WorldWideScience

Sample records for nuclear core analysis

  1. CAC - NUCLEAR THERMAL ROCKET CORE ANALYSIS CODE

    Science.gov (United States)

    Clark, J. S.

    1994-01-01

    One of the most important factors in the development of nuclear rocket engine designs is to be able to accurately predict temperatures and pressures throughout a fission nuclear reactor core with axial hydrogen flow through circular coolant passages. CAC is an analytical prediction program to study the heat transfer and fluid flow characteristics of a circular coolant passage. CAC predicts as a function of time axial and radial fluid conditions, passage wall temperatures, flow rates in each coolant passage, and approximate maximum material temperatures. CAC incorporates the hydrogen properties model STATE to provide fluid-state relations, thermodynamic properties, and transport properties of molecular hydrogen in any fixed ortho-para combination. The program requires the general core geometry, the core material properties as a function of temperature, the core power profile, and the core inlet conditions as function of time. Although CAC was originally developed in FORTRAN IV for use on an IBM 7094, this version is written in ANSI standard FORTRAN 77 and is designed to be machine independent. It has been successfully compiled on IBM PC series and compatible computers running MS-DOS with Lahey F77L, a Sun4 series computer running SunOS 4.1.1, and a VAX series computer running VMS 5.4-3. CAC requires 300K of RAM under MS-DOS, 422K of RAM under SunOS, and 220K of RAM under VMS. No sample executable is provided on the distribution medium. Sample input and output data are included. The standard distribution medium for this program is a 5.25 inch 360K MS-DOS format diskette. CAC was developed in 1966, and this machine independent version was released in 1992. IBM-PC and IBM are registered trademarks of International Business Machines. Lahey F77L is a registered trademark of Lahey Computer Systems, Inc. SunOS is a trademark of Sun Microsystems, Inc. VMS is a trademark of Digital Equipment Corporation. MS-DOS is a registered trademark of Microsoft Corporation.

  2. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  3. Analysis of suprathermal nuclear processes in the solar core plasma

    Science.gov (United States)

    Voronchev, Victor T.; Nakao, Yasuyuki; Watanabe, Yukinobu

    2017-04-01

    A consistent model for the description of suprathermal processes in the solar core plasma naturally triggered by fast particles generated in exoergic nuclear reactions is formulated. This model, based on the formalism of in-flight reaction probability, operates with different methods of treating particle slow-down in the plasma, and allows for the influence of electron degeneracy and electron screening on processes in the matter. The model is applied to examine slowing-down of 8.7 MeV α-particles produced in the {}7{Li}(p,α )α reaction of the pp chain, and to analyze suprathermal processes in the solar CNO cycle induced by them. Particular attention is paid to the suprathermal {}14{{N}}{(α ,{{p}})}17{{O}} reaction unappreciated in standard solar model simulations. It is found that an appreciable non-standard (α ,p) nuclear flow due to this reaction appears in the matter and modifies running of the CNO cycle in ∼95% of the solar core region. In this region at R> 0.1{R}ȯ , normal branching of nuclear flow {}14{{N}}≤ftarrow {}17{{O}}\\to {(}18{{F}})\\to {}18{{O}} transforms to abnormal sequential flow {}14{{N}}\\to {}17{{O}}\\to {(}18{{F}})\\to {}18{{O}}, altering some element abundances. In particular, nuclear network calculations reveal that in the outer core the abundances of 17O and 18O isotopes can increase by a factor of 20 as compared with standard estimates. A conjecture is made that other CNO suprathermal (α ,p) reactions may also affect abundances of CNO elements, including those generating solar neutrinos.

  4. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-06-01

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

  6. Heat transfer analysis of fuel assemblies in a heterogeneous gas core nuclear rocket

    Science.gov (United States)

    Watanabe, Yoichi; Appelbaum, Jacob; Diaz, Nils; Maya, Isaac

    1991-01-01

    Heat transfer problems of a heterogeneous gaseous core nuclear rocket were studied. The reactor core consists of 1.5-m long hexagonal fuel assemblies filled with pressurized uranium tetrafluoride (UF4) gas. The fuel gas temperature ranges from 3500 to 7000 K at a nominal operating condition of 40 atm. Each fuel assembly has seven coolant tubes, through which hydrogen propellant flows. The propellant temperature is not constrained by the fuel temperature but by the maximum temperature of the graphite coolant tube. For a core achieving a fission power density of 1000 MW/cu m, the propellant core exit temperature can be as high as 3200 K. The physical size of a 1250 MW gaseous core nuclear rocket is comparable with that of a NERVA-type solid core nuclear rocket. The engine can deliver a specific impulse of 1020 seconds and a thrust of 330 kN.

  7. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    Energy Technology Data Exchange (ETDEWEB)

    Prabhu Gaunkar, N., E-mail: neelampg@iastate.edu; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C. [Department of Electrical and Computer Engineering, Iowa State University, Ames, Iowa 50011 (United States); Bulu, I.; Ganesan, K.; Song, Y. Q. [Schlumberger-Doll Research, Cambridge, Massachusetts 02139 (United States)

    2015-05-07

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors.

  8. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, Jonathan, E-mail: JMitchell16@slb.com; Fordham, Edmund J. [Schlumberger Gould Research, High Cross, Madingley Road, Cambridge CB3 0EL (United Kingdom)

    2014-11-15

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general.

  9. Contributed review: nuclear magnetic resonance core analysis at 0.3 T.

    Science.gov (United States)

    Mitchell, Jonathan; Fordham, Edmund J

    2014-11-01

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general.

  10. Analysis of nuclear characteristics and fuel economics for PWR core with homogeneous thorium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Joo, H. K.; Noh, J. M.; Yoo, J. W.; Song, J. S.; Kim, J. C.; Noh, T. W

    2000-12-01

    The nuclear core characteristics and economics of an once-through homogenized thorium cycle for PWR were analyzed. The lattice code, HELIOS has been qualified against BNL and B and W critical experiments and the IAEA numerical benchmark problem in advance of the core analysis. The infinite multiplication factor and the evolution of main isotopes with fuel burnup were investigated for the assessment of depletion charateristics of thorium fuel. The reactivity of thorium fuel at the beginning of irradiation is smaller than that of uranium fuel having the same inventory of {sup 235}U, but it decrease with burnup more slowly than in UO{sub 2} fuel. The gadolinia worth in thorium fuel assembly is also slightly smaller than in UO{sub 2} fuel. The inventory of {sup 233}U which is converted from {sup 232}Th is proportional to the initial mass of {sup 232}Th and is about 13kg per one tones of initial heavy metal mass. The followings are observed for thorium fuel cycle compared with UO{sub 2} cycle ; shorter cycle length, more positive MTC at EOC, more negative FTC, similar boron worth and control rod. Fuel economics of thorium cycle was analyzed by investigating the natural uranium requirements, the separative work requirements, and the cost for burnable poison rods. Even though less number of burnable poison rods are required in thorium fuel cycle, the costs for the natural uranium requirements and the separative work requirements are increased in thorium fuel cycle. So within the scope of this study, once through cycle concept, homogenized fuel concept, the same fuel management scheme as uranium cycle, the thorium fuel cycle for PWR does not have any economic incentives in preference to uranium.

  11. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores. [PEBBLE code

    Energy Technology Data Exchange (ETDEWEB)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.

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

    Science.gov (United States)

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

    2007-01-01

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

  13. Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

    Science.gov (United States)

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

    2013-01-01

    Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

  14. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  15. Nuclear gas core propulsion research program

    Science.gov (United States)

    Diaz, Nils J.; Dugan, Edward T.; Anghaie, Samim

    1993-01-01

    Viewgraphs on the nuclear gas core propulsion research program are presented. The objectives of this research are to develop models and experiments, systems, and fuel elements for advanced nuclear thermal propulsion rockets. The fuel elements under investigation are suitable for gas/vapor and multiphase fuel reactors. Topics covered include advanced nuclear propulsion studies, nuclear vapor thermal rocket (NVTR) studies, and ultrahigh temperature nuclear fuels and materials studies.

  16. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant; Analisis de documentos de los materiales de la envolvente del nucleo del reactor nuclear de la CLV U-1

    Energy Technology Data Exchange (ETDEWEB)

    Zamora R, L.; Medina F, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  17. Development of core design/analysis technology for integral reactor; verification of SMART nuclear design by Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Hong, In Seob; Han, Beom Seok; Jeong, Jong Seong [Seoul National University, Seoul (Korea)

    2002-03-01

    The objective of this project is to verify neutronics characteristics of the SMART core design as to compare computational results of the MCNAP code with those of the MASTER code. To achieve this goal, we will analyze neutronics characteristics of the SMART core using the MCNAP code and compare these results with results of the MASTER code. We improved parallel computing module and developed error analysis module of the MCNAP code. We analyzed mechanism of the error propagation through depletion computation and developed a calculation module for quantifying these errors. We performed depletion analysis for fuel pins and assemblies of the SMART core. We modeled a 3-D structure of the SMART core and considered a variation of material compositions by control rods operation and performed depletion analysis for the SMART core. We computed control-rod worths of assemblies and a reactor core for operation of individual control-rod groups. We computed core reactivity coefficients-MTC, FTC and compared these results with computational results of the MASTER code. To verify error analysis module of the MCNAP code, we analyzed error propagation through depletion of the SMART B-type assembly. 18 refs., 102 figs., 36 tabs. (Author)

  18. Scale Model Test and Transient Analysis of Steam Injector Driven Passive Core Injection System for Innovative-Simplified Nuclear Power Plant

    Science.gov (United States)

    Ohmori, Shuichi; Narabayashi, Tadashi; Mori, Michitsugu

    A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. This provides SI with capability to serve also as a direct-contact feed-water heater that heats up feed-water by using extracted steam from turbine. Our technology development aims to significantly simplify equipment and reduce physical quantities by applying "high-efficiency SI", which are applicable to a wide range of operation regimes beyond the performance and applicable range of existing SIs and enables unprecedented multistage and parallel operation, to the low-pressure feed-water heaters and emergency core cooling system of nuclear power plants, as well as achieve high inherent safety to prevent severe accidents by keeping the core covered with water (a severe accident-free concept). This paper describes the results of the scale model test, and the transient analysis of SI-driven passive core injection system (PCIS).

  19. Safe, Compact Nuclear Propulsion: Solid Core Nuclear Propulsion Concept

    Science.gov (United States)

    1988-10-01

    analysis group developed ROM component cost estimates given in representative ranges. 4.1 Engine System A representative nuclear thermal rocket engine...nuclear thermal rocket engine cycle balance computer code. The design requirements for the engine were: Thrust : 15,000 lbf Champer Pressure 500 psia...advanced nuclear thermal rockets . Our analysis was based on an examination of presentation material provided by Martin, some independent calculations of

  20. Wire core reactor for nuclear thermal propulsion

    Science.gov (United States)

    Harty, Richard B.; Brengle, Robert G.

    1993-01-01

    Studies have been performed of a compact high-performance nuclear rocket reactor that incorporates a tungsten alloy wire fuel element. This reactor, termed the wire core reactor, can deliver a specific impulse of 1,000 s using an expander cycle and a nozzle expansion ratio of 500 to 1. The core is constructed of layers of 0.8-mm-dia fueled tungsten wires wound over alternate layers of spacer wires, which forms a rugged annular lattice. Hydrogen flow in the core is annular, flowing from inside to outside. In addition to the concepts compact size and good heat transfer, the core has excellent power-flow matching features and can resist vibration and thermal stresses during star-up and shutdown.

  1. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    Science.gov (United States)

    Walton, James T.

    1992-01-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code.

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

    Science.gov (United States)

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

    2010-01-01

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

  3. Computer code and users' guide for the preliminary analysis of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239b

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, R.A.; Smith, W.W.

    1976-06-30

    The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The second volume describes the computer code and users' guide for the preliminary analysis of the system.

  4. Open cycle gas core nuclear rockets

    Science.gov (United States)

    Ragsdale, Robert

    1991-01-01

    The open cycle gas core engine is a nuclear propulsion device. Propulsion is provided by hot hydrogen which is heated directly by thermal radiation from the nuclear fuel. Critical mass is sustained in the uranium plasma in the center. It has typically 30 to 50 kg of fuel. It is a thermal reactor in the sense that fissions are caused by absorption of thermal neutrons. The fast neutrons go out to an external moderator/reflector material and, by collision, slow down to thermal energy levels, and then come back in and cause fission. The hydrogen propellant is stored in a tank. The advantage of the concept is very high specific impulse because you can take the plasma to any temperature desired by increasing the fission level by withdrawing or turning control rods or control drums.

  5. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    Science.gov (United States)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

  6. Forces in bolted joints: analysis methods and test results utilized for nuclear core applications (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Crescimanno, P.J.; Keller, K.L.

    1981-03-01

    Analytical methods and test data employed in the core design of bolted joints for the LWBR core are presented. The effects of external working loads, thermal expansion, and material stress relaxation are considered in the formulation developed to analyze joint performance. Extensions of these methods are also provided for bolted joints having both axial and bending flexibilities, and for the effect of plastic deformation on internal forces developed in a bolted joint. Design applications are illustrated by examples.

  7. Development of Core Monitoring System for Nuclear Power Plants (I)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.H.; Kim, Y.B.; Park, M.G; Lee, E.K.; Shin, H.C.; Lee, D.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    1.Object and Necessity of the Study -The main objectives of this study are (1)conversion of APOLLO version BEACON system to HP-UX version core monitoring system, (2)provision of the technical bases to enhance the in-house capability of developing more advanced core monitoring system. 2.Results of the Study - In this study, the revolutionary core monitoring technologies such as; nodal analysis and isotope depletion calculation method, advanced schemes for power distribution control, and treatment of nuclear databank were established. The verification and validation work has been successfully performed by comparing the results with those of the design code and measurement data. The advanced graphic user interface and plant interface method have been implemented to ensure the future upgrade capability. The Unix shell scripts and system dependent software are also improved to support administrative functions of the system. (author). 14 refs., 112 figs., 52 tabs.

  8. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  9. Depth-charge static and time-dependence perturbation/sensitivity system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1981-09-01

    This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code block for both static and time-dependence perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Labortary. The DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analysis of realistic multidimensional reactor models.

  10. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  11. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  12. Nuclear statistical equilibrium at core-collapse supernova

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    A new improved nuclear partition function is employed to calculate the nuclear statistical equilibrium (NSE) in core-collapse supernova environment. The results show that the change of nucleus abundance is slight even though the temperature is higher than 1011 K when shock propagates, which indicates that the effect of the nuclear partition function is not so important as shown in the previous calculations, but it can also be considered in detailed simulation if it is sensitive to weak interaction rates in core-collapse supernova.

  13. Nuclear factor Y regulates ancient budgerigar hepadnavirus core promoter activity.

    Science.gov (United States)

    Shen, Zhongliang; Liu, Yanfeng; Luo, Mengjun; Wang, Wei; Liu, Jing; Liu, Wei; Pan, Shaokun; Xie, Youhua

    2016-09-16

    Endogenous viral elements (EVE) in animal genomes are the fossil records of ancient viruses and provide invaluable information on the origin and evolution of extant viruses. Extant hepadnaviruses include avihepadnaviruses of birds and orthohepadnaviruses of mammals. The core promoter (Cp) of hepadnaviruses is vital for viral gene expression and replication. We previously identified in the budgerigar genome two EVEs that contain the full-length genome of an ancient budgerigar hepadnavirus (eBHBV1 and eBHBV2). Here, we found eBHBV1 Cp and eBHBV2 Cp were active in several human and chicken cell lines. A region from nt -85 to -11 in eBHBV1 Cp was critical for the promoter activity. Bioinformatic analysis revealed a putative binding site of nuclear factor Y (NF-Y), a ubiquitous transcription factor, at nt -64 to -50 in eBHBV1 Cp. The NF-Y core binding site (ATTGG, nt -58 to -54) was essential for eBHBV1 Cp activity. The same results were obtained with eBHBV2 Cp and duck hepatitis B virus Cp. The subunit A of NF-Y (NF-YA) was recruited via the NF-Y core binding site to eBHBV1 Cp and upregulated the promoter activity. Finally, the NF-Y core binding site is conserved in the Cps of all the extant avihepadnaviruses but not of orthohepadnaviruses. Interestingly, a putative and functionally important NF-Y core binding site is located at nt -21 to -17 in the Cp of human hepatitis B virus. In conclusion, our findings have pinpointed an evolutionary conserved and functionally critical NF-Y binding element in the Cps of avihepadnaviruses.

  14. Development of an automated core model for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R.D.

    1998-12-31

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

  15. 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.

  16. Nuclear Human Resources Development Program using Educational Core Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yu Sun; Hong, Soon Kwan [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides.

  17. Economic Analysis of Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Man Ki; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Oh, K. B

    2006-12-15

    It has been well recognized that securing economic viabilities along with technologies are very important elements in the successful implementation of nuclear R and D projects. The objective of the Project is to help nuclear energy to be utilized in an efficient way by analyzing major issues related with nuclear economics. The study covers following subjects: the role of nuclear in the future electric supply system, economic analysis of nuclear R and D project, contribution to the regional economy from nuclear power. In addition, the study introduces the international cooperation in the methodological area of efficient use of nuclear energy by surveying the international activities related with nuclear economics.

  18. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Science.gov (United States)

    Rabir, Mohamad Hairie; Hah, Chang Joo; Ju, Cho Sung

    2015-04-01

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  19. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  20. 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.

  1. Robustness of nuclear core activity reconstruction by data assimilation

    Energy Technology Data Exchange (ETDEWEB)

    Bouriquet, Bertrand, E-mail: bertrand.bouriquet@cerfacs.f [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Argaud, Jean-Philippe [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Erhard, Patrick [Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Massart, Sebastien [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Poncot, Angelique [Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Ricci, Sophie [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Thual, Olivier [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No. 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Universite de Toulouse, INPT, UPS, IMFT, Allee Camille Soula, F-31400 Toulouse (France)

    2011-02-11

    We apply a data assimilation technique, inspired from meteorological applications, to perform an optimal reconstruction of the neutronic activity field in a nuclear core. Both measurements and information coming from a numerical model are used. We first study the robustness of the method when the amount of measured information decreases. We then study the influence of the nature of the instruments and their spatial repartition on the efficiency of the field reconstruction.

  2. Piezoelectric material for use in a nuclear reactor core

    Science.gov (United States)

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R.

    2012-05-01

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d33 was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d33 for many as-grown samples.

  3. Review of coaxial flow gas core nuclear rocket fluid mechanics

    Science.gov (United States)

    Weinstein, H.

    1976-01-01

    Almost all of the fluid mechanics research associated with the coaxial flow gas core reactor ended abruptly with the interruption of NASA's space nuclear program because of policy and budgetary considerations in 1973. An overview of program accomplishments is presented through a review of the experiments conducted and the analyses performed. Areas are indicated where additional research is required for a fuller understanding of cavity flow and of the factors which influence cold and hot flow containment. A bibliography is included with graphic material.

  4. Computational fluid dynamics analysis of core bypass flow and crossflow in a prismatic very high temperature gas-cooled nuclear reactor based on a two-layer block model

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Huhu, E-mail: huhuwang@tamu.edu [Department of Nuclear Engineering, Texas A and M University, 3133 TAMU, College Station, TX 77840 (United States); Dominguez-Ontiveros, Elvis, E-mail: elvisdom@tamu.edu [Department of Nuclear Engineering, Texas A and M University, 3133 TAMU, College Station, TX 77840 (United States); Hassan, Yassin A., E-mail: y-hassan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, 3133 TAMU, College Station, TX 77840 (United States); Department of Mechanical Engineering, Texas A and M University, 3123 TAMU, College Station, TX 77840 (United States)

    2014-03-15

    Highlights: • A CFD model was built based on a two-layer block experimental facility at Texas A and M University. • The coolant characterizations within the uniform and wedge-shaped crossflow gap regions were investigated. • The influence on the coolant distribution from the bypass flow gap width was studied. • Discretization and iterative errors involved in the simulations were quantified. - Abstract: The very high temperature gas-cooled nuclear reactor (VHTR) has been designated as one of the promising reactors that will serve for the Next Generation (Generation IV) Nuclear Plant. For a prismatic VHTR core, the bypass flow and crossflow phenomena are important design considerations. To investigate the coolant distribution in the reactor core based on the two-layer block facility built at Texas A and M University, a three-dimensional steady-state CFD analysis was performed using the commercial code STAR-CCM+ v6.04. Results from this work serve as a guideline and validating source for the related experiments. A grid independence study was conducted to quantify related errors in the simulations. The simulation results show that the bypass flow fraction was not a strong function of the Reynolds number. The presence of the crossflow gap had a significant effect on the distribution of the coolant in the core. Uniform and wedge-shape crossflow gaps were studied. It was found that a significant secondary flow in the crossflow gap region moved from the bypass flow gap toward coolant holes, which resulted in up to a 28% reduction of the coolant mass flow rate in the bypass flow gap.

  5. Nuclear equation of state for core-collapse supernova simulations with realistic nuclear forces

    Energy Technology Data Exchange (ETDEWEB)

    Togashi, H., E-mail: hajime.togashi@riken.jp [Nishina Center for Accelerator-Based Science, Institute of Physical and Chemical Research (RIKEN), 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Nakazato, K. [Faculty of Arts and Science, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Takehara, Y.; Yamamuro, S.; Suzuki, H. [Department of Physics, Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan); Takano, M. [Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Department of Pure and Applied Physics, Graduate School of Advanced Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan)

    2017-05-15

    A new table of the nuclear equation of state (EOS) based on realistic nuclear potentials is constructed for core-collapse supernova numerical simulations. Adopting the EOS of uniform nuclear matter constructed by two of the present authors with the cluster variational method starting from the Argonne v18 and Urbana IX nuclear potentials, the Thomas–Fermi calculation is performed to obtain the minimized free energy of a Wigner–Seitz cell in non-uniform nuclear matter. As a preparation for the Thomas–Fermi calculation, the EOS of uniform nuclear matter is modified so as to remove the effects of deuteron cluster formation in uniform matter at low densities. Mixing of alpha particles is also taken into account following the procedure used by Shen et al. (1998, 2011). The critical densities with respect to the phase transition from non-uniform to uniform phase with the present EOS are slightly higher than those with the Shen EOS at small proton fractions. The critical temperature with respect to the liquid–gas phase transition decreases with the proton fraction in a more gradual manner than in the Shen EOS. Furthermore, the mass and proton numbers of nuclides appearing in non-uniform nuclear matter with small proton fractions are larger than those of the Shen EOS. These results are consequences of the fact that the density derivative coefficient of the symmetry energy of our EOS is smaller than that of the Shen EOS.

  6. Nuclear Reactor Engineering Analysis Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  7. Reducing the risk to Mars: The gas core nuclear rocket

    Science.gov (United States)

    Howe, S. D.; DeVolder, B.; Thode, L.; Zerkle, D.

    1998-01-01

    The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas-core nuclear rocket (GCNR) has the potential to be such a system. The authors have completed a comparative study of the potential impact that a GCNR could have on a manned Mars mission. The total IMLEO, transit times, and accumulated radiation dose to the crew will be compared with the NASA Design Reference Missions.

  8. Expression Analysis and Nuclear Import Study of Full-length Isoforms Importin α as 6x Histidin-tagged Fusion Protein on the Intracellular Localization of Recombinant HBV Core Protein

    Directory of Open Access Journals (Sweden)

    Aris Haryanto

    2015-10-01

    Full Text Available Isoform importin α molecules play a central role in the classical nuclear import pathway, that occurs throughthe nuclear pore complex (NPC and typically requires a specific nuclear localization signal (NLS. In this study,it was investigated the role of isoforms importin α in the nuclear import of wild type recombinant hepatitis B viruscore protein (WT rHBc, phosphorylated recombinant HBV core (rHBc and recombinant HBV core without NLSby co-immunoprecipitation. Four recombinant full-length isoforms importin α as 6x histidin-tagged fusion proteinwere expressed and analysed from expression plasmid vectors Rch1, pHM 1969, pHM 1967 and pHM 1965. Theresults indicated that importin α-1, importin α-3, importin α-4 and importin α-5 can be expressed and isolatedfrom E. coli transformed recombinant DNA plasmid as protein in size around 58-60 kDa. By the nuclear transportstudy shown that isoforms importin α are involved in the nuclear import of WT rHBc, phosphorylated rHBc andrHBc without NLS. It also indicated that they have an important role for nuclear transport of from cytoplasm intothe nucleus.Keywords: NPC, NLS, importin α, importin β, isoforms importin α as 6x histidin-tagged fusion protein, WTrHBc, SV40 Tag, co-immunoprecipitation, westernblotting.

  9. Bimodal Nuclear Thermal Rocket Analysis Developments

    Science.gov (United States)

    Belair, Michael; Lavelle, Thomas; Saimento, Charles; Juhasz, Albert; Stewart, Mark

    2014-01-01

    Nuclear thermal propulsion has long been considered an enabling technology for human missions to Mars and beyond. One concept of operations for these missions utilizes the nuclear reactor to generate electrical power during coast phases, known as bimodal operation. This presentation focuses on the systems modeling and analysis efforts for a NERVA derived concept. The NERVA bimodal operation derives the thermal energy from the core tie tube elements. Recent analysis has shown potential temperature distributions in the tie tube elements that may limit the thermodynamic efficiency of the closed Brayton cycle used to generate electricity with the current design. The results of this analysis are discussed as well as the potential implications to a bimodal NERVA type reactor.

  10. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    Science.gov (United States)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  11. Localization of Vibrating Noise Sources in Nuclear Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Hultqvist, Pontus

    2004-09-01

    In this thesis the possibility of locating vibrating noise sources in a nuclear reactor core from the neutron noise has been investigated using different localization methods. The influence of the vibrating noise source has been considered to be a small perturbation of the neutron flux inside the reactor. Linear perturbation theory has been used to construct the theoretical framework upon which the localization methods are based. Two different cases have been considered: one where a one-dimensional one-group model has been used and another where a two-dimensional two-energy group noise simulator has been used. In the first case only one localization method is able to determine the position with good accuracy. This localization method is based on finding roots of an equation and is sensitive to other perturbations of the neutron flux. It will therefore work better with the assistance of approximative methods that reconstruct the noise source to determine if the results are reliable or not. In the two-dimensional case the results are more promising. There are several different localization techniques that reproduce both the vibrating noise source position and the direction of vibration with enough precision. The approximate methods that reconstruct the noise source are substantially better and are able to support the root finding method in a more constructive way. By combining the methods, the results will be more reliable.

  12. Analysis of circuits including magnetic cores (MTRAC)

    Science.gov (United States)

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

    1972-01-01

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

  13. Economic analysis of nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Dong; Lee, M. K.; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Kim, H. S

    2000-12-01

    This study identified the role of nuclear energy in the following three major aspects. First of all, this study carried out cost effectiveness of nuclear as a CDM technology, which is one of means of GHG emission reduction in UNFCCC. Secondly, environmental externalities caused by air pollutants emitted by power options were estimated. The 'observed market behaviour' method and 'responses to hypothetical market' method were used to estimate objectively the environmental external costs by electric source, respectively. Finally, the role of nuclear power in securing electricity supply in a liberalized electricity market was analyzed. This study made efforts to investigate whether nuclear power generation with high investment cost could be favored in a liberalized market by using 'option value' analysis of investments.

  14. Economic analysis of nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Dong; Lee, M. K.; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Kim, H. S

    2000-12-01

    This study identified the role of nuclear energy in the following three major aspects. First of all, this study carried out cost effectiveness of nuclear as a CDM technology, which is one of means of GHG emission reduction in UNFCCC. Secondly, environmental externalities caused by air pollutants emitted by power options were estimated. The 'observed market behaviour' method and 'responses to hypothetical market' method were used to estimate objectively the environmental external costs by electric source, respectively. Finally, the role of nuclear power in securing electricity supply in a liberalized electricity market was analyzed. This study made efforts to investigate whether nuclear power generation with high investment cost could be favored in a liberalized market by using 'option value' analysis of investments.

  15. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Directory of Open Access Journals (Sweden)

    Andrea Cerutti

    Full Text Available Hepatitis C virus (HCV infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS, but no nuclear export signal (NES has yet been identified.We show here that the aa(109-133 region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126 in the identified NES or in the sequence encoding the mature core aa(1-173 significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  16. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Science.gov (United States)

    Cerutti, Andrea; Maillard, Patrick; Minisini, Rosalba; Vidalain, Pierre-Olivier; Roohvand, Farzin; Pecheur, Eve-Isabelle; Pirisi, Mario; Budkowska, Agata

    2011-01-01

    Hepatitis C virus (HCV) infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS), but no nuclear export signal (NES) has yet been identified.We show here that the aa(109-133) region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126) in the identified NES or in the sequence encoding the mature core aa(1-173) significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  17. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, C.L.; Stewart, C.W.; Cena, R.J.; Rowe, D.S.; Sutey, A.M.

    1976-03-01

    The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels.

  18. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Webb, B.J.

    1988-01-01

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.

  19. Multiphysics Nuclear Thermal Rocket Thrust Chamber Analysis

    Science.gov (United States)

    Wang, Ten-See

    2005-01-01

    The objective of this effort is t o develop an efficient and accurate thermo-fluid computational methodology to predict environments for hypothetical thrust chamber design and analysis. The current task scope is to perform multidimensional, multiphysics analysis of thrust performance and heat transfer analysis for a hypothetical solid-core, nuclear thermal engine including thrust chamber and nozzle. The multiphysics aspects of the model include: real fluid dynamics, chemical reactivity, turbulent flow, and conjugate heat transfer. The model will be designed to identify thermal, fluid, and hydrogen environments in all flow paths and materials. This model would then be used to perform non- nuclear reproduction of the flow element failures demonstrated in the Rover/NERVA testing, investigate performance of specific configurations and assess potential issues and enhancements. A two-pronged approach will be employed in this effort: a detailed analysis of a multi-channel, flow-element, and global modeling of the entire thrust chamber assembly with a porosity modeling technique. It is expected that the detailed analysis of a single flow element would provide detailed fluid, thermal, and hydrogen environments for stress analysis, while the global thrust chamber assembly analysis would promote understanding of the effects of hydrogen dissociation and heat transfer on thrust performance. These modeling activities will be validated as much as possible by testing performed by other related efforts.

  20. Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

    Science.gov (United States)

    Weinstein, H.; Lavan, Z.

    1975-01-01

    Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

  1. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  2. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Directory of Open Access Journals (Sweden)

    Gholam Reza Ansarifar

    2015-12-01

    Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  3. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Energy Technology Data Exchange (ETDEWEB)

    Ansarifar, Gholam Reza; Saadatzi, Saeed [Dept. of Nuclear Engineering, Faculty of Advanced Sciences and Technology, University of Isfahan, Isfahan (Iran, Islamic Republic of)

    2015-12-15

    One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC), which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR) power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  4. Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system

    Science.gov (United States)

    Kahook, Samer D.; Dugan, Edward T.

    1991-01-01

    Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (≊20%), small radiator size (≊5 m2/MWe), and high specific power (≊5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

  5. Nucleoporins as components of the nuclear pore complex core structure and Tpr as the architectural element of the nuclear basket.

    Science.gov (United States)

    Krull, Sandra; Thyberg, Johan; Björkroth, Birgitta; Rackwitz, Hans-Richard; Cordes, Volker C

    2004-09-01

    The vertebrate nuclear pore complex (NPC) is a macromolecular assembly of protein subcomplexes forming a structure of eightfold radial symmetry. The NPC core consists of globular subunits sandwiched between two coaxial ring-like structures of which the ring facing the nuclear interior is capped by a fibrous structure called the nuclear basket. By postembedding immunoelectron microscopy, we have mapped the positions of several human NPC proteins relative to the NPC core and its associated basket, including Nup93, Nup96, Nup98, Nup107, Nup153, Nup205, and the coiled coil-dominated 267-kDa protein Tpr. To further assess their contributions to NPC and basket architecture, the genes encoding Nup93, Nup96, Nup107, and Nup205 were posttranscriptionally silenced by RNA interference (RNAi) in HeLa cells, complementing recent RNAi experiments on Nup153 and Tpr. We show that Nup96 and Nup107 are core elements of the NPC proper that are essential for NPC assembly and docking of Nup153 and Tpr to the NPC. Nup93 and Nup205 are other NPC core elements that are important for long-term maintenance of NPCs but initially dispensable for the anchoring of Nup153 and Tpr. Immunogold-labeling for Nup98 also results in preferential labeling of NPC core regions, whereas Nup153 is shown to bind via its amino-terminal domain to the nuclear coaxial ring linking the NPC core structures and Tpr. The position of Tpr in turn is shown to coincide with that of the nuclear basket, with different Tpr protein domains corresponding to distinct basket segments. We propose a model in which Tpr constitutes the central architectural element that forms the scaffold of the nuclear basket.

  6. Nuclear cardiology core syllabus of the European Association of Cardiovascular Imaging (EACVI).

    Science.gov (United States)

    Gimelli, Alessia; Neglia, Danilo; Schindler, Thomas H; Cosyns, Bernard; Lancellotti, Patrizio; Kitsiou, Anastasia

    2015-04-01

    The European Association of Cardiovascular Imaging (EACVI) Core Syllabus for Nuclear Cardiology is now available online. The syllabus lists key elements of knowledge in nuclear cardiology. It represents a framework for the development of training curricula and provides expected knowledge-based learning outcomes to the nuclear cardiology trainees. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2015. For permissions please email: journals.permissions@oup.com.

  7. Hanging core support system for a nuclear reactor. [LMFBR

    Science.gov (United States)

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-04-26

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  8. Nuclear Power Plant Module, NPP-1: Nuclear Power Cost Analysis.

    Science.gov (United States)

    Whitelaw, Robert L.

    The purpose of the Nuclear Power Plant Modules, NPP-1, is to determine the total cost of electricity from a nuclear power plant in terms of all the components contributing to cost. The plan of analysis is in five parts: (1) general formulation of the cost equation; (2) capital cost and fixed charges thereon; (3) operational cost for labor,…

  9. Differential influence of instruments in nuclear core activity evaluation by data assimilation

    Energy Technology Data Exchange (ETDEWEB)

    Bouriquet, Bertrand, E-mail: bertrand.bouriquet@cerfacs.f [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Argaud, Jean-Philippe [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Erhard, Patrick [Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Massart, Sebastien [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Poncot, Angelique [Electricite de France, 1 avenue du General de Gaulle, F-92141 Clamart Cedex (France); Ricci, Sophie [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Thual, Olivier [Sciences de l' Univers au CERFACS, URA CERFACS/CNRS No 1875, 42 avenue Gaspard Coriolis, F-31057 Toulouse Cedex 01 (France); Universite de Toulouse, INPT, UPS, IMFT, Allee Camille Soula, F-31400 Toulouse (France)

    2011-01-21

    The global neutronic activity fields of a nuclear core can be reconstructed using data assimilation. Indeed, data assimilation allows to combine both measurements from instruments and information from a model, to evaluate the best possible neutronic activity within the core. We present and apply a specific procedure which evaluates the influence of measures by adding or removing instruments in a given measurement network (possibly empty). The study of various network configurations for the instruments in the nuclear core establishes that the influence of the instruments depends both on the independent instrumentation location and on the chosen network.

  10. Verification and uncertainty evaluation of CASMO-3/MASTER nuclear analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jae Seung; Cho, Byung Oh; Joo, Han Kyu; Zee, Sung Quun; Lee, Chung Chan; Park, Sang Yoon

    2000-06-01

    MASTER is a nuclear design code developed by KAERI. It uses group constants generated by CASMO-3 developed by Studsvik. In this report the verification and evaluation of uncertainty were performed for the code system application in nuclear reactor core analysis and design. The verification is performed via various benchmark comparisons for static and transient core condition, and core follow calculations with startup physics test predictions of total 14 cycles of pressurized water reactors. Benchmark calculation include comparisons with reference solutions of IAEA and OECA/NEA problems and critical experiment measurements. The uncertainty evaluation is focused to safety related parameters such as power distribution, reactivity coefficients, control rod worth and core reactivity. It is concluded that CASMO-3/MASTER can be applied for PWR core nuclear analysis and design without any bias factors. Also, it is verified that the system can be applied for SMART core, via supplemental comparisons with reference calculations by MCNP which is a probabilistic nuclear calculation code.

  11. Reducing numerical costs for core wide nuclear reactor CFD simulations by the Coarse-Grid-CFD

    Science.gov (United States)

    Viellieber, Mathias; Class, Andreas G.

    2013-11-01

    Traditionally complete nuclear reactor core simulations are performed with subchannel analysis codes, that rely on experimental and empirical input. The Coarse-Grid-CFD (CGCFD) intends to replace the experimental or empirical input with CFD data. The reactor core consists of repetitive flow patterns, allowing the general approach of creating a parametrized model for one segment and composing many of those to obtain the entire reactor simulation. The method is based on a detailed and well-resolved CFD simulation of one representative segment. From this simulation we extract so-called parametrized volumetric forces which close, an otherwise strongly under resolved, coarsely-meshed model of a complete reactor setup. While the formulation so far accounts for forces created internally in the fluid others e.g. obstruction and flow deviation through spacers and wire wraps, still need to be accounted for if the geometric details are not represented in the coarse mesh. These are modelled with an Anisotropic Porosity Formulation (APF). This work focuses on the application of the CGCFD to a complete reactor core setup and the accomplishment of the parametrization of the volumetric forces.

  12. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  13. Nuclear Proliferation Technology Trends Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zentner, Michael D.; Coles, Garill A.; Talbert, Robert J.

    2005-10-04

    A process is underway to develop mature, integrated methodologies to address nonproliferation issues. A variety of methodologies (both qualitative and quantitative) are being considered. All have one thing in common, a need for a consistent set of proliferation related data that can be used as a basis for application. One approach to providing a basis for predicting and evaluating future proliferation events is to understand past proliferation events, that is, the different paths that have actually been taken to acquire or attempt to acquire special nuclear material. In order to provide this information, this report describing previous material acquisition activities (obtained from open source material) has been prepared. This report describes how, based on an evaluation of historical trends in nuclear technology development, conclusions can be reached concerning: (1) The length of time it takes to acquire a technology; (2) The length of time it takes for production of special nuclear material to begin; and (3) The type of approaches taken for acquiring the technology. In addition to examining time constants, the report is intended to provide information that could be used to support the use of the different non-proliferation analysis methodologies. Accordingly, each section includes: (1) Technology description; (2) Technology origin; (3) Basic theory; (4) Important components/materials; (5) Technology development; (6) Technological difficulties involved in use; (7) Changes/improvements in technology; (8) Countries that have used/attempted to use the technology; (9) Technology Information; (10) Acquisition approaches; (11) Time constants for technology development; and (12) Required Concurrent Technologies.

  14. Degraded core analysis for the PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-10-01

    The paper presents an analysis of the probability and consequences of degraded core accidents for the PWR. The article is based on a paper which was presented by the author to the Sizewell-B public inquiry. Degraded core accidents are examined with respect to:- the initiating events, safety plant failure, and processes with a bearing on containment failure. Accident types and frequencies are discussed, as well as the dispersion of radionuclides. Accident risks, i.e. individual and societal risks in degraded core accidents are assessed from:- the amount of radionuclides released, the weather, the population distribution, and the accident frequencies. Uncertainties in the assessment of degraded core accidents are also summarized. (U.K.).

  15. Overview on Hydrate Coring, Handling and Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jon Burger; Deepak Gupta; Patrick Jacobs; John Shillinglaw

    2003-06-30

    Gas hydrates are crystalline, ice-like compounds of gas and water molecules that are formed under certain thermodynamic conditions. Hydrate deposits occur naturally within ocean sediments just below the sea floor at temperatures and pressures existing below about 500 meters water depth. Gas hydrate is also stable in conjunction with the permafrost in the Arctic. Most marine gas hydrate is formed of microbially generated gas. It binds huge amounts of methane into the sediments. Worldwide, gas hydrate is estimated to hold about 1016 kg of organic carbon in the form of methane (Kvenvolden et al., 1993). Gas hydrate is one of the fossil fuel resources that is yet untapped, but may play a major role in meeting the energy challenge of this century. In June 2002, Westport Technology Center was requested by the Department of Energy (DOE) to prepare a ''Best Practices Manual on Gas Hydrate Coring, Handling and Analysis'' under Award No. DE-FC26-02NT41327. The scope of the task was specifically targeted for coring sediments with hydrates in Alaska, the Gulf of Mexico (GOM) and from the present Ocean Drilling Program (ODP) drillship. The specific subjects under this scope were defined in 3 stages as follows: Stage 1: Collect information on coring sediments with hydrates, core handling, core preservation, sample transportation, analysis of the core, and long term preservation. Stage 2: Provide copies of the first draft to a list of experts and stakeholders designated by DOE. Stage 3: Produce a second draft of the manual with benefit of input from external review for delivery. The manual provides an overview of existing information available in the published literature and reports on coring, analysis, preservation and transport of gas hydrates for laboratory analysis as of June 2003. The manual was delivered as draft version 3 to the DOE Project Manager for distribution in July 2003. This Final Report is provided for records purposes.

  16. Uranium droplet nuclear reactor core with MHD generator

    Science.gov (United States)

    Anghaie, Samim; Kumar, Ratan

    An innovative concept employing liquid uranium droplets as fuel in an ultrahigh-temperature vapor core reactor (UTVR) magnetohydrodynamic (MHD) generator power system for space power generation has been studied. Metallic vapor in superheated form acts as a working fluid for a closed-Rankine-type thermodynamic cycle. Usage of fuel and working fluid in this form assures certain advantages. The major technical issues emerging as a result involve a method for droplet generation, droplet transport in the reactor core, heat generation in the fuel and transport to the metallic vapor, and materials compatibility. A qualitative and quantitative attempt to resolve these issues has indicated the promise and tentative feasibility of the system.

  17. Gas core nuclear thermal rocket engine research and development in the former USSR

    Energy Technology Data Exchange (ETDEWEB)

    Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G. [eds.; Gurfink, M.M.

    1992-09-01

    Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept.

  18. Unified nuclear core activity map reconstruction using heterogeneous instruments with data assimilation

    CERN Document Server

    Bouriquet, Bertrand; Erhard, Patrick; Ponçot, Angélique

    2011-01-01

    Evaluating the neutronic state of the whole nuclear core is a very important topic that have strong implication for nuclear core management and for security monitoring. The core state is evaluated using measurements. Usually, part of the measurements are used, and only one kind of instruments are taken into account. However, the core state evaluation should be more accurate when more measurements are collected in the core. But using information from heterogeneous sources is at glance a difficult task. This difficulty can be overcome by Data Assimilation techniques. Such a method allows to combine in a coherent framework the information coming from model and the one coming from various type of observations. Beyond the inner advantage to use heterogeneous instruments, this leads to obtain a significant increasing of the quality of neutronic global state reconstruction with respect to individual use of measures. In order to present this approach, we will introduce here the basic principles of data assimilation f...

  19. Pinning down nuclear. To the core of the matter

    Energy Technology Data Exchange (ETDEWEB)

    Boeck, Helmut; Gerstmayr, Michael [Technische Univ., Vienna (Austria); International Atomic Energy Agency, Vienna (Austria); Radde, Eileen [Nuclear Engineering Seibersdorf GmbH (Austria); International Atomic Energy Agency, Vienna (Austria)

    2014-07-01

    The nuclear disaster in Fukushima shocked the world tremendously. The call to pull out of nuclear energy is getting louder - and more often than not by politicians trying to lure the favour of voters. Through the media there are half-truths and false information floating about the global consequences of the disaster and sensational prognoses for the future, all of which are in turn unsettling for the general public. Are the opposers to nuclear energy playing with the fear of the public or is the threat real? This book tells, in a captivating manner - authenticated with examples and incidents not known by many - what the threat for the area actually looks like. They confront the level of truth in the frightening scenarios and inform about the situation in case of emergency. Furthermore, they examine factors that preceded the disaster and broach the subject of the incredible hunger for energy, which dominates the world and continues to drive the commercial use of nuclear energy. Also the ghost of Chernobyl and its aftermath, which has been dismissed from our minds, is re-examined based on current knowledge. The book impresses with insider know-how, latest detailed knowledge, amazing facts and an entertaining narrative style.

  20. Nuclear Fusion in the Deuterated cores of inflated hot Jupiters

    CERN Document Server

    Ouyed, Rachid

    2015-01-01

    In Ouyed et al. (1998), Deuterium-Deuterium (DD) burning in the deep interior of giant planets (at the core-mantle interface) was proposed as a mechanism to explain their observed heat excess. An issue with such a mechanism is the extreme condition of high interior temperatures (~ 10^5 K) in a concentrated D layer needed to account for the excess heat. In this paper, we show that screened DD fusion in a deuterated core is a more plausible mechanism to explain the excess heat and observed inflated radii of some Jovian exoplanets ("hot Jupiters"). The screening alleviates the extreme temperature constraint and removes the requirement of a stratified D layer, so that DD-fusion is a significant internal energy source (~ 10^(25)-10^(27) erg/s) even within the expected range of core temperature (~ 10^4 K) and density of hot Jupiters. The mechanism is universal, long-lasting (Gigayears), and should be effective as long as the metallicity is not too high and the core has not been significantly eroded away already. Ap...

  1. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and thermal-hydraulic

  2. Evaluation of nuclear characteristics of DCA modification core for sub-critical measurement

    Energy Technology Data Exchange (ETDEWEB)

    Hazama, Taira [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-10-01

    Critical experiments were carried out on Deuterium Critical Assembly (DCA) modification core. DCA modification core has two regions, that is, test region and driver region. The test region consists of various types of fuel and moderator, while the driver region remains the same as the original DCA core (ATR simulated core). Critical characteristics were measured with various types of core patterns and were compared with calculated values based on SCALE code system. Monte Calro code KENO was found to be very accurate in the core analysis. The accuracy stays below 0.5 %dk/k in keff even if core configuration is extremely complicated. (author)

  3. Simulation of Thermopower Influence on Fuel Core of Power Rod in Nuclear Power Plant (NPP Active Zone

    Directory of Open Access Journals (Sweden)

    I. S. Kulikov

    2010-01-01

    Full Text Available The paper considers problems of modern methods for  calculation of designs and materials of nuclear power. A model of numerical analysis for stress-strain state of fuel pins in the NPP active zone is proposed in the paper. The paper contains simulation concerning a fuel core section of a nuclear reactor heat-generating element with subsequent solution of a temperature and thermoelastic problem in computer program complex FEA ANSYS Workbench 11.0. All the obtained results have passed through checking procedure.

  4. (129)I record of nuclear activities in marine sediment core from Jiaozhou Bay in China.

    Science.gov (United States)

    Fan, Yukun; Hou, Xiaolin; Zhou, Weijian; Liu, Guangshan

    2016-04-01

    Iodine-129 has been used as a powerful tool for environmental tracing of human nuclear activities. In this work, a sediment core collected from Jiaozhou Bay, the east coast of China, in 2002 was analyzed for (129)I to investigate the influence of human nuclear activities in this region. Significantly enhanced (129)I level was observed in upper 70 cm of the sediment core, with peak values in the layer corresponding to 1957, 1964, 1974, 1986, and after 1990. The sources of (129)I and corresponding transport processes in this region are discussed, including nuclear weapons testing at the Pacific Proving Grounds, global fallout from a large numbers of nuclear weapon tests in 1963, the climax of Chinese nuclear weapons testing in the early 1970s, the Chernobyl accident in 1986, and long-distance dispersion of European reprocessing derived (129)I. The very well (129)I records of different human nuclear activities in the sediment core illustrate the potential application of (129)I in constraining ages and sedimentation rates of the recent sediment. The releases of (129)I from the European nuclear fuel reprocessing plants at La Hague (France) and Sellafield (UK) were found to dominate the inventory of (129)I in the Chinese sediments after 1990, not only the directly atmospheric releases of these reprocessing plants, but also re-emission of marine discharged (129)I of these reprocessing plants in the highly contaminated European seas.

  5. CFD Analysis of Core Bypass Phenomena

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the sector grid can be set as a symmetry boundary

  6. CFD Analysis of Core Bypass Phenomena

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-11-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary

  7. Nuclear weapon detection categorization analysis

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    This statement of work is for the Proof of Concept for nuclear weapon categories utility in Arms control. The focus of the project will be to collect, analyze and correlate Intrinsic Radiation (INRAD) calculation results for the purpose of defining measurable signatures that differentiate categories of nuclear weapons. The project will support START III negotiations by identifying categories of nuclear weapons. The categories could be used to clarify sub-limits on the total number of nuclear weapons.

  8. Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

  9. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  10. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  11. Nuclear determination of saturation profiles in core plugs. Status report

    Energy Technology Data Exchange (ETDEWEB)

    Sletsgaard, J. [DTU, Inst. for Automation (Denmark)

    1996-01-01

    A method to determine liquid saturations in core plugs during flooding is of importance when the relative permeability and capillary pressure function are to be determined. This part of the EFP-93 project uses transmission of {gamma}-radiation to determine these saturations. In {gamma}-transmission measurements, the electron density of the given substance is measured. This is an advantage as compared to methods that use electric conductivity, since neither oil nor gas conducts electricity. At the moment a single {sup 137}Cs-source is used, but a theoretical investigation of whether it is possible to determine three saturations, using two radioactive sources with different {gamma}-energies, has been performed. Measurements were made on three core plugs. To make sure that the measurements could be reproduced, all the plugs had a point of reference, i.e. a mark so that it was possible to place the plug same way every time. Two computer programs for calculation of saturation and porosity and the experimental setup are listed. (EG).

  12. The open-cycle gas-core nuclear rocket engine - Some engineering considerations.

    Science.gov (United States)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyk, L. C.

    1971-01-01

    A preliminary design study of a conceptual 6000-MW open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 44,200 lb and a specific impulse of 4400 sec. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel) and the waste heat rejection system were considered conceptually and were sized.

  13. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  14. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  15. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    Science.gov (United States)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  16. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  17. New Nuclear Equation of State for Core-Collapse Supernovae with the Variational Method

    Directory of Open Access Journals (Sweden)

    Togashi H.

    2014-03-01

    Full Text Available We report the current status of our project to construct a new nuclear equation of state (EOS with the variational method for core-collapse supernova (SN simulations. Starting from the realistic nuclear Hamiltonian, the EOS for uniform nuclear matter is constructed with the cluster variational method: For non-uniform nuclear matter, the EOS is calculated with the Thomas-Fermi method. The obtained thermodynamic quantities of uniform matter are in good agreement with those with more sophisticated Fermi Hypernetted Chain variational calculations, and phase diagrams constructed so far are close to those of the Shen-EOS. The structure of neutron stars calculated with this EOS at zero temperature is consistent with recent observational data, and the maximum mass of the neutron star is slightly larger than that with the Shen-EOS. Using the present EOS of uniform nuclear matter, we also perform the 1D simulation of the core-collapse supernovae by a simplified prescription of adiabatic hydrodynamics. The stellar core with the present EOS is more compact than that with the Shen-EOS, and correspondingly, the explosion energy in this simulation with the present EOS is larger than that with the Shen-EOS.

  18. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

    2008-09-15

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

  19. Dynamical Behavior of Core 3 He Nuclear Reaction-Diffusion Systems and Sun's Gravitational Field

    Institute of Scientific and Technical Information of China (English)

    DU Jiulin; SHEN Hong

    2005-01-01

    The coupling of the sun's gravitational field with processes of diffusion and convection exerts a significant influence on the dynamical behavior of the core 3He nuclear reaction-diffusion system. Stability analyses of the system are made in this paper by using the theory of nonequilibrium dynamics. It is showed that, in the nuclear reaction regions extending from the center to about 0.38 times of the radius of the sun, the gravitational field enables the core 3He nuclear reaction-diffusion system to become unstable and, after the instability, new states to appear in the system have characteristic of time oscillation. This may change the production rates of both 7Be and 8B neutrinos.

  20. An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

    Science.gov (United States)

    Clark, John S.; Walton, James T.; Mcguire, Melissa L.

    1992-01-01

    Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.

  1. Numerical simulation of a Hypothetical Core Disruptive Accident in a small-scale model of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. E-mail: robbe@aquilon.cea.frmfrobbe@cea.fr; Lepareux, M.; Treille, E.; Cariou, Y

    2003-08-01

    In the case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between molten fuel and liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant. The MARA 10 experimental test simulates a HCDA in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge. This paper presents a numerical simulation of the test with the EUROPLEXUS code and an analysis of the computed results. In particular, the evolution of the fluid flows and the deformations of the internal and external structures are analysed in detail. Finally, the current computed results are compared with the experimental ones and with previous numerical results computed with the SIRIUS and CASTEM-PLEXUS codes.

  2. Preliminary Nuclear Analysis for the HANARO Fuel Element with Burnable Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, So Young; In, Won Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Burnable absorber is used for reducing reactivity swing and power peaking in high performance research reactors. Development of the HANARO fuel element with burnable absorber was started in the U-Mo fuel development program at HANARO, but detailed full core analysis was not performed because the current HANARO fuel management system is uncertain to analysis the HANARO core with burnable absorber. A sophisticated reactor physics system is required to analysis the core. The McCARD code was selected and the detailed McCARD core models, in which the basic HANARO core model was developed by one of the McCARD developers, are used in this study. The development of nuclear fuel requires a long time and correct developing direction especially by the nuclear analysis. This paper presents a preliminary nuclear analysis to promote the fuel development. Based on the developed fuel, the further nuclear analysis will improve reactor performance and safety. Basic nuclear analysis for the HANARO and the AHR were performed for getting the proper fuel elements with burnable absorber. Addition of 0.3 - 0.4% Cd to the fuel meat is promising for the current HANARO fuel element. Small addition of burnable absorber may not change any fuel characteristics of the HANARO fuel element, but various basic tests and irradiation tests at the HANARO core are required.

  3. CORE DESIGNS OF ABWR FOR PROPOSED OF THE FIRST NUCLEAR POWER PLANT IN INDONESIA

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2015-04-01

    Full Text Available Indonesia as an archipelago has been experiencing high growth industry and energy demand due to high population growth, dynamic economic activities. The total population is around 230 million people and 75 % to the total population is living in Java. The introduction of Nuclear Power Plant on Java Bali electricity grid will be possible in 2022 for 2 GWe, using proven technology reactor like ABWR or others light water reactor with nominal power 1000 MWe. In this case, the rated thermal power for the equilibrium cycles is 3926 MWt, the cycle length is 18 month and overall capacity factor is 87 %. The designs were performed for an 872-fuel bundles ABWR core using GE-11 fuel type in an 9×9 fuel rod arrays with 2 Large Central Water Rods (LCWR. The calculations were divided into two steps; the first is to generate bundle library and the other is to make the thermal and reactivity limits satisfied for the core designs. Toshiba General Electric Bundle lattice Analysis (TGBLA and PANACEA computer codes were used as designs tools. TGBLA is a General Electric proprietary computer code which is used to generate bundle lattice library for fuel designs. PANACEA is General Electric proprietary computer code which is used as thermal hydraulic and neutronic coupled BWR core simulator. This result of core designs describes reactivity and thermal margins i.e.; Maximum Linear Heat Generation rate (MLHGR is lower than 14.4 kW/ft, Minimum Critical Power Ratio (MCPR is upper than 1.25, Hot Excess Reactivity (HOTXS is upper than 1 %Dk at BOC and 0.8 %Dk at 200 MWD/ST and Cold Shutdown Margin Reactivity (CSDM is upper than 1 %Dk. It is concluded that the equilibrium core design using GE-11 fuel bundle type satisfies the core design objectives for the proposed of the firs Indonesia ABWR Nuclear Power Plant. Keywords: The first NPP in Indonesia, ABWR-1000 MWe, and core designs.   Indonesia adalah sebagai negara kepulauan yang laju pertumbuhan industri, energi, penduduk

  4. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  5. Chernobyl nuclear accident revealed from the 7010 m Muztagata ice core record

    Institute of Scientific and Technical Information of China (English)

    TIAN LiDe; YAO TanDong; WU GuangJian; LI Zhen; XU BaiQing; LI YueFang

    2007-01-01

    The total activity variation with depth from a 41.6 m Muztagata ice core drilled at 7010 m,recorded not only the 1963 radioactive layer due to the thermonuclear test,but also clearly the radioactive peak released by the Chernobyl accident in 1986.This finding indicates that the Chernobyl nuclear accident was clearly recorded in alpine glaciers in the Pamirs of west China,and the layer can be potentially used for ice core dating in other high alpine glaciers in the surrounding regions.

  6. Comparative assessment of out-of-core nuclear thermionic power systems

    Science.gov (United States)

    Estabrook, W. C.; Koenig, D. R.; Prickett, W. Z.

    1975-01-01

    The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds.

  7. An assessment of coupling algorithms for nuclear reactor core physics simulations

    Science.gov (United States)

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C. T.; Evans, Thomas; Philip, Bobby

    2016-04-01

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss-Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton-Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  8. Nuclear data evaluation and group constant generation for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Doh; Kil, Chung Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-12-01

    Data testing of ENDF/B-VI.2 was performed and ACE-format continuous point-wise cross section library from ENDF/B-VI.2 for MCNP was validated through CSEWG benchmark and power plant mockup experiments. The calculated k-effective of ORNL-1, -2, -3, -4 and -10 with ENDF/B-VI are low by about 0.5% but those of L-7, -8, -9, -10 and -11 show good agreement with experiments. Overall results for uranium core with ENDF/B-VI is low in critically than with ENDF/B-V. The calculated results with ENDF/B-VI for PNL-6 {approx} 12 of plutonium core and PNL-30 {approx} 35 of mixed oxide core show good agreement with the experiments. The results of critically calculation for fast core benchmark do not show large difference between ENDF/B-VI and -V. But the calculated results of reaction rate ratio with ENDF/B-VI are improved, compared with ENDF/B-V. The calculated power distribution for VENUS PWR mockup core and typical BWR core of GE with both of ENDF/B-VI and -V agree well with measured values. From the above results, newly generated MCNP library from ENDF/B-VI is useful for nuclear and shielding design and analysis. 5 figs, 13 tabs, 11 refs. (Author).

  9. Design and Performance of South Ukraine Nuclear Power Plant Mixed Cores

    Energy Technology Data Exchange (ETDEWEB)

    Abdullayev, A. M.; Baydulin, V.; Zhukov, A. I.; Latorre, Richard

    2011-09-24

    In 2010, 42 Westinghouse fuel assemblies (WFAs) were loaded into the core of South Ukraine Nuclear Power Plant (SUNPP) Unit 3 after four successful cycles with 6 Westinghouse Lead Test Assemblies. The scope of safety substantiating documents required for the regulatory approval of this mixed core was extended considerably, particularly with development and implementation of new methodologies and 3-D kinetic codes. Additional verification for all employed codes was also performed. Despite the inherent hydraulic non-uniformity of a mixed core, it was possible to demonstrate that all design and operating restrictions for three different types of fuel (TVS-M, TVSA and WFA) loaded in the core were conservatively met. This paper provides the main results from the first year of operation of the core loaded with 42 WFAs, the predicted parameters for the transition and equilibrium cycles with WFAs, comparisons of predicted versus measured core parameters, as well as the acceptable margin evaluation results for reactivity accidents using the 3-D kinetic codes. To date WFA design parameters have been confirmed by operation experience.

  10. Autoclave nuclear criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)

    1991-12-31

    Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

  11. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    Science.gov (United States)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

  12. Neutrino-pair emission from nuclear de-excitation in core-collapse supernova simulations

    CERN Document Server

    Fischer, Tobias; Martinez-Pinedo, Gabriel

    2013-01-01

    We study the impact of neutrino-pair production from the de-excitation of highly excited heavy nuclei on core-collapse supernova simulations, following the evolution up to several 100 ms after core bounce. Our study is based on the AGILE-Boltztran supernova code, which features general relativistic radiation hydrodynamics and accurate three-flavor Boltzmann neutrino transport in spherical symmetry. In our simulations the nuclear de-excitation process is described in two different ways. At first we follow the approach proposed by Fuller and Meyer [Astrophys. J. 376,701 (1991)], which is based on strength functions derived in the framework of the nuclear Fermi-gas model of non-interacting nucleons. Secondly, we parametrize the allowed and forbidden strength distributions in accordance with measurements for selected nuclear ground states. We determine the de-excitation strength by applying the Brink hypothesis and detailed balance. For both approaches, we find that nuclear de-excitation has no effect on the supe...

  13. The Sensitivity of Core-Collapse Supernovae to Nuclear Electron Capture

    CERN Document Server

    Sullivan, Chris; Zegers, Remco G T; Grubb, Thomas; Austin, Sam M

    2015-01-01

    A weak-rate library aimed at investigating the sensitivity of astrophysical environments to variations of electron-capture rates on medium-heavy nuclei has been developed. With this library, the sensitivity of the core-collapse and early post-bounce phases of core-collapse supernovae to nuclear electron-capture is examined by systematically and statistically varying electron-capture rates of individual nuclei. The rates are adjusted by factors consistent with uncertainties indicated by comparing theoretical rates to those deduced from charge-exchange and $\\beta$-decay measurements. To ensure a model independent assessment, sensitivity studies across a comprehensive set of progenitors and equations of state are performed. In our systematic study, we find a +16/-4 % range in the mass of the inner-core at the time of shock formation and a $\\pm$20% range of peak {\

  14. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  15. Core Backbone Convergence Mechanisms and Microloops Analysis

    Directory of Open Access Journals (Sweden)

    Abdelali Ala

    2012-07-01

    Full Text Available In this article we study approaches that can be used to minimise the convergence time, we also make a focus on microloops phenomenon, analysis and means to mitigate them. The convergence time reflects the time required by a network to react to a failure of a link or a router failure itself. When all nodes (routers have updated their respective routing and forwarding databases, we can say the network has converged. This study will help in building real-time and resilient network infrastructure, the goal is to make any evenement in the core network, as transparent as possible to any sensitive and real-time flows. This study is also, a deepening of earlier works presented in [10] and [11].

  16. Multi-Core Processor Memory Contention Benchmark Analysis Case Study

    Science.gov (United States)

    Simon, Tyler; McGalliard, James

    2009-01-01

    Multi-core processors dominate current mainframe, server, and high performance computing (HPC) systems. This paper provides synthetic kernel and natural benchmark results from an HPC system at the NASA Goddard Space Flight Center that illustrate the performance impacts of multi-core (dual- and quad-core) vs. single core processor systems. Analysis of processor design, application source code, and synthetic and natural test results all indicate that multi-core processors can suffer from significant memory subsystem contention compared to similar single-core processors.

  17. Nuclear charge and neutron radii and nuclear matter: trend analysis

    CERN Document Server

    Reinhard, P -G

    2016-01-01

    Radii of charge and neutron distributions are fundamental nuclear properties. They depend on both nuclear interaction parameters related to the equation of state of infinite nuclear matter and on quantal shell effects, which are strongly impacted by the presence of nuclear surface. In this work, by studying the dependence of charge and neutron radii, and neutron skin, on nuclear matter parameters, we assess different mechanisms that drive nuclear sizes. We apply nuclear density functional theory using a family of Skyrme functionals obtained by means of different optimization protocols targeting specific nuclear properties. By performing the Monte-Carlo sampling of reasonable functionals around the optimal parametrization, we study correlations between nuclear matter paramaters and observables characterizing charge and neutron distributions. We demonstrate the existence of the strong converse relation between the nuclear charge radii and the saturation density of symmetric nuclear matter and also between the n...

  18. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  19. A nuclear heuristic for application to metaheuristics in-core fuel management optimization

    Energy Technology Data Exchange (ETDEWEB)

    Meneses, Anderson Alvarenga de Moura, E-mail: ameneses@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program; Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Gambardella, Luca Maria, E-mail: luca@idsia.c [Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Schirru, Roberto, E-mail: schirru@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program

    2009-07-01

    The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

  20. Core safety of Indian nuclear power plants (NPPs) under extreme conditions

    Indian Academy of Sciences (India)

    J B Joshi; A K Nayak; M Singhal; D Mukhopadhaya

    2013-10-01

    Nuclear power is currently the fourth largest source of electricity production in India after thermal, hydro and renewable sources of electricity. Currently, India has 20 nuclear reactors in operation and seven other reactors are under construction. Most of these reactors are indigenously designed and built Heavy Water Reactors. In addition, a 300 MWe Advanced Heavy Water Reactor has already been designed and in the process of deployment in near future for demonstration of power production from Thorium apart from enhanced safety features by passive means. India has ambitious plans to enhance the share of electricity production from nuclear. The recent Fukushima accident has raised concerns of safety of Nuclear Power Plants worldwide. The Fukushima accident was caused by extreme events, i.e., large earthquake followed by gigantic Tsunami which are not expected to hit India’s coast considering the geography of India and historical records. Nevertheless, systematic investigations have been conducted by nuclear scientists in India to evaluate the safety of the current Nuclear Power Plants in case of occurrence of such extreme events in any nuclear site. This paper gives a brief outline of the safety features of Indian Heavy Water Reactors for prevention and mitigation of such extreme events. The probabilistic safety analysis revealed that the risk from Indian Heavy Water Reactors are negligibly small.

  1. A global DGLAP analysis of nuclear PDFs

    Science.gov (United States)

    Eskola, K. J.; Kolhinen, V. J.; Paukkunen, H.; Salgado, C. A.

    2008-05-01

    In this talk, we shortly report results from our recent global DGLAP analysis of nuclear parton distributions. This is an extension of our former EKS98-analysis improved with an automated χ2 minimization procedure and uncertainty estimates. Although our new analysis show no significant deviation from EKS98, a sign of a significantly stronger gluon shadowing could be seen in the RHIC BRAHMS data.

  2. Nuclear structure and the fate of core collapse (Type II) supernova

    Energy Technology Data Exchange (ETDEWEB)

    Gai, Moshe [LNS at Avery Point, University of Connecticut, Groton, CT 06340-6097 (United States); Wright Lab, Dept. of Physics, Yale University, New Haven, CT 06520-8124 (United States)

    2014-08-15

    For a long time Gerry Brown and his collaborator Hans Bethe considered the question of the final fate of a core collapse (Type II) supernova. Recalling ideas from nuclear structure on Kaon condensate and a soft equation of state of the dense nuclear matter they concluded that progenitor stars with mass as low as 17–18M{sub ⊙} (including supernova 1987A) could collapse to a small mass black hole with a mass just beyond 1.5M{sub ⊙}, the upper bound they derive for a neutron star. We discuss another nuclear structure effect that determines the carbon to oxygen ratio (C/O) at the end of helium burning. This ratio also determines the fate of a Type II supernova with a carbon rich progenitor star producing a neutron star and oxygen rich collapsing to a black hole. While the C/O ratio is one of the most important nuclear inputs to stellar evolution it is still not known with sufficient accuracy. We discuss future efforts to measure with gamma-beam and TPC detector of the {sup 12}C(α,γ){sup 16}O reaction that determines the C/O ratio in stellar helium burning.

  3. Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

    Science.gov (United States)

    Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

    2017-01-01

    The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

  4. Nuclear Structure and the Fate of Core Collapse (Type II) Supernova

    CERN Document Server

    Gai, Moshe

    2014-01-01

    For a long time Gerry Brown and his collaborator Hans Bethe considered the question of the final fate of a core collapse (Type II) supernova. Recalling ideas from nuclear structure on Kaon condensate and a soft equation of state of the dense nuclear matter they concluded that progenitor stars with mass as low a 17-18M$_\\odot$ (including supernova 1987A) could collapse to a small mass black hole with a mass just beyond 1.5M$_\\odot$, the upper bound they derive for a neutron star. We discuss another nuclear structure effect that determines the carbon to oxygen ratio (C/O) at the end of helium burning. This ratio also determines the fate of a Type II supernova with a carbon rich progenitor star producing a neutron star and oxygen rich collapsing to a black hole. While the C/O ratio is one of the most important nuclear input to stellar evolution it is still not known with sufficient accuracy. We discuss future efforts to measure with gamma-beam and TPC detector the 12C(a,g)16O reaction that determines the C/O rat...

  5. The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems

    Energy Technology Data Exchange (ETDEWEB)

    Gallup, D.R.

    1990-03-01

    In this document, masses of the STAR-C power system and an optimized out-of-core thermionic reactor (OTR) power system versus power level are investigated. The impacts of key system parameters on system performance are also addressed. The STAR-C is mass competitive below about 15 kWe, but at higher power levels the scalability is relatively poor. An optimized OR is the least massive space nuclear power system below 25 kWe, and scales well to 50 kWe. The system parameters that have a significant impact on the scalability of the STAR-C are core thermal flux, thermionic converter efficiency, and core length to diameter ratio. The emissivity of the core surface is shown to be a relatively unimportant parameter. For an optimized OR power system, the most significant system parameter is the maximum allowable fuel temperature. It is also shown that if advanced radiation-hardened electronics are used in the satellite payload, a very large mass savings is realized. 10 refs., 23 figs., 7 tabs.

  6. SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

    Directory of Open Access Journals (Sweden)

    WOLFGANG HARTMANN

    2013-10-01

    Full Text Available This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

  7. Molecular forensic science analysis of nuclear materials

    Science.gov (United States)

    Reilly, Dallas David

    Concerns over the proliferation and instances of nuclear material in the environment have increased interest in the expansion of nuclear forensics analysis and attribution programs. A new related field, molecular forensic science (MFS) has helped meet this expansion by applying common scientific analyses to nuclear forensics scenarios. In this work, MFS was applied to three scenarios related to nuclear forensics analysis. In the first, uranium dioxide was synthesized and aged at four sets of static environmental conditions and studied for changes in chemical speciation. The second highlighted the importance of bulk versus particle characterizations by analyzing a heterogeneous industrially prepared sample with similar techniques. In the third, mixed uranium/plutonium hot particles were collected from the McGuire Air Force Base BOMARC Site and analyzed for chemical speciation and elemental surface composition. This work has identified new signatures and has indicated unexpected chemical behavior under various conditions. These findings have lead to an expansion of basic actinide understanding, proof of MFS as a tool for nuclear forensic science, and new areas for expansion in these fields.

  8. Preliminaries on core image analysis using fault drilling samples; Core image kaiseki kotohajime (danso kussaku core kaisekirei)

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, T.; Ito, H. [Geological Survey of Japan, Tsukuba (Japan)

    1996-05-01

    This paper introduces examples of image data analysis on fault drilling samples. The paper describes the following matters: core samples used in the analysis are those obtained from wells drilled piercing the Nojima fault which has moved in the Hygoken-Nanbu Earthquake; the CORESCAN system made by DMT Corporation, Germany, used in acquiring the image data consists of a CCD camera, a light source and core rotation mechanism, and a personal computer, its resolution being about 5 pixels/mm in both axial and circumferential directions, and 24-bit full color; with respect to the opening fractures in core samples collected by using a constant azimuth coring, it was possible to derive values of the opening width, inclination angle, and travel from the image data by using a commercially available software for the personal computer; and comparison of this core image with the BHTV record and the hydrophone VSP record (travel and inclination obtained from the BHTV record agree well with those obtained from the core image). 4 refs., 4 figs.

  9. Hybrid Analysis of Engine Core Noise

    Science.gov (United States)

    O'Brien, Jeffrey; Kim, Jeonglae; Ihme, Matthias

    2015-11-01

    Core noise, or the noise generated within an aircraft engine, is becoming an increasing concern for the aviation industry as other noise sources are progressively reduced. The prediction of core noise generation and propagation is especially challenging for computationalists since it involves extensive multiphysics including chemical reaction and moving blades in addition to the aerothermochemical effects of heated jets. In this work, a representative engine flow path is constructed using experimentally verified geometries to simulate the physics of core noise. A combustor, single-stage turbine, nozzle and jet are modeled in separate calculations using appropriate high fidelity techniques including LES, actuator disk theory and Ffowcs-Williams Hawkings surfaces. A one way coupling procedure is developed for passing fluctuations downstream through the flowpath. This method effectively isolates the core noise from other acoustic sources, enables straightforward study of the interaction between core noise and jet exhaust, and allows for simple distinction between direct and indirect noise. The impact of core noise on the farfield jet acoustics is studied extensively and the relative efficiency of different disturbance types and shapes is examined in detail.

  10. Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    Science.gov (United States)

    Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

    2017-09-01

    The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

  11. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Energy Technology Data Exchange (ETDEWEB)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki [Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan) and Advanced Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Numazu College of Technology, Ooka 3600, Numazu, Shizuoka 410-8501 (Japan); Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan)

    2012-11-12

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to {approx} 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  12. Possible generation of heat from nuclear fusion in Earth’s inner core

    Science.gov (United States)

    Fukuhara, Mikio

    2016-11-01

    The cause and source of the heat released from Earth’s interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2D + 2D + 2D → 21H + 4He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 1012 J/m3, based on the assumption that Earth’s primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth’s interior to the universe, and pass through Earth, respectively.

  13. Possible generation of heat from nuclear fusion in Earth's inner core.

    Science.gov (United States)

    Fukuhara, Mikio

    2016-11-23

    The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: (2)D + (2)D + (2)D → 2(1)H + (4)He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10(12) J/m(3), based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

  14. Economic analysis of nuclear power generation

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Dong; Choi, Young Myung; Kim, Hwa Sup; Lee, Man Ki; Moon, Kee Hwan; Kim, Seung Su; Chae, Kyu Nam

    1996-12-01

    The major contents in this study are as follows : (1) Efforts are made to examine the role of nuclear energy considering environmental regulation. An econometric model for energy demand and supply including carbon tax imposition is established. (2) Analysis for the learning effect of nuclear power plant operation is performed. The study is focused to measure the effect of technology homogeneity on the operation performance. (3) A preliminary capital cost of the KALIMER is estimated by using cost computer program, which is developed in this study. (author). 36 refs.,46 tabs., 15 figs.

  15. Advanced Materials and Solids Analysis Research Core (AMSARC)

    Science.gov (United States)

    The Advanced Materials and Solids Analysis Research Core (AMSARC), centered at the U.S. Environmental Protection Agency's (EPA) Andrew W. Breidenbach Environmental Research Center in Cincinnati, Ohio, is the foundation for the Agency's solids and surfaces analysis capabilities. ...

  16. Advanced Materials and Solids Analysis Research Core (AMSARC)

    Science.gov (United States)

    The Advanced Materials and Solids Analysis Research Core (AMSARC), centered at the U.S. Environmental Protection Agency's (EPA) Andrew W. Breidenbach Environmental Research Center in Cincinnati, Ohio, is the foundation for the Agency's solids and surfaces analysis capabilities. ...

  17. Nuclear radiation analysis for TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Ku, L.P.

    1980-09-01

    A multi-dimensional analysis of the radiation problems was performed for TFTR using the system established at PPPL. Simple, clean geometries were first used to establish the expected reference level. Different calculational models were compared. The characteristics of streaming and activation were then studied. It is shown that the objectives set forth for the TFTR shielding design are not entirely satisfied, based on our calculation. Using the results of this study as a baseline, a review of the shielding for TFTR and its upgrade, TFM, has been initiated with the objective of a shielding design modified to satisfy both modes of operation.

  18. Molecular double core-hole electron spectroscopy for chemical analysis

    CERN Document Server

    Tashiro, Motomichi; Fukuzawa, Hironobu; Ueda, Kiyoshi; Buth, Christian; Kryzhevoi, Nikolai V; Cederbaum, Lorenz S

    2010-01-01

    We explore the potential of double core hole electron spectroscopy for chemical analysis in terms of x-ray two-photon photoelectron spectroscopy (XTPPS). The creation of deep single and double core vacancies induces significant reorganization of valence electrons. The corresponding relaxation energies and the interatomic relaxation energies are evaluated by CASSCF calculations. We propose a method how to experimentally extract these quantities by the measurement of single and double core-hole ionization potentials (IPs and DIPs). The influence of the chemical environment on these DIPs is also discussed for states with two holes at the same atomic site and states with two holes at two different atomic sites. Electron density difference between the ground and double core-hole states clearly shows the relaxations accompanying the double core-hole ionization. The effect is also compared with the sensitivity of single core hole ionization potentials (IPs) arising in single core hole electron spectroscopy. We have ...

  19. Quantitative Analysis in Nuclear Medicine Imaging

    CERN Document Server

    2006-01-01

    This book provides a review of image analysis techniques as they are applied in the field of diagnostic and therapeutic nuclear medicine. Driven in part by the remarkable increase in computing power and its ready and inexpensive availability, this is a relatively new yet rapidly expanding field. Likewise, although the use of radionuclides for diagnosis and therapy has origins dating back almost to the discovery of natural radioactivity itself, radionuclide therapy and, in particular, targeted radionuclide therapy has only recently emerged as a promising approach for therapy of cancer and, to a lesser extent, other diseases. As effort has, therefore, been made to place the reviews provided in this book in a broader context. The effort to do this is reflected by the inclusion of introductory chapters that address basic principles of nuclear medicine imaging, followed by overview of issues that are closely related to quantitative nuclear imaging and its potential role in diagnostic and therapeutic applications. ...

  20. Multimegawatt nuclear electric propulsion with gaseous and vapor core reactors with MHD

    Science.gov (United States)

    Knight, Travis; Anghaie, Samim; Smith, Blair; Houts, Michael

    2001-02-01

    This study investigated the development of a system concept for space power generation and nuclear electric propulsion based on a fissioning plasma core reactor (FPCR) with magnetohydrodynamic (MHD) power conversion system, coupled to a magnetoplasmadynamic (MPD) thruster. The FPCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UF4) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. Candidate working fluids include K, Li, Na, KF, LiF, NaF, etc. The system features core outlet temperatures of 3000 to 4000 K at pressures of about 1 to 10 MPa, MHD temperatures of 2000 to 3000 K, and radiator temperatures of 1200 to 2000 K. This combination of parameters offers the potential for low total system specific mass in the range of 0.4 to 0.6 kg/kWe. The MHD output could be coupled with minimal power conditioning to the variable specific impulse magnetoplasma rocket (VASIMR), MPD thrusters or other types of thruster for producing thrust at very high specific impulse (Isp=1500 to 10,000 s). .

  1. Nuclear inputs of key iron isotopes for core-collapse modeling and simulation

    CERN Document Server

    Nabi, Jameel-Un

    2014-01-01

    From the modeling and simulation results of presupernova evolution of massive stars, it was found that isotopes of iron, $^{54,55,56}$Fe, play a significant role inside the stellar cores, primarily decreasing the electron-to-baryon ratio ($Y_{e}$) mainly via electron capture processes thereby reducing the pressure support. The neutrinos produced, as a result of these capture processes, are transparent to the stellar matter and assist in cooling the core thereby reducing the entropy. The structure of the presupernova star is altered both by the changes in $Y_{e}$ and the entropy of the core material. Here we present the microscopic calculation of Gamow-Teller strength distributions for isotopes of iron. The calculation is also compared with other theoretical models and experimental data. Presented also are stellar electron capture rates and associated neutrino cooling rates, due to isotopes of iron, in a form suitable for simulation and modeling codes. It is hoped that the nuclear inputs presented here should ...

  2. 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)

  3. Core Competence Analysis--Toyota Production System

    Institute of Scientific and Technical Information of China (English)

    钱璐宜

    2013-01-01

      Core competencies are the wel spring of new business development. It is the sharpest sword to penetrate the mature market, hold and enlarge the existing share. Toyota makes wel use of its TPS and form its own style which other car manufacturers hard to imitate.In contrast,the Chinese company---FAW only imitating the superficial aspects from Toyota and ignoring its own problems in manufacture line.

  4. NEW SOIL VOC SAMPLERS: EN CORE AND ACCU CORE SAMPLING/STORAGE DEVICES FOR VOC ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Susan S. Sorini; John F. Schabron; Joseph F. Rovani Jr

    2006-06-01

    Soil sampling and storage practices for volatile organic analysis must be designed to minimize loss of volatile organic compounds (VOCs) from samples. The En Core{reg_sign} sampler is designed to collect and store soil samples in a manner that minimizes loss of contaminants due to volatilization and/or biodegradation. An ASTM International (ASTM) standard practice, D 6418, Standard Practice for Using the Disposable En Core Sampler for Sampling and Storing Soil for Volatile Organic Analysis, describes use of the En Core sampler to collect and store a soil sample of approximately 5 grams or 25 grams for volatile organic analysis and specifies sample storage in the En Core sampler at 4 {+-} 2 C for up to 48 hours; -7 to -21 C for up to 14 days; or 4 {+-} 2 C for up to 48 hours followed by storage at -7 to -21 C for up to five days. This report discusses activities performed during the past year to promote and continue acceptance of the En Core samplers based on their performance to store soil samples for VOC analysis. The En Core sampler is designed to collect soil samples for VOC analysis at the soil surface. To date, a sampling tool for collecting and storing subsurface soil samples for VOC analysis is not available. Development of a subsurface VOC sampling/storage device was initiated in 1999. This device, which is called the Accu Core{trademark} sampler, is designed so that a soil sample can be collected below the surface using a dual-tube penetrometer and transported to the laboratory for analysis in the same container. Laboratory testing of the current Accu Core design shows that the device holds low-level concentrations of VOCs in soil samples during 48-hour storage at 4 {+-} 2 C and that the device is ready for field evaluation to generate additional performance data. This report discusses a field validation exercise that was attempted in Pennsylvania in 2004 and activities being performed to plan and conduct a field validation study in 2006. A draft ASTM

  5. Design Optimization of Nuclear Vapor Thermal Rocket Core - A Thermo-Mechanical Study

    Science.gov (United States)

    Keshavmurthy, Shyam P.; Watanabe, Yoichi; Dugan, Edward T.; Diaz, Nils J.

    1994-07-01

    Fuel structural materials for the Nuclear Vapor Thermal Rocket (NVTR) are exposed to very high temperature vapor fuel in the fuel channel and to high temperature but cooler propellant in the coolant channel. This temperature difference leads to thermal stress in the fuel element. There is also a mismatch in the value of coefficients of thermal expansion between the fuel element material and the coating material that could lead to failure of the coating. The stress in the coating and the fuel element material is dependent on the power density of the core and also on the arrangement of fuel and coolant channels. In order to achieve higher power density, the fuel element design has to be optimized to yield lower stress. Analytical studies found that carbon/carbon composite hexagonal fuel elements employing a square lattice arrangement of multiple UF4 fuel and hydrogen coolant channels yield maximum stress intensities well below fuel element materials stress limit.

  6. Cost-based optimization of a nuclear reactor core design: a preliminary model

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F.; Alves Filho, Hermes [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Inst. Politecnico. Dept. de Modelagem Computacional]. E-mails: wfsacco@iprj.uerj.br; halves@iprj.uerj.br; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil). Div. de Reatores]. E-mail: cmnap@ien.gov.br

    2007-07-01

    A new formulation of a nuclear core design optimization problem is introduced in this article. Originally, the optimization problem consisted in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the radial power peaking factor in a three-enrichment zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. Here, we address the same problem using the minimization of the fuel and cladding materials costs as the objective function, and the radial power peaking factor as an operational constraint. This cost-based optimization problem is attacked by two metaheuristics, the standard genetic algorithm (SGA), and a recently introduced Metropolis algorithm called the Particle Collision Algorithm (PCA). The two algorithms are submitted to the same computational effort and their results are compared. As the formulation presented is preliminary, more elaborate models are also discussed (author)

  7. Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

    Energy Technology Data Exchange (ETDEWEB)

    Wan, Chenghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Cao, Liangzhi, E-mail: caolz@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Shen, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2017-04-15

    Highlights: • Uncertainty analysis has been completed based on the “two-step” scheme. • Uncertainty analysis has been performed to BEAVRS at HZP. • For lattice calculations, the few-group constant’s uncertainty was quantified. • For core simulation, uncertainties of k{sub eff} and power distributions were quantified. - Abstract: Based on the “two-step” scheme for the reactor-physics calculations, the capability of uncertainty analysis for the core simulations has been implemented in the UNICORN code, an in-house code for the sensitivity and uncertainty analysis of the reactor-physics calculations. Applying the statistical sampling method, the nuclear-data uncertainties can be propagated to the important predictions of the core simulations. The uncertainties of the few-group constants introduced by the uncertainties of the multigroup microscopic cross sections are quantified first for the lattice calculations; the uncertainties of the few-group constants are then propagated to the core multiplication factor and core power distributions for the core simulations. Up to now, our in-house lattice code NECP-CACTI and the neutron-diffusion solver NECP-VIOLET have been implemented in UNICORN for the steady-state core simulations based on the “two-step” scheme. With NECP-CACTI and NECP-VIOLET, the modeling and simulation of the steady-state BEAVRS benchmark problem at the HZP conditions was performed, and the results were compared with those obtained by CASMO-4E. Based on the modeling and simulation, the UNICORN code has been applied to perform the uncertainty analysis for BAEVRS at HZP. The uncertainty results of the eigenvalues and two-group constants for the lattice calculations and the multiplication factor and the power distributions for the steady-state core simulations are obtained and analyzed in detail.

  8. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  9. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  10. Assessment of water hammer effects on boiling water nuclear reactor core dynamics

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2007-01-01

    Full Text Available Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

  11. Assessing the feasibility and consequences of nuclear georeactors in the Earths core mantle boundary

    CERN Document Server

    De Meijer, R J

    2008-01-01

    We assess the likelihood and geochemical consequences of the presence of nuclear georeactors in the core mantle boundary region (CMB) between Earths silicate mantle and metallic core. Current geochemical models for the Earths interior predict that U and Th in the CMB are concentrated exclusively in the mineral calcium silicate perovskite (CaPv), leading to predicted concentration levels of approximately 12 ppm combined U and Th, 4.5 Ga ago if CaPv is distributed evenly throughout the CMB. Assuming a similar behaviour for primordial 244Pu provides a considerable flux of neutrons from spontaneous fission. We show that an additional concentration factor of only an order of magnitude is required to both ignite and maintain self sustaining georeactors based on fast fission. Continuously operating georeactors with a power of 5 TW can explain the observed isotopic compositions of helium and xenon in the Earths mantle. Our hypothesis requires the presence of elevated concentrations of U and Th in the CMB, and is amen...

  12. Exergoeconomic analysis of a nuclear power plant

    Science.gov (United States)

    Moreno, Roman Miguel

    Exergoeconomic analysis of a nuclear power plant is a focus of this dissertation. Specifically, the performance of the Palo Verde Nuclear Power Plant in Arizona is examined. The analysis combines thermodynamic second law exergy analysis with economics in order to assign costs to the loss and destruction of exergy. This work was done entirely with an interacting spreadsheets notebook. The procedures are to first determine conventional energy flow, where the thermodynamic stream state points are calculated automatically. Exergy flow is then evaluated along with destruction and losses. The capital cost and fixed investment rate used for the economics do not apply specifically to the Palo Verde Plant. Exergy costing is done next involving the solution of about 90 equations by matrix inversion. Finally, the analysis assigns cost to the exergy destruction and losses in each component. In this work, the cost of electricity (exergy), including capital cost, leaving the generator came to 38,400 /hr. The major exergy destruction occurs in the reactor where fission energy transfer is limited by the maxiμm permissible clad temperature. Exergy destruction costs were: reactor--18,207 hr, the low pressure turbine-2,000 /hr, the condenser--1,700 hr, the steam generator-1,200 $/hr. The inclusion of capital cost and O&M are important in new system design assessments. When investigating operational performance, however, these are sunk costs; only fuel cost needs to be considered. The application of a case study is included based on a real modification instituted at Palo Verde to reduce corrosion steam generator problems; the pressure in the steam generator was reduced from 1072 to 980 psi. Exergy destruction costs increased in the low pressure turbine and in the steam generator, but decreased in the reactor vessel and the condenser. The dissertation demonstrates the procedures and tools required for exergoeconomic analysis whether in the evaluation of a new nuclear reactor system

  13. Trends in nuclear systems design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G. (Commission of the European Communities, Joint Research Centre/Ispra Establishment, Inst. for Systems Engineering and Design (Italy))

    1991-12-01

    The papers presented at the poster session 'Nuclear Systems Design and Analyses' are reviewed and trends discussed. The attention is focussed on the following areas: (i) Reactor Studies and (ii) Neutronic Tests and Analysis. Present studies on D-T power reactor conceptual designs should be encouraged. Recent conceptual studies on D-{sup 3}He systems seem adequate to identify the critical issues. The economic feasibility of {sup 3}He supply from the moon should be assessed. Contributions on Neutronics are found to be of particular relevance and represent an important step towards the implementation of data base for nuclear design of plasma facing and breeding blanket components. (orig.).

  14. Mathematical Modeling for Simulation of Nuclear Reactor Analysis

    OpenAIRE

    Salah Ud-Din Khan; Shahab Ud-Din Khan

    2013-01-01

    In this paper, we have developed a mathematical model for the nuclear reactor analysis to be implemented in the nuclear reactor code. THEATRe is nuclear reactor analysis code which can only work for the cylindrical type fuel reactor and cannot applicable for the plate type fuel nuclear reactor. Therefore, the current studies encompasses on the modification of THEATRe code for the plate type fuel element. This mathematical model is applicable to the thermal analysis of the reactor which is ver...

  15. 100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

    Energy Technology Data Exchange (ETDEWEB)

    HARRINGTON RA

    2010-01-15

    On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery

  16. Continuous flow analysis of labile iron in ice-cores.

    Science.gov (United States)

    Hiscock, William T; Fischer, Hubertus; Bigler, Matthias; Gfeller, Gideon; Leuenberger, Daiana; Mini, Olivia

    2013-05-07

    The important active and passive role of mineral dust aerosol in the climate and the global carbon cycle over the last glacial/interglacial cycles has been recognized. However, little data on the most important aeolian dust-derived biological micronutrient, iron (Fe), has so far been available from ice-cores from Greenland or Antarctica. Furthermore, Fe deposition reconstructions derived from the palaeoproxies particulate dust and calcium differ significantly from the Fe flux data available. The ability to measure high temporal resolution Fe data in polar ice-cores is crucial for the study of the timing and magnitude of relationships between geochemical events and biological responses in the open ocean. This work adapts an existing flow injection analysis (FIA) methodology for low-level trace Fe determinations with an existing glaciochemical analysis system, continuous flow analysis (CFA) of ice-cores. Fe-induced oxidation of N,N'-dimethyl-p-pheylenediamine (DPD) is used to quantify the biologically more important and easily leachable Fe fraction released in a controlled digestion step at pH ~1.0. The developed method was successfully applied to the determination of labile Fe in ice-core samples collected from the Antarctic Byrd ice-core and the Greenland Ice-Core Project (GRIP) ice-core.

  17. NUCLEAR FORENSICS ANALYSIS CENTER FORENSIC ANALYSIS TO DATA INTERPRETATION

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, T.

    2011-02-07

    The Nuclear Forensics Analysis Center (NFAC) is part of Savannah River National Laboratory (SRNL) and is one of only two USG National Laboratories accredited to perform nuclear forensic analyses to the requirements of ISO 17025. SRNL NFAC is capable of analyzing nuclear and radiological samples from bulk material to ultra-trace samples. NFAC provides analytical support to the FBI's Radiological Evidence Examination Facility (REEF), which is located within SRNL. REEF gives the FBI the capability to perform traditional forensics on material that is radiological and/or is contaminated. SRNL is engaged in research and development efforts to improve the USG technical nuclear forensics capabilities. Research includes improving predictive signatures and developing a database containing comparative samples.

  18. A new equation of state for core-collapse supernovae based on realistic nuclear forces and including a full nuclear ensemble

    Science.gov (United States)

    Furusawa, S.; Togashi, H.; Nagakura, H.; Sumiyoshi, K.; Yamada, S.; Suzuki, H.; Takano, M.

    2017-09-01

    We have constructed a nuclear equation of state (EOS) that includes a full nuclear ensemble for use in core-collapse supernova simulations. It is based on the EOS for uniform nuclear matter that two of the authors derived recently, applying a variational method to realistic two- and three-body nuclear forces. We have extended the liquid drop model of heavy nuclei, utilizing the mass formula that accounts for the dependences of bulk, surface, Coulomb and shell energies on density and/or temperature. As for light nuclei, we employ a quantum-theoretical mass evaluation, which incorporates the Pauli- and self-energy shifts. In addition to realistic nuclear forces, the inclusion of in-medium effects on the full ensemble of nuclei makes the new EOS one of the most realistic EOSs, which covers a wide range of density, temperature and proton fraction that supernova simulations normally encounter. We make comparisons with the FYSS EOS, which is based on the same formulation for the nuclear ensemble but adopts the relativistic mean field theory with the TM1 parameter set for uniform nuclear matter. The new EOS is softer than the FYSS EOS around and above nuclear saturation densities. We find that neutron-rich nuclei with small mass numbers are more abundant in the new EOS than in the FYSS EOS because of the larger saturation densities and smaller symmetry energy of nuclei in the former. We apply the two EOSs to 1D supernova simulations and find that the new EOS gives lower electron fractions and higher temperatures in the collapse phase owing to the smaller symmetry energy. As a result, the inner core has smaller masses for the new EOS. It is more compact, on the other hand, due to the softness of the new EOS and bounces at higher densities. It turns out that the shock wave generated by core bounce is a bit stronger initially in the simulation with the new EOS. The ensuing outward propagations of the shock wave in the outer core are very similar in the two simulations, which

  19. Neutron analysis of the fuel of high temperature nuclear reactors; Analisis neutronico del combustible de reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  20. Nuclear analysis of the ITER Cryopump Ports

    Energy Technology Data Exchange (ETDEWEB)

    Moro, Fabio, E-mail: fabio.moro@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Flammini, Davide [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Antipenkov, Alexander; Dremel, Matthias; Levesy, Bruno; Loughlin, Michael [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Juarez, Rafael; Perez, Lucia [UNED, Energetic Engineering Department, C/Juan del Rosal 12, Madrid (Spain); Petrizzi, Luigino [European Commission, DG Research & Innovation G5, CDMA 00/030, B-1049 Brussels (Belgium)

    2015-10-15

    Highlights: • Evaluation the shielding effectiveness of the TCPHs by means of 3-D neutrons and gamma maps. • Assessment of the nuclear heating induced by neutron and photons on the TCP and TCPHs. • Calculation of the dose rate at 12 days after shutdown in the maintenance area of the Lower Ports with the Advanced D1S method, in order to verify the design target (100 μSv/h). • Potential improvements of the shielding configuration aimed at the reduction of the dose level in the Port Cell have been proposed and discussed. - Abstract: The ITER machine will be equipped with 6 torus Cryopumps (TCP) that are positioned in their housings (TCPH) and integrated into the cryostat walls at B1 level in the port cells. A comprehensive nuclear analysis of the Cryopump Ports #4 and #12 has been carried out by means of the MCNP-5 Monte Carlo code in a full 3-D geometry, providing guidelines for the design of the embedded components. Radiation transport calculations have been performed in order to determine the radiation field inside the Lower Ports, up the Port Cell: 3-D neutrons and gamma maps have been provided in order to evaluate the shielding effectiveness of the TCPHs. Nuclear heating induced by neutron and photons have been estimated on the TCP and TCPH to assess the nuclear loads during plasma operations. The shutdown dose rate in the maintenance area of the Lower Ports has been assessed with the Advanced D1S method to verify the design limits.

  1. Adiabatic hyperspherical analysis of realistic nuclear potentials

    CERN Document Server

    Daily, K M; Greene, Chris H

    2015-01-01

    Using the hyperspherical adiabatic method with the realistic nuclear potentials Argonne V14, Argonne V18, and Argonne V18 with the Urbana IX three-body potential, we calculate the adiabatic potentials and the triton bound state energies. We find that a discrete variable representation with the slow variable discretization method along the hyperradial degree of freedom results in energies consistent with the literature. However, using a Laguerre basis results in missing energy, even when extrapolated to an infinite number of basis functions and channels. We do not include the isospin $T=3/2$ contribution in our analysis.

  2. Risk and safety analysis of nuclear systems

    CERN Document Server

    Lee, John C

    2011-01-01

    The book has been developed in conjunction with NERS 462, a course offered every year to seniors and graduate students in the University of Michigan NERS program. The first half of the book covers the principles of risk analysis, the techniques used to develop and update a reliability data base, the reliability of multi-component systems, Markov methods used to analyze the unavailability of systems with repairs, fault trees and event trees used in probabilistic risk assessments (PRAs), and failure modes of systems. All of this material is general enough that it could be used in non-nuclear a

  3. Application Analysis of Strengthened Story in Frame-Core Structures

    Institute of Scientific and Technical Information of China (English)

    SU Yuan; CHEN Chuan-yao; LI Li

    2009-01-01

    Lateral deflection formulas are presented for analysis of the strengthened story applied to frame-core structures. For the frame-core structures with top outriggers and with middle outriggers, the relationship between stiffness characteristic parameters of frame and outriggers and the top drift of structures under different loads is analyzed. It is indicated that when stiffness characteristic parameter of frame is large, outrigger efficiency for top drift reduction is low, and the mutation of internal forces occurs; when the stiffness characteristic parameter of frame is less than 3, installing the strengthened story is advantageous to frame-core structures.

  4. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  5. TMI-2 accident: core heat-up analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ardron, K.H.; Cain, D.G.

    1981-01-01

    This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions.

  6. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  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. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-01-01

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

  10. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  11. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-04-15

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  12. A study on the nuclear foreign policy analysis

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Keun Bae; Choi, Y. M.; Lee, D. J.; Lee, K. S.; Lee, B. W.; Cho, I. H.; Ko, H. S.

    1996-12-01

    This study aims to analyses recent trends of international situation relating to nuclear non-proliferation and the adverse conditions in Korea`s pursuing self-support of such technology, so that it may map out effective strategies for the promotion of nuclear energy. This study analyses developments of international nuclear non-proliferation regime, which plays a main role in preventing the international proliferation of nuclear weapons. This study includes NPT, IAEA safeguards system, international export control regimes, CTBT, and NWFZs as the subjects of analysis. Second theme is international organizations concerning nuclear activities. This study mainly analyses IAEA activities which pursues the promotion of peaceful use of nuclear energy and nuclear non-proliferation simultaneously as a pivotal body of international nuclear cooperation. Third focus of this study is Northeast Asian circumstances pertaining to nuclear non-proliferation. The study looks into the DPRK nuclear issues, and reviews the developments of the proposed regional body for nuclear cooperation and the discussion on the Northeast Asian NWFZ. Fourth, but the most influential to Korean nuclear activities, is the U. S. nuclear policy, since U. S. takes the overwhelming initiative in the field of international nuclear non-proliferation. Therefore, this study gives much weight in analyzing the structure, procedures, recent trend, and pending issues of U. S. nuclear policy. (author). 78 refs., 5 tabs., 4 figs.

  13. A study on the nuclear foreign policy analysis

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Keun Bae; Choi, Y. M.; Lee, D. J.; Lee, K. S.; Lee, B. W.; Cho, I. H.; Ko, H. S.

    1996-12-01

    This study aims to analyses recent trends of international situation relating to nuclear non-proliferation and the adverse conditions in Korea`s pursuing self-support of such technology, so that it may map out effective strategies for the promotion of nuclear energy. This study analyses developments of international nuclear non-proliferation regime, which plays a main role in preventing the international proliferation of nuclear weapons. This study includes NPT, IAEA safeguards system, international export control regimes, CTBT, and NWFZs as the subjects of analysis. Second theme is international organizations concerning nuclear activities. This study mainly analyses IAEA activities which pursues the promotion of peaceful use of nuclear energy and nuclear non-proliferation simultaneously as a pivotal body of international nuclear cooperation. Third focus of this study is Northeast Asian circumstances pertaining to nuclear non-proliferation. The study looks into the DPRK nuclear issues, and reviews the developments of the proposed regional body for nuclear cooperation and the discussion on the Northeast Asian NWFZ. Fourth, but the most influential to Korean nuclear activities, is the U. S. nuclear policy, since U. S. takes the overwhelming initiative in the field of international nuclear non-proliferation. Therefore, this study gives much weight in analyzing the structure, procedures, recent trend, and pending issues of U. S. nuclear policy. (author). 78 refs., 5 tabs., 4 figs.

  14. Core Power Control of the fast nuclear reactors with estimation of the delayed neutron precursor density using Sliding Mode method

    Energy Technology Data Exchange (ETDEWEB)

    Ansarifar, G.R., E-mail: ghr.ansarifar@ast.ui.ac.ir; Nasrabadi, M.N.; Hassanvand, R.

    2016-01-15

    Highlights: • We present a S.M.C. system based on the S.M.O for control of a fast reactor power. • A S.M.O has been developed to estimate the density of delayed neutron precursor. • The stability analysis has been given by means Lyapunov approach. • The control system is guaranteed to be stable within a large range. • The comparison between S.M.C. and the conventional PID controller has been done. - Abstract: In this paper, a nonlinear controller using sliding mode method which is a robust nonlinear controller is designed to control a fast nuclear reactor. The reactor core is simulated based on the point kinetics equations and one delayed neutron group. Considering the limitations of the delayed neutron precursor density measurement, a sliding mode observer is designed to estimate it and finally a sliding mode control based on the sliding mode observer is presented. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Sliding Mode Control (SMC) is one of the robust and nonlinear methods which have several advantages such as robustness against matched external disturbances and parameter uncertainties. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability.

  15. Impact of the symmetry energy on nuclear pasta phases and crust-core transition in neutron stars

    CERN Document Server

    Bao, S S

    2015-01-01

    We study the impact of the symmetry energy on properties of nuclear pasta phases and crust-core transition in neutron stars. We perform a self-consistent Thomas--Fermi calculation employing the relativistic mean-field model. The properties of pasta phases presented in the inner crust of neutron stars are investigated and the crust-core transition is examined. It is found that the slope of the symmetry energy plays an important role in determining the pasta phase structure and the crust-core transition. The correlation between the symmetry energy slope and the crust-core transition density obtained in the Thomas--Fermi approximation is consistent with that predicted by the liquid-drop model.

  16. Petrographic Analysis of Cores from Plant 42

    Science.gov (United States)

    2016-10-01

    27  DISCLAIMER: The contents of this report are not to be used for advertising , publication...Department of the Army position unless so designated by other authorized documents. DESTROY THIS REPORT WHEN NO LONGER NEEDED. DO NOT RETURN IT TO THE... graphic analysis was polished using diamond incrusted polishing pads. The polished sample was imaged using a Zeiss Stereo Discovery V20 mi- croscope

  17. Development of a standard data base for FBR core nuclear design. 8. Compilation of JUPITER analytical results

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Makoto; Sugino, Kazuteru; Yokoyama, Kenji [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Sato, Wakaei; Numata, Kazuyuki; Iwai, Takehiko

    1997-11-01

    A standard data base for LMFBR core nuclear design has been developed to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as demonstration or commercial FBRs. To develop the data base, extensive work has been performed to accumulate and evaluate many kinds of results from fast reactor physics experiments and their analyses. The present report summarizes the analytical results of the JUPITER experiments, using the most recent nuclear data library (JENDL-3.2) and the latest analytical methods in a consistent manner. The total number of JUPITER C/E values obtained here exceeds 2,300, which cover most of the JUPITER data in the nuclear design data base. The analytical results will be combined with the sensitivity coefficients and experimental and analytical error values as a whole, and are expected to contribute the improvement of large FBR core design methods by means of a unified cross-section set for the demonstration FBR and various physical information. (J.P.N.). 236 refs.

  18. Multiscale Methods for Nuclear Reactor Analysis

    Science.gov (United States)

    Collins, Benjamin S.

    The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques, that use diffusion theory and pin power reconstruction (PPR). Two different multiscale methods were developed and analyzed; the post-refinement multiscale method and the embedded multiscale method. The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is "post-refinement" and thus has no impact on the global solution. The embedded multiscale method allows the local solver to change the global solution to provide an improved global and local solution. The post-refinement multiscale method is assessed using three core designs. When the local solution has more energy groups, the fixed source method has some difficulties near the interface: however the albedo method works well for all cases. In order to remedy the issue with boundary condition errors for the fixed source method, a buffer region is used to act as a filter, which decreases the sensitivity of the solution to the boundary condition. Both the albedo and fixed source methods benefit from the use of a buffer region. Unlike the post-refinement method, the embedded multiscale method alters the global solution. The ability to change the global solution allows for refinement in areas where the errors in the few group nodal diffusion are typically large. The embedded method is shown to improve the global solution when it is applied to a MOX/LEU assembly

  19. The proteins of intra-nuclear bodies: a data-driven analysis of sequence, interaction and expression

    Directory of Open Access Journals (Sweden)

    Bodén Mikael

    2010-04-01

    Full Text Available Abstract Background Cajal bodies, nucleoli, PML nuclear bodies, and nuclear speckles are morpohologically distinct intra-nuclear structures that dynamically respond to cellular cues. Such nuclear bodies are hypothesized to play important regulatory roles, e.g. by sequestering and releasing transcription factors in a timely manner. While the nucleolus and nuclear speckles have received more attention experimentally, the PML nuclear body and the Cajal body are still incompletely characterized in terms of their roles and protein complement. Results By collating recent experimentally verified data, we find that almost 1000 proteins in the mouse nuclear proteome are known to associate with one or more of the nuclear bodies. Their gene ontology terms highlight their regulatory roles: splicing is confirmed to be a core activity of speckles and PML nuclear bodies house a range of proteins involved in DNA repair. We train support-vector machines to show that nuclear proteins contain discriminative sequence features that can be used to identify their intra-nuclear body associations. Prediction accuracy is highest for nucleoli and nuclear speckles. The trained models are also used to estimate the full protein complement of each nuclear body. Protein interactions are found primarily to link proteins in the nuclear speckles with proteins from other compartments. Cell cycle expression data provide support for increased activity in nucleoli, nuclear speckles and PML nuclear bodies especially during S and G2 phases. Conclusions The large-scale analysis of the mouse nuclear proteome sheds light on the functional organization of physically embodied intra-nuclear compartments. We observe partial support for the hypothesis that the physical organization of the nucleus mirrors functional modularity. However, we are unable to unambiguously identify proteins' intra-nuclear destination, suggesting that critical drivers behind of intra-nuclear translocation are yet to

  20. Core Handling and Real-Time Non-Destructive Characterization at the Kochi Core Center: An Example of Core Analysis from the Chelungpu Fault

    Directory of Open Access Journals (Sweden)

    W. Lin

    2007-11-01

    Full Text Available As an example of core analysis carried out inactive fault drilling programs, we report the procedures of core handling on the drilling site and non-destructive characterization in the laboratory. This analysis was employed onthe core samples taken from HoleBof the Taiwan Chelungpu-fault Drilling Project (TCDP, which penetrated through the active fault that slipped during the 1999 Chi-Chi, Taiwan earthquake. We show results of the non-destructive physical property measurements carried out at the Kochi Core Center (KCC, Japan. Distinct anomalies of lower bulk density and higher magnetic susceptibilitywere recognized in all three fault zones encountered in HoleB. To keep the core samples in good condition before they are used for variousanalyses is crucial. In addition, careful planning for core handlingand core analyses is necessary for successfulinvestigations. doi:10.2204/iodp.sd.s01.35.2007

  1. Seismic analysis of nuclear power plant structures

    Science.gov (United States)

    Go, J. C.

    1973-01-01

    Primary structures for nuclear power plants are designed to resist expected earthquakes of the site. Two intensities are referred to as Operating Basis Earthquake and Design Basis Earthquake. These structures are required to accommodate these seismic loadings without loss of their functional integrity. Thus, no plastic yield is allowed. The application of NASTRAN in analyzing some of these seismic induced structural dynamic problems is described. NASTRAN, with some modifications, can be used to analyze most structures that are subjected to seismic loads. A brief review of the formulation of seismic-induced structural dynamics is also presented. Two typical structural problems were selected to illustrate the application of the various methods of seismic structural analysis by the NASTRAN system.

  2. Comparative analysis of the core inflation for Russia

    Directory of Open Access Journals (Sweden)

    A. K. Sapova

    2016-01-01

    Full Text Available Consumer price index is a measure of inflation and it consists of two parts: persistent component (trend inflation and short-term shocks. Inflation targeting requires index of core inflation, that independent from shortterm shocks and demonstrates the changes of the trend inflation. Reserve Banks pay attention on the changes in the trend inflation, when they take decisions about monetary policy, because it’s more informative than consumer price index for estimation of medium-term inflation risks. The objective of this article is detecting the index of core inflation that could be appropriate for monetary policy. There are some different measures of core inflation based on practice of Reserve Banks from different countries and economic articles. The comparative analysis presented in this article is based on several types of tests. The result of the research is that core consumer price index that is used today has got both advantages and weaknesses. Moreover, there is index of core inflation based on new methodology that is better than core consumer price index of Federal Sate Statistics Service. It is concluded that the Central Bank should focus precisely on this indicator when it takes decisions about monetary policy.

  3. Reliability of sprinkler systems. Exploration and analysis of data from nuclear and non-nuclear installations

    Energy Technology Data Exchange (ETDEWEB)

    Roenty, V.; Keski-Rahkonen, O.; Hassinen, J.P. [VTT Building and Transport, Espoo (Finland)

    2004-12-01

    Sprinkler systems are an important part of fire safety of nuclear installations. As a part of effort to make fire-PSA of our utilities more quantitative a literature survey from open sources worldwide of available reliability data on sprinkler systems was carried out. Since the result of the survey was rather poor quantitatively, it was decided to mine available original Finnish nuclear and non-nuclear data, since nuclear power plants present a rather small device population. Sprinklers are becoming a key element for the fire safety in modern, open non-nuclear buildings. Therefore, the study included both nuclear power plants and non-nuclear buildings protected by sprinkler installations. Data needed for estimating of reliability of sprinkler systems were collected from available sources in Finnish nuclear and non-nuclear installations. Population sizes on sprinkler system installations and components therein as well as covered floor areas were counted individually from Finnish nuclear power plants. From non-nuclear installations corresponding data were estimated by counting relevant things from drawings of 102 buildings, and plotting from that sample needed probability distributions. The total populations of sprinkler systems and components were compiled based on available direct data and these distributions. From nuclear power plants electronic maintenance reports were obtained, observed failures and other reliability relevant data were selected, classified according to failure severity, and stored on spreadsheets for further analysis. A short summary of failures was made, which was hampered by a small sample size. From non-nuclear buildings inspection statistics from years 1985.1997 were surveyed, and observed failures were classified and stored on spreadsheets. Finally, a reliability model is proposed based on earlier formal work, and failure frequencies obtained by preliminary data analysis of this work. For a model utilising available information in the non-nuclear

  4. Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

    2011-02-15

    The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

  5. Use of bone core biopsies for cytogenetic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Martin, P.; Rowley, J.D.; Baron, J.M.

    1979-01-01

    Cultures of bone core specimens have proved satisfactory for cytogenetic analysis in patients from whom it was impossible to obtain a bone marrow aspirate, or in whose peripheral blood dividing myeloid cells were absent or insufficient in number. The quality of the metaphase chromosome is adequate for banding studies.

  6. Stress analysis of portable safety platform (Core Sampler Truck)

    Energy Technology Data Exchange (ETDEWEB)

    Ziada, H.H.

    1995-03-30

    This document provides the stress analysis and evaluation of the portable platform of the rotary mode core sampler truck No. 2 (RMCST {number_sign}2). The platform comprises railing, posts, deck, legs, and a portable ladder; it is restrained from lateral motion by means of two brackets added to the drill-head service platform.

  7. The features of Drosophila core promoters revealed by statistical analysis

    Directory of Open Access Journals (Sweden)

    Trifonov Edward N

    2006-06-01

    Full Text Available Abstract Background Experimental investigation of transcription is still a very labor- and time-consuming process. Only a few transcription initiation scenarios have been studied in detail. The mechanism of interaction between basal machinery and promoter, in particular core promoter elements, is not known for the majority of identified promoters. In this study, we reveal various transcription initiation mechanisms by statistical analysis of 3393 nonredundant Drosophila promoters. Results Using Drosophila-specific position-weight matrices, we identified promoters containing TATA box, Initiator, Downstream Promoter Element (DPE, and Motif Ten Element (MTE, as well as core elements discovered in Human (TFIIB Recognition Element (BRE and Downstream Core Element (DCE. Promoters utilizing known synergetic combinations of two core elements (TATA_Inr, Inr_MTE, Inr_DPE, and DPE_MTE were identified. We also establish the existence of promoters with potentially novel synergetic combinations: TATA_DPE and TATA_MTE. Our analysis revealed several motifs with the features of promoter elements, including possible novel core promoter element(s. Comparison of Human and Drosophila showed consistent percentages of promoters with TATA, Inr, DPE, and synergetic combinations thereof, as well as most of the same functional and mutual positions of the core elements. No statistical evidence of MTE utilization in Human was found. Distinct nucleosome positioning in particular promoter classes was revealed. Conclusion We present lists of promoters that potentially utilize the aforementioned elements/combinations. The number of these promoters is two orders of magnitude larger than the number of promoters in which transcription initiation was experimentally studied. The sequences are ready to be experimentally tested or used for further statistical analysis. The developed approach may be utilized for other species.

  8. Source term analysis for a nuclear submarine accident

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J.; Hugron, J.J.M.R. [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    1999-07-01

    A source term analysis has been conducted to determine the activity release into the environment as a result of a large-break loss-of-coolant accident aboard a visiting nuclear-powered submarine to a Canadian port. This best-estimate analysis considers the fractional release from the core, and fission product transport in the primary heat transport system, primary containment (i.e. reactor compartment) and submarine hull. Physical removal mechanisms such as vapour and aerosol deposition are treated in the calculation. Since a thermalhydraulic analysis indicated that the integrity of the reactor compartment is maintained, release from the reactor compartment will only occur by leakage; however, it is conservatively assumed that the secondary containment is not isolated for a 24-h period where release occurs through an open hatch in the submarine hull. Consequently, during this period, the activity release into the atmosphere is estimated as 4.6 TBq, leading to a maximum individual dose equivalent of 0.5 mSv at 800 metres from the berthing location. This activity release is comparable to that obtained in the BEREX TSA study (for a similar accident scenario) but is four orders of magnitude less than that reported in the earlier Davis study where, unrealistically, no credit had been taken for the containment system or for any physical removal processes. (author)

  9. Magnetic resonance imaging in laboratory petrophysical core analysis

    Science.gov (United States)

    Mitchell, J.; Chandrasekera, T. C.; Holland, D. J.; Gladden, L. F.; Fordham, E. J.

    2013-05-01

    Magnetic resonance imaging (MRI) is a well-known technique in medical diagnosis and materials science. In the more specialized arena of laboratory-scale petrophysical rock core analysis, the role of MRI has undergone a substantial change in focus over the last three decades. Initially, alongside the continual drive to exploit higher magnetic field strengths in MRI applications for medicine and chemistry, the same trend was followed in core analysis. However, the spatial resolution achievable in heterogeneous porous media is inherently limited due to the magnetic susceptibility contrast between solid and fluid. As a result, imaging resolution at the length-scale of typical pore diameters is not practical and so MRI of core-plugs has often been viewed as an inappropriate use of expensive magnetic resonance facilities. Recently, there has been a paradigm shift in the use of MRI in laboratory-scale core analysis. The focus is now on acquiring data in the laboratory that are directly comparable to data obtained from magnetic resonance well-logging tools (i.e., a common physics of measurement). To maintain consistency with well-logging instrumentation, it is desirable to measure distributions of transverse (T2) relaxation time-the industry-standard metric in well-logging-at the laboratory-scale. These T2 distributions can be spatially resolved over the length of a core-plug. The use of low-field magnets in the laboratory environment is optimal for core analysis not only because the magnetic field strength is closer to that of well-logging tools, but also because the magnetic susceptibility contrast is minimized, allowing the acquisition of quantitative image voxel (or pixel) intensities that are directly scalable to liquid volume. Beyond simple determination of macroscopic rock heterogeneity, it is possible to utilize the spatial resolution for monitoring forced displacement of oil by water or chemical agents, determining capillary pressure curves, and estimating

  10. Planning of the development of the MMIS core technology based on nuclear-IT convergence

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Kee Choon; Kim, Chang Hwoi; Hwang, In Koo [KAERI, Daejeon (Korea, Republic of); and others

    2012-01-15

    - Drive nuclear-IT convergence technologies such as middleware applied new concept nuclear instrumentation and control architecture, automated operation of future nuclear power plant, virtual reality/augmented reality, design and verification technology of a nuclear power plant main control room, software dependability, and cyber security technology - Write state-of-the-art report for the nuclear instrumentation and control based on IT convergence - A prototype which implemented related equipment and software subject to nuclear reactor operator that reside in the main control room (Reactor Operator, RO) order to a on-site operator (Local Operator, LO) and confirm the task performance matches the RO's intention - 'IT Convergence intelligent instrumentation and control technology' project planning for the Fourth Nuclear Power Research and Development in the long-term plan.

  11. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach; Simulacao termohidraulica do nucleo do reator nuclear HTR-10 com o uso da abordagem realistica CFD

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S.; Dominguez, Dany S., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil); Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas, La Habana (Cuba); Lira, Carlos Alberto Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  12. Size analysis of single-core magnetic nanoparticles

    Science.gov (United States)

    Ludwig, Frank; Balceris, Christoph; Viereck, Thilo; Posth, Oliver; Steinhoff, Uwe; Gavilan, Helena; Costo, Rocio; Zeng, Lunjie; Olsson, Eva; Jonasson, Christian; Johansson, Christer

    2017-04-01

    Single-core iron-oxide nanoparticles with nominal core diameters of 14 nm and 19 nm were analyzed with a variety of non-magnetic and magnetic analysis techniques, including transmission electron microscopy (TEM), dynamic light scattering (DLS), static magnetization vs. magnetic field (M-H) measurements, ac susceptibility (ACS) and magnetorelaxometry (MRX). From the experimental data, distributions of core and hydrodynamic sizes are derived. Except for TEM where a number-weighted distribution is directly obtained, models have to be applied in order to determine size distributions from the measurand. It was found that the mean core diameters determined from TEM, M-H, ACS and MRX measurements agree well although they are based on different models (Langevin function, Brownian and Néel relaxation times). Especially for the sample with large cores, particle interaction effects come into play, causing agglomerates which were detected in DLS, ACS and MRX measurements. We observed that the number and size of agglomerates can be minimized by sufficiently strong diluting the suspension.

  13. Optimization Design and Finite Element Analysis of Core Cutter

    Institute of Scientific and Technical Information of China (English)

    CAO Pin-lu; YIN Kun; PENG Jian-ming; LIU Jian-lin

    2007-01-01

    The hydro-hammer sampler is a new type of sampler compared with traditional ones. An important part of this new offshore sampler is that the structure of the core cutter has a significant effect on penetration and core recovery. In our experiments, a commercial finite element code with a capability of simulating large-strain frictional contact between two or more solid bodies is used to simulate the core cutter-soil interaction. The effects of the cutting edge shape, the diameter and the edge angle on penetration are analyzed by non-liner transient dynamic analysis using a finite element method (FEM). Simulation results show that the cutter shape clearly has an effect on the penetration and core recovery. In addition, the penetration of the sampler increases with an increase in the inside diameter of the cutter, but decreases with an increase in the cutting angle. Based on these analyses, an optimum structure of the core cutter is designed and tested in the north margin of the Dalian gulf. Experiment results show that the penetration rate is about 16.5 m/h in silty clay and 15.4 m/h in cohesive clay, while the recovery is 68% and 83.3% respectively.

  14. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

    Energy Technology Data Exchange (ETDEWEB)

    Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L. (Science Applications International Corp., Albuquerque, NM (USA); Sandia National Labs., Albuquerque, NM (USA))

    1989-08-01

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs.

  15. Preliminary safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

    2001-06-01

    KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

  16. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

    This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

  17. Building Foundations for Nuclear Security Enterprise Analysis Utilizing Nuclear Weapon Data.

    Energy Technology Data Exchange (ETDEWEB)

    Josserand, Terry Michael; Young, Leone; Chamberlin, Edwin Phillip,

    2017-10-01

    T he Nuclear Security Enterprise , managed by the National Nuclear Security Administration - a semiautonomous agency within the Department of Energy - has been associated with numerous assessments with respect to the estimating, management capabilities, and practices pertaining to nuclear weapon modernization efforts. This report identifies challenges in estimating and analyzing the N uclear S ecurity E nterprise through an analysis of analogous timeframe conditions utilizing two types of nuclear weapon data - (1) a measure of effort and (2) a function of time. The analysis of analogous timeframe conditions that utilizes only two types of nuclear weapon d ata yields four summary observations that estimators and analysts of the N uclear S ecurity E nterprise will find useful. This Page Intentionally Left Blank

  18. Degraded core analysis for the pressurized-water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-02-09

    An analysis of the likelihood and the consequences of 'degraded-core accidents' has been undertaken for the proposed Sizewell B PWR. In such accidents, degradation of the core geometry occurs as a result of overheating. Radionuclides are released and may enter the environment, causing harmful effects. The analysis concludes that degraded-core accidents are highly improbable, the plant having been designed to reduce the frequency of such accidents to a value of order 10/sup -6/ per year. Tbe building containing the reactor would only fail in a small proportion of degraded-core accidents. In the great majority of cases the containment would remain intact and the release of radioactivity to the environment would be small. The risk to individuals have been calculated for both immediate and long term effects. Although the estimates of risk are approximate, studies to investigate the uncertainties, and sensitivities to different assumptions, show that potential errors are small compared with the very large 'margin of safety' between the risks estimated for Sizewell B and those that already exist in society.

  19. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  20. High Resolution Continuous Flow Analysis System for Polar Ice Cores

    Science.gov (United States)

    Dallmayr, Remi; Azuma, Kumiko; Yamada, Hironobu; Kjær, Helle Astrid; Vallelonga, Paul; Azuma, Nobuhiko; Takata, Morimasa

    2014-05-01

    In the last decades, Continuous Flow Analysis (CFA) technology for ice core analyses has been developed to reconstruct the past changes of the climate system 1), 2). Compared with traditional analyses of discrete samples, a CFA system offers much faster and higher depth resolution analyses. It also generates a decontaminated sample stream without time-consuming sample processing procedure by using the inner area of an ice-core sample.. The CFA system that we have been developing is currently able to continuously measure stable water isotopes 3) and electrolytic conductivity, as well as to collect discrete samples for the both inner and outer areas with variable depth resolutions. Chemistry analyses4) and methane-gas analysis 5) are planned to be added using the continuous water stream system 5). In order to optimize the resolution of the current system with minimal sample volumes necessary for different analyses, our CFA system typically melts an ice core at 1.6 cm/min. Instead of using a wire position encoder with typical 1mm positioning resolution 6), we decided to use a high-accuracy CCD Laser displacement sensor (LKG-G505, Keyence). At the 1.6 cm/min melt rate, the positioning resolution was increased to 0.27mm. Also, the mixing volume that occurs in our open split debubbler is regulated using its weight. The overflow pumping rate is smoothly PID controlled to maintain the weight as low as possible, while keeping a safety buffer of water to avoid air bubbles downstream. To evaluate the system's depth-resolution, we will present the preliminary data of electrolytic conductivity obtained by melting 12 bags of the North Greenland Eemian Ice Drilling (NEEM) ice core. The samples correspond to different climate intervals (Greenland Stadial 21, 22, Greenland Stadial 5, Greenland Interstadial 5, Greenland Interstadial 7, Greenland Stadial 8). We will present results for the Greenland Stadial -8, whose depths and ages are between 1723.7 and 1724.8 meters, and 35.520 to

  1. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-05-15

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  2. Preliminary Core Analysis of a Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Chang Keun; Chang, Jongwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Venneri, Francesco [Ultra Safe Nuclear Corporation, Los Alamos (United States); Hawari, Ayman [NC State Univ., Raleigh (United States)

    2014-05-15

    The Micro Modular Reactor (MMR) will be 'melt-down proof'(MDP) under all circumstances, including the complete loss of coolant, and will be easily transportable and retrievable, and suitable for use with very little site preparation and Balance of Plant (BOP) requirements for a variety of applications, from power generation and process heat applications in remote areas to grid-unattached locations, including ship propulsion. The Micro Modular Reactor design proposed in this paper has 3 meter diameter core (2 meter active core) which is suitable for 'factory manufactured' and has few tens year of service life for remote deployment. We confirmed the feasibility of long term service life by a preliminary neutronic analysis in terms of the excess reactivity, the temperature feedback coefficient, and the control margins. We are able to achieve a reasonably long core life time of 5 ∼ 10 years under typical thermal hydraulic condition of a helium cooled reactor. However, on a situation where longer service period and safety is important, we can reduce the power density to the level of typical pebble bed reactor. In this case we can design 10 MWt MMR with core diameter for 10 ∼ 40 years core life time without much loss in the economics. Several burnable poisons are studied and it is found that erbia mixed in the compact matrix seems reasonably good poison. The temperature feedback coefficients were remaining negative during lifetime. Drum type control rods at reflector region and few control rods inside core region are sufficient to control the reactivity during operation and to achieve safe cold shutdown state.

  3. Accelerator-Driven Subcritical Fission in a Molten Salt Core: Green Nuclear Power for the New Millennium

    Science.gov (United States)

    McIntyre, Peter

    2011-10-01

    Scientists at Texas A&M University, Brookhaven National Lab, and Idaho National Lab are developing a design for accelerator-drive subcritical fission in a molten salt core (ADSMS). Three high-power proton beams are delivered to spallation targets in a molten salt core, where they provide ˜3% of the fast neutrons required to sustain 600 MW of fission. The proton beams are produced by a flux-coupled stack of superconducting strong-focusing cyclotrons. The fuel consists of a eutectic of sodium chloride with either spent nuclear fuel from a conventional U power reactor (ADSMS-U) or thorium (ADSMS-Th). The subcritical core cannot go critical under any failure mode. The core cannot melt down even if all power is suddenly lost to the facility for a prolonged period. The ultra-fast neutronics of the core makes it possible to operate in an isobreeding mode, in which neutron capture breeds the fertile nuclide into a fissile nuclide at the same rate that fission burns the fissile nuclide, and consumes 90% of the fertile inventory instead of the 5% consumed in the original use in a conventional power plant. The ultra-fast neutronics produces a very low equilibrium inventory of the long-lived minor actinides, ˜10^4 less than what is produced in conventional power plants. ADSMS offers a method to safely produce the energy needs for all mankind for the next 3000 years.

  4. 10th Australian conference on nuclear techniques of analysis. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-06-01

    These proceedings contains abstracts and extended abstracts of 80 lectures and posters presented at the 10th Australian conference on nuclear techniques of analysis hosted by the Australian National University in Canberra, Australia from 24-26 of November 1997. The conference was divided into sessions on the following topics : ion beam analysis and its applications; surface science; novel nuclear techniques of analysis, characterization of thin films, electronic and optoelectronic material formed by ion implantation, nanometre science and technology, plasma science and technology. A special session was dedicated to new nuclear techniques of analysis, future trends and developments. Separate abstracts were prepared for the individual presentation included in this volume.

  5. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  6. Analysis of the core genome and pangenome of Pseudomonas putida.

    Science.gov (United States)

    Udaondo, Zulema; Molina, Lázaro; Segura, Ana; Duque, Estrella; Ramos, Juan L

    2016-10-01

    Pseudomonas putida are strict aerobes that proliferate in a range of temperate niches and are of interest for environmental applications due to their capacity to degrade pollutants and ability to promote plant growth. Furthermore solvent-tolerant strains are useful for biosynthesis of added-value chemicals. We present a comprehensive comparative analysis of nine strains and the first characterization of the Pseudomonas putida pangenome. The core genome of P. putida comprises approximately 3386 genes. The most abundant genes within the core genome are those that encode nutrient transporters. Other conserved genes include those for central carbon metabolism through the Entner-Doudoroff pathway, the pentose phosphate cycle, arginine and proline metabolism, and pathways for degradation of aromatic chemicals. Genes that encode transporters, enzymes and regulators for amino acid metabolism (synthesis and degradation) are all part of the core genome, as well as various electron transporters, which enable aerobic metabolism under different oxygen regimes. Within the core genome are 30 genes for flagella biosynthesis and 12 key genes for biofilm formation. Pseudomonas putida strains share 85% of the coding regions with Pseudomonas aeruginosa; however, in P. putida, virulence factors such as exotoxins and type III secretion systems are absent.

  7. A feasibility study for the application of enriched gadolinia burnable absorber rods in nuclear core design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chung Chan; Zee, Sung Quun; Kim, Kang Seog; Song, Jae Seung

    2000-12-01

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions.

  8. Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, E. A.; Kulak, R. F.; Therios, I. U.; Wei, T. Y. C.

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics

  9. Error Analysis of High Frequency Core Loss Measurement for Low-Permeability Low-Loss Magnetic Cores

    DEFF Research Database (Denmark)

    Niroumand, Farideh Javidi; Nymand, Morten

    2016-01-01

    in magnetic cores is B-H loop measurement where two windings are placed on the core under test. However, this method is highly vulnerable to phase shift error, especially for low-permeability, low-loss cores. Due to soft saturation and very low core loss, low-permeability low-loss magnetic cores are favorable....... The analysis has been validated by experimental measurements for relatively low-loss magnetic cores with different permeability values.......Magnetic components significantly contribute to the dissipated loss in power electronic converters. Measuring the true value of dissipated power in these components is highly desirable, since it can be used to verify the optimum design of these components. The common approach for measuring the loss...

  10. Dynamic Analysis of Nuclear Waste Generation Based on Nuclear Fuel Cycle Transition Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, S. R. [University of Science and Technology, Daejeon (Korea, Republic of); Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    According to the recommendations submitted by the Public Engagement Commission on Spent Nuclear Fuel Management (PECOS), the government was advised to pick the site for an underground laboratory and interim storage facilities before the end of 2020 followed by the related research for permanent and underground disposal of spent fuel after 10 years. In the middle of the main issues, the factors of environmentally friendly and safe way to handle nuclear waste are inextricable from nuclear power generating nation to ensure the sustainability of nuclear power. For this purposes, the closed nuclear fuel cycle has been developed regarding deep geological disposal, pyroprocessing, and burner type sodium-cooled fast reactors (SFRs) in Korea. Among two methods of an equilibrium model and a dynamic model generally used for screening nuclear fuel cycle system, the dynamic model is more appropriate to envisage country-specific environment with the transition phase in the long term and significant to estimate meaningful impacts based on the timedependent behavior of harmful wastes. This study aims at analyzing the spent nuclear fuel generation based on the long-term nuclear fuel cycle transition scenarios considered at up-to-date country specific conditions and comparing long term advantages of the developed nuclear fuel cycle option between once-through cycle and Pyro-SFR cycle. In this study, a dynamic analysis was carried out to estimate the long-term projection of nuclear electricity generation, installed capacity, spent nuclear fuel arising in different fuel cycle scenarios based on the up-to-date national energy plans.

  11. 47{sup th} Annual meeting on nuclear technology (AMNT 2016). About cores, coal and cash

    Energy Technology Data Exchange (ETDEWEB)

    Podivinsky, Tomas Jan

    2016-07-15

    Rationality and - especially with regard to reducing emissions - technological neutrality are two commitments for nuclear fission. The Czech Republic, where conditions are not suitable for economical large-scale operation of facilities based on renewables, there is no alternative in environmental or business policy to the reasonable use of nuclear energy. The aim of the updated Czech energy strategy is to increase the proportion of nuclear energy from 35 % to approx. 50 % of power generation and to cover the rest - together with ultra-high efficiency coal fired power plants - with energy from renewable sources and gas fired power plants.

  12. Multivariate Regression Analysis of Gravitational Waves from Rotating Core Collapse

    CERN Document Server

    Engels, William J; Ott, Christian D

    2014-01-01

    We present a new multivariate regression model for analysis and parameter estimation of gravitational waves observed from well but not perfectly modeled sources such as core-collapse supernovae. Our approach is based on a principal component decomposition of simulated waveform catalogs. Instead of reconstructing waveforms by direct linear combination of physically meaningless principal components, we solve via least squares for the relationship that encodes the connection between chosen physical parameters and the principal component basis. Although our approach is linear, the waveforms' parameter dependence may be non-linear. For the case of gravitational waves from rotating core collapse, we show, using statistical hypothesis testing, that our method is capable of identifying the most important physical parameters that govern waveform morphology in the presence of simulated detector noise. We also demonstrate our method's ability to predict waveforms from a principal component basis given a set of physical ...

  13. The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    Science.gov (United States)

    Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

    2017-01-01

    The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

  14. Single hepatitis-B virus core capsid binding to individual nuclear pore complexes in Hela cells.

    Science.gov (United States)

    Lill, Yoriko; Lill, Markus A; Fahrenkrog, Birthe; Schwarz-Herion, Kyrill; Paulillo, Sara; Aebi, Ueli; Hecht, Bert

    2006-10-15

    We investigate the interaction of hepatitis B virus capsids lacking a nuclear localization signal with nuclear pore complexes (NPCs) in permeabilized HeLa cells. Confocal and wide-field optical images of the nuclear envelope show well-spaced individual NPCs. Specific interactions of capsids with single NPCs are characterized by extended residence times of capsids in the focal volume which are characterized by fluorescence correlation spectroscopy. In addition, single-capsid-tracking experiments using fast wide-field fluorescence microscopy at 50 frames/s allow us to directly observe specific binding via a dual-color colocalization of capsids and NPCs. We find that binding occurs with high probability on the nuclear-pore ring moiety, at 44 +/- 9 nm radial distance from the central axis.

  15. Study on core concept for commercial LMFBR plant toward self-consistent nuclear energy system concept

    Energy Technology Data Exchange (ETDEWEB)

    Toukura, A. [Institute of Applied Energy, Tokyo (Japan); Yamazaki, M. [Toshiba Corp., Fuchu, Tokyo (Japan). Fuchu Works; Ohashi, M. [Hitachi Ltd., Ibaraki (Japan). Hitachi Works; Ikeda, K. [Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan); Saito, M.; Fujiie, Y. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    1995-12-31

    Fast Breeder Reactor (FBR) is expected to be commercialized in Japan to overcome foreseeable problems such as reactor safety, increasing energy demand, final disposal of high level radioactive waste and fuel resource shortage. We have been studying three FBR core concepts enhancing its potential abilities; ultra-large type, simplified type and friendly to fuel cycle type core. This study is sponsored by Ministry of International Trade and Industry. (author).

  16. Health physics activities in support of the thermal shield removal/disposal and core support barrel repair at the St. Lucie Nuclear Power Plant.

    Science.gov (United States)

    Maisler, J J; Buchanan, H F

    1988-02-01

    The health physics activities related to the removal and disposal of a thermal shield at a nuclear power plant and subsequent repairs to the core support barrel required increased planning relative to a normal refueling/maintenance outage. The repair of the core support barrel was a "first" in the nuclear power industry. Pre-job planning was of great concern because of extremely high radiation levels associated with the irradiated stainless steel thermal shield and core support barrel. ALARA techniques used in the preparation of the thermal shield for removal and shipment to the disposal site are discussed.

  17. Pygmy and core polarization dipole modes in 206Pb: Connecting nuclear structure to stellar nucleosynthesis

    Science.gov (United States)

    Tonchev, A. P.; Tsoneva, N.; Bhatia, C.; Arnold, C. W.; Goriely, S.; Hammond, S. L.; Kelley, J. H.; Kwan, E.; Lenske, H.; Piekarewicz, J.; Raut, R.; Rusev, G.; Shizuma, T.; Tornow, W.

    2017-10-01

    A high-resolution study of the electromagnetic response of 206Pb below the neutron separation energy is performed using a (γ → ,γ‧) experiment at the HI γ → S facility. Nuclear resonance fluorescence with 100% linearly polarized photon beams is used to measure spins, parities, branching ratios, and decay widths of excited states in 206Pb from 4.9 to 8.1 MeV. The extracted ΣB (E 1) ↑ and ΣB (M 1) ↑ values for the total electric and magnetic dipole strength below the neutron separation energy are 0.9 ± 0.2 e2fm2 and 8.3 ± 2.0 μN2, respectively. These measurements are found to be in very good agreement with the predictions from an energy-density functional (EDF) plus quasiparticle phonon model (QPM). Such a detailed theoretical analysis allows to separate the pygmy dipole resonance from both the tail of the giant dipole resonance and multi-phonon excitations. Combined with earlier photonuclear experiments above the neutron separation energy, one extracts a value for the electric dipole polarizability of 206Pb of αD = 122 ± 10 mb /MeV. When compared to predictions from both the EDF+QPM and accurately calibrated relativistic EDFs, one deduces a range for the neutron-skin thickness of Rskin206 = 0.12- 0.19 fm and a corresponding range for the slope of the symmetry energy of L = 48- 60 MeV. This newly obtained information is also used to estimate the Maxwellian-averaged radiative cross section 205Pb (n , γ)206Pb at 30 keV to be σ = 130 ± 25 mb. The astrophysical impact of this measurement-on both the s-process in stellar nucleosynthesis and on the equation of state of neutron-rich matter-is discussed.

  18. Ultrasonic spectral analysis for nuclear fuel characterization

    Energy Technology Data Exchange (ETDEWEB)

    Baroni, Douglas B.; Bittencourt, Marcelo S.Q.; Leal, Antonio M.M., E-mail: douglasbaroni@ien.gov.b, E-mail: bittenc@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    Ceramic materials have been widely used for various purposes in many different industries due to certain characteristics, such as high melting point and high resistance to corrosion. Concerning the areas of applications, automobile, aeronautics, naval and even nuclear, the characteristics of these materials should be strictly controlled. In the nuclear area, ceramics are of great importance once they are the nuclear fuel pellets and must have, among other features, a well controlled porosity due to mechanical strength and thermal conductivity required by the application. Generally, the techniques used to characterize nuclear fuel are destructive and require costly equipment and facilities. This paper aims to present a nondestructive technique for ceramic characterization using ultrasound. This technique differs from other ultrasonic techniques because it uses ultrasonic pulse in frequency domain instead of time domain, associating the characteristics of the analyzed material with its frequency spectrum. In the present work, 40 Alumina (Al{sub 2}O{sub 3}) ceramic pellets with porosities ranging from 5% to 37%, in absolute terms measured by Archimedes technique, were tested. It can be observed that the frequency spectrum of each pellet varies according to its respective porosity and microstructure, allowing a fast and non-destructive association of the same characteristics with the same spectra pellets. (author)

  19. Analysis and prediction of leucine-rich nuclear export signals

    DEFF Research Database (Denmark)

    La Cour, T.; Kiemer, Lars; Mølgaard, Anne;

    2004-01-01

    We present a thorough analysis of nuclear export signals and a prediction server, which we have made publicly available. The machine learning prediction method is a significant improvement over the generally used consensus patterns. Nuclear export signals (NESs) are extremely important regulators...... this analysis is that the most important properties of NESs are accessibility and flexibility allowing relevant proteins to interact with the signal. Furthermore, we show that not only the known hydrophobic residues are important in defining a nuclear export signals. We employ both neural networks and hidden...

  20. Experimental determination of nuclear parameters for RP-0 reactor core; Determinacion experimental de los parametros nucleares para el nucleo tipo MTR del reactor nuclear RP-0

    Energy Technology Data Exchange (ETDEWEB)

    Cajacuri, Rafael A. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica

    2000-07-01

    In the nuclear reactor for investigations RP-0 which is in Lima, Peru, that is a open pool class reactor with 1 to 10 watts of power and as a nuclear fuel uranium 238 enriched to 20% constituted by elements of Material Testing Reactor fuel class. This has reflectors of graphite and moderator of water demineralized. In 1996/1997 was measured in this reactor the following parameters: position of the control bar that make critic the reactor, critic height of moderator, excess of reactivity of the nucleus, parameter of reactivity for vacuum, parameter of reactivity for temperature, reactivity of its control bar, levels of doses in the reactor. (author)

  1. Neutronic analysis of LMFBRs during severe core disruptive accidents

    Energy Technology Data Exchange (ETDEWEB)

    Tomlinson, E.T.

    1979-01-01

    A number of numerical experiments were performed to assess the validity of diffusion theory and various perturbation methods for calculating the reactivity state of a severely disrupted liquid metal cooled fast breeder reactor (LMFBR). The disrupted configurations correspond, in general, to phases through which an LMFBR core could pass during a core disruptive accident (CDA). Two-reactor models were chosen for this study, the two zone, homogeneous Clinch River Breeder Reactor and the Large Heterogeneous Reactor Design Study Core. The various phases were chosen to approximate the CDA results predicted by the safety analysis code SAS3D. The calculational methods investigated in this study include the eigenvalue difference technique based on both discrete ordinate transport theory and diffusion theory, first-order perturbation theory, exact perturbation theory, and a new hybrid perturbation theory. Selected cases were analyzed using Monte Carlo methods. It was found that in all cases, diffusion theory and perturbation theory yielded results for the change in reactivity that significantly disagreed with both the discrete ordinate and Monte Carlo results. These differences were, in most cases, in a nonconservative direction.

  2. Analysis of Core Degradation in Fukushima Unit 1 Accident with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Kim, Tae Woon; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    In this study, an accident analysis of Fukushima Daiichi Unit 1 was performed using MELCOR 1.8.6. The behavior of the initial stage of the accident was focused during 30 hours after the reactor scram, because it was predicted that the vessel failure (severe accident) occurred before 20 hours. A hydrogen explosion also occurred at about 24 hours after the accident, and thus the phenomenon of core degradation before 30 hours was highlighted. Moreover, the effect of the amount of fresh water injection on the core degradation was performed by changing the amount of injection water. It was expected that a large portion of the injection water could not reach the core because of leakage. Thus, core damage was observed according to the amount of water that reached the core. The plant geometries and operating conditions were obtained from TEPCO (Tokyo Electric Power Company) through the OECD/NEA BSAF (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station) Project. An analysis of the Fukushima accident was also performed by Sandia National Laboratories, but several conditions were revised and added in this study. First, the flow rate of the steam into turbine and water into downcomer were considered at the initial stage of accident. The water level followed well the measured data by adding this mechanism. Second, SRV stuck open was included in this calculation. SRV stuck open can occur due to high temperature and frequent operation, and was modeled in calculation. The timing of SRV stuck open was closed to the timing of MSL failure, and thus the depressurization of RPV could have originated from both of MSL failure and SRV stuck open. The effect of injection water was observed, and it was found that the proper water injection can prevent a severe accident at the initial stage of the accident. In conclusion, an analysis of the severe accident occurring in Fukushima Unit 1 was conducted by using MELCOR. The analysis results were consistent with the

  3. Eigenvalue analysis using a full-core Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Okafor, K.C.; Zino, J.F. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1992-01-01

    The reactor physics codes used at the Savannah River Site (SRS) to predict reactor behavior have been continually benchmarked against experimental and operational data. A particular benchmark variable is the observed initial critical control rod position. Historically, there has been some difficulty predicting this position because of the difficulties inherent in using computer codes to model experimental or operational data. The Monte Carlo method is applied in this paper to study the initial critical control rod positions for the SRS K Reactor. A three-dimensional, full-core MCNP model of the reactor was developed for this analysis.

  4. Geoantineutrino Spectrum and Slow Nuclear Burning on the Boundary of the Liquid and Solid Phases of the Earth's core

    CERN Document Server

    Rusov, V D; Khotyaintseva, E N; Kosenko, S I; Litvinov, D A; Pavlovich, V N; Tarasov, V A; Vaschenko, V N; Zelentsova, T N

    2004-01-01

    The problem of the geoantineutrino deficit and the experimental results of the interaction of uranium dioxide and carbide with iron-nickel and silica-alumina melts at high pressure (5-10 GPa) and temperature (1600- 22000 C) have induced us to consider the possible consequences of made by V. Anisichkin and A. Ershov supposition that there is an actinoid shell on boundary of liquid and solid phases of the Earth's core. We have shown that the activation of a natural nuclear reactor operating as the solitary waves of nuclear burning in 238U- and/or 232Th-medium (in particular, the neutron-fission progressive wave of Feoktistov and/or Teller-Ishikawa-Wood) such physical consequent can be. The simplified model of the kinetics of accumulation and burnup in U-Pu fuel cycle of Feoktistov is developed. The results of the numerical simulation of neutron-fission wave in two-phase UO2/Fe medium on a surface of the Earth's solid core are presented. On the basis of O'Nions-Ivensen-Hamilton model of the geochemical evolution...

  5. Application of the nuclear equation of state obtained by the variational method to core-collapse supernovae

    CERN Document Server

    Togashi, H; Sumiyoshi, K; Nakazato, K

    2014-01-01

    The equation of state (EOS) for hot asymmetric nuclear matter which is constructed with the variational method starting from the Argonne v18 and Urbana IX nuclear forces is applied to spherically symmetric core-collapse supernovae (SNe). We first investigate the EOS of isentropic beta-stable SN matter, and find that the matter with the variational EOS is more neutron-rich than that with the Shen EOS. Using the variational EOS for uniform matter supplemented by the Shen EOS of non-uniform matter at low densities, we perform general-relativistic spherically symmetric simulations of core-collapse SNe with and without neutrino transfer, starting from a presupernova model of 15 solar mass. In the adiabatic simulation without neutrino transfer, the explosion is successful, and the explosion energy with the variational EOS is larger than that with the Shen EOS. In the case of the simulation with neutrino transfer, the shock wave stalls and then the explosion fails, as in other spherically symmetric simulations. The ...

  6. Minimization of the energy loss of nuclear power plants in case of partial in-core monitoring system failure

    Science.gov (United States)

    Zagrebaev, A. M.; Ramazanov, R. N.; Lunegova, E. A.

    2017-01-01

    In this paper we consider the optimization problem minimize of the energy loss of nuclear power plants in case of partial in-core monitoring system failure. It is possible to continuation of reactor operation at reduced power or total replacement of the channel neutron measurements, requiring shutdown of the reactor and the stock of detectors. This article examines the reconstruction of the energy release in the core of a nuclear reactor on the basis of the indications of height sensors. The missing measurement information can be reconstructed by mathematical methods, and replacement of the failed sensors can be avoided. It is suggested that a set of ‘natural’ functions determined by means of statistical estimates obtained from archival data be constructed. The procedure proposed makes it possible to reconstruct the field even with a significant loss of measurement information. Improving the accuracy of the restoration of the neutron flux density in partial loss of measurement information to minimize the stock of necessary components and the associated losses.

  7. Rethinking Sensitivity Analysis of Nuclear Simulations with Topology

    Energy Technology Data Exchange (ETDEWEB)

    Dan Maljovec; Bei Wang; Paul Rosen; Andrea Alfonsi; Giovanni Pastore; Cristian Rabiti; Valerio Pascucci

    2016-01-01

    In nuclear engineering, understanding the safety margins of the nuclear reactor via simulations is arguably of paramount importance in predicting and preventing nuclear accidents. It is therefore crucial to perform sensitivity analysis to understand how changes in the model inputs affect the outputs. Modern nuclear simulation tools rely on numerical representations of the sensitivity information -- inherently lacking in visual encodings -- offering limited effectiveness in communicating and exploring the generated data. In this paper, we design a framework for sensitivity analysis and visualization of multidimensional nuclear simulation data using partition-based, topology-inspired regression models and report on its efficacy. We rely on the established Morse-Smale regression technique, which allows us to partition the domain into monotonic regions where easily interpretable linear models can be used to assess the influence of inputs on the output variability. The underlying computation is augmented with an intuitive and interactive visual design to effectively communicate sensitivity information to the nuclear scientists. Our framework is being deployed into the multi-purpose probabilistic risk assessment and uncertainty quantification framework RAVEN (Reactor Analysis and Virtual Control Environment). We evaluate our framework using an simulation dataset studying nuclear fuel performance.

  8. Analysis of Fracture in Cores from the Tuff Confining Unit beneath Yucca Flat, Nevada Test Site

    Energy Technology Data Exchange (ETDEWEB)

    Lance Prothro

    2008-03-01

    The role fractures play in the movement of groundwater through zeolitic tuffs that form the tuff confining unit (TCU) beneath Yucca Flat, Nevada Test Site, is poorly known. This is an important uncertainty, because beneath most of Yucca Flat the TCU lies between the sources of radionuclide contaminants produced by historic underground nuclear testing and the regional carbonate aquifer. To gain a better understanding of the role fractures play in the movement of groundwater and radionuclides through the TCU beneath Yucca Flat, a fracture analysis focusing on hydraulic properties was performed on conventional cores from four vertical exploratory holes in Area 7 of Yucca Flat that fully penetrate the TCU. The results of this study indicate that the TCU is poorly fractured. Fracture density for all fractures is 0.27 fractures per vertical meter of core. For open fractures, or those observed to have some aperture, the density is only 0.06 fractures per vertical meter of core. Open fractures are characterized by apertures ranging from 0.1 to 10 millimeter, and averaging 1.1 millimeter. Aperture typically occurs as small isolated openings along the fracture, accounting for only 10 percent of the fracture volume, the rest being completely healed by secondary minerals. Zeolite is the most common secondary mineral occurring in 48 percent of the fractures observed.

  9. Multilayer Network Analysis of Nuclear Reactions

    Science.gov (United States)

    Zhu, Liang; Ma, Yu-Gang; Chen, Qu; Han, Ding-Ding

    2016-08-01

    The nuclear reaction network is usually studied via precise calculation of differential equation sets, and much research interest has been focused on the characteristics of nuclides, such as half-life and size limit. In this paper, however, we adopt the methods from both multilayer and reaction networks, and obtain a distinctive view by mapping all the nuclear reactions in JINA REACLIB database into a directed network with 4 layers: neutron, proton, 4He and the remainder. The layer names correspond to reaction types decided by the currency particles consumed. This combined approach reveals that, in the remainder layer, the β-stability has high correlation with node degree difference and overlapping coefficient. Moreover, when reaction rates are considered as node strength, we find that, at lower temperatures, nuclide half-life scales reciprocally with its out-strength. The connection between physical properties and topological characteristics may help to explore the boundary of the nuclide chart.

  10. Multilayer network analysis of nuclear reactions

    CERN Document Server

    Zhu, Liang; Chen, Qu; Han, Ding-Ding

    2016-01-01

    The nuclear reaction network is usually studied via precise calculation of differential equation sets, and much research interest has been focused on the characteristics of nuclides, such as half-life and size limit. In this paper, however, we adopt the methods from both multilayer and reaction networks, and obtain a distinctive view by mapping all the nuclear reactions in JINA REACLIB database into a directed network with 4 layers: neutron, proton, $^4$He and the remainder. The layer names correspond to reaction types decided by the currency particles consumed. This combined approach reveals that, in the remainder layer, the $\\beta$-stability has high correlation with node degree difference and overlapping coefficient. Moreover, when reaction rates are considered as node strength, we find that, at lower temperatures, nuclide half-life scales reciprocally with its out-strength. The connection between physical properties and topological characteristics may help to explore the boundary of the nuclide chart.

  11. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  12. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: pnch@yonsei.ac.kr [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)

    2017-03-15

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  13. Standard test methods for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium metal

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2004-01-01

    1.1 These test methods cover procedures for the chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium metal to determine compliance with specifications.

  14. Pygmy and core polarization dipole modes in 206 Pb: Connecting nuclear structure to stellar nucleosynthesis

    Energy Technology Data Exchange (ETDEWEB)

    Tonchev, A. P.; Tsoneva, N.; Bhatia, C.; Arnold, C. W.; Goriely, S.; Hammond, S. L.; Kelley, J. H.; Kwan, E.; Lenske, H.; Piekarewicz, J.; Raut, R.; Rusev, G.; Shizuma, T.; Tornow, W.

    2017-10-01

    A high-resolution study of the electromagnetic response of 206Pb below the neutron separation energy is performed using a (γ→,γ') experiment at the HIγ→S facility. Nuclear resonance fluorescence with 100% linearly polarized photon beams is used to measure spins, parities, branching ratios, and decay widths of excited states in 206Pb from 4.9 to 8.1 MeV. The extracted ΣB(E1)↑ and ΣB(M1)↑ values for the total electric and magnetic dipole strength below the neutron separation energy are 0.9±0.2e2fm2 and 8.3±2.0μ$2\\atop{N}$, respectively. These measurements are found to be in very good agreement with the predictions from an energy-density functional (EDF) plus quasiparticle phonon model (QPM). Such a detailed theoretical analysis allows to separate the pygmy dipole resonance from both the tail of the giant dipole resonance and multi-phonon excitations. Combined with earlier photonuclear experiments above the neutron separation energy, one extracts a value for the electric dipole polarizability of 206Pb of αD=122±10mb/MeV. When compared to predictions from both the EDF+QPM and accurately calibrated relativistic EDFs, one deduces a range for the neutron-skin thickness of R$206\\atop{skin}$=0.12–0.19fm and a corresponding range for the slope of the symmetry energy of L=48–60MeV. This newly obtained information is also used to estimate the Maxwellian-averaged radiative cross section 205Pb(n,γ)Pb206 at 30 keV to be σ=130±25mb. The astrophysical impact of this measurement—on both the s-process in stellar nucleosynthesis and on the equation of state of neutron-rich matter—is discussed.

  15. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  16. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  17. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the si

  18. Comparative Evaluation of a CORE Based Learning Environment for Nuclear Medicine.

    Science.gov (United States)

    Hogg, Peter; Boyle, Tom; Lawson, Richard

    1999-01-01

    Reports on a comparative assessment of a multimedia learning environment based on a guided discovery approach called CORE (Concept Object Refinement Expression) with two control conditions, lecture and electronic book, in an undergraduate radiography course. Discusses results of qualitative and quantitative measures of effectiveness, pretests and…

  19. Nuclear Cyber Security Case Study and Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sunae [ChungNam National Univ., Daejeon (Korea, Republic of); Kim, Kyung doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Due to the new trend in cyber attacks, there is an increased security threat towards every country's infrastructure. So, security measures are required now than ever before. Previous cyber attacks normal process consists of paralyzing a server function, data extraction, or data control into the IT system for trespassing. However, nowadays control systems and infrastructures are also targeted and attacking methods have changed a lot. These days, the virus is becoming increasingly serious and hacker attacks are also becoming more frequent. This virus is a computer virus produced for the purpose of destroying the infrastructure, such as power plants, airports, railways June 2010, and it was first discovered in Belarus. Israel, the US, and other countries are believed culprits behind Stuxnet attacks on other nations such as Iran. Recent malware distribution, such as website hacking threat is growing. In surveys today one of the most long-term posing security threats is from North Korea. In particular, North Korea has been caught launching ongoing cyber-attacks after their latest nuclear test. South Korea has identified national trends regarding North Korean nuclear tests and analyzed them in order to catch disclosed confidential information. Especially, many nuclear power plants in the world are found to be vulnerable to cyber-attacks. Industrial facilities should be more wary of the risk of a serious cyber attack in the middle is going to increase the reliance on universal and commercial digital systems (off the shelf) software, civilian nuclear infrastructure. Senior executives’ current risk rate levels are increasing. Digitalization of the perception of risk is lacking in nuclear power plants and workers are creating prevention methods to make them fully aware of the risks of cyber-attacks. It is suggested that it may be inappropriate to assume we are prepared for potential attacks. Due to advances in technology, a warning that the growing sense of crisis

  20. Analysis of High Temperature Reactor Control Rod Worth for the Initial and Full Core

    Science.gov (United States)

    Oktajianto, Hammam; Setiawati, Evi; Anam, Khoirul; Sugito, Heri

    2017-01-01

    Control rod is one important component in a nuclear reactor. In nuclear reactor operations the control rod functions to shut down the reactor. This research analyses ten control rods worth of HTR (High Temperature Reactor) at initial and full core. The HTR in this research adopts HTR-10 China and HTR- of pebble bed. Core calculations are performed by using MCNPX code after modelling the entire parts of core in condition of ten control rods fully withdrawn, all control rods in with 20 cm ranges of depth and the use of one control rod. Pebble bed and moderator balls are distributed in the core zone using a Body Centred Cubic (BCC) lattice by ratio of 57:43. The research results are obtained that the use of one control rod will decrease the reactor criticality of 2.04±0.12 %Δk/k at initial core and 1.57±0.10 %Δk/k at full core. The deeper control rods are in, the lesser criticality of reactor is with reactivity of ten control rods of 16.41±0.11 %Δk/k at initial core and 15.43±0.11 %Δk/k at full core. The results show that the use of ten control rods at full core will keep achieving subcritical condition even though the reactivity is smaller than reactivity at initial core.

  1. Reactor Core Scheme for Small Nuclear Power Plant%小型核电站堆芯方案

    Institute of Scientific and Technical Information of China (English)

    解家春; 刘天才

    2012-01-01

    The small nuclear power planl enjoys advantages of long life and passive safely and is an important choice in the future nuclear power development. A conceptual core is designed for the small nuclear power planl. It is a pool-type fast reactor with sodium as coolant, the movable reflector and the fixed absorber as the reactivity control system for long-life. Further calculation results show thai the life of the reactor could be as long as 30 years, with a reasonable power distribution, all the reactivity coefficients negative, enough reactivity control worth, and all parameters satisfy the design requirements.%具有长寿命、非能动安全的小型核电站是核电发展的一个重要方向.本研究设计了一个小型核电站堆芯方案.该方案为池式钠冷快堆,采用移动反射层和堆内固定吸收体实现较长的堆芯寿期.进一步计算表明,该堆芯方案的寿期可达30年,功率分布合理,各种反应性系数为负值,控制方式的价值足够,满足设计要求.

  2. Lack of nuclear clusters in dwarf spheroidal galaxies: implications for massive black holes formation and the cusp/core problem

    Science.gov (United States)

    Arca-Sedda, Manuel; Capuzzo-Dolcetta, Roberto

    2017-01-01

    One of the leading scenarios for the formation of nuclear star clusters in galaxies is related to the orbital decay of globular clusters (GCs) and their subsequent merging, though alternative theories are currently debated. The availability of high-quality data for structural and orbital parameters of GCs allows us to test different nuclear star cluster formation scenarios. The Fornax dwarf spheroidal (dSph) galaxy is the heaviest satellite of the Milky Way and it is the only known dSph hosting five GCs, whereas there are no clear signatures for the presence of a central massive black hole. For this reason, it represents a suited place to study the orbital decay process in dwarf galaxies. In this paper, we model the future evolution of the Fornax GCs simulating them and the host galaxy by means of direct N-body simulations. Our simulations also take into account the gravitational field generated by the Milky Way. We found that if the Fornax galaxy is embedded in a standard cold dark matter halo, the nuclear cluster formation would be significantly hampered by the high central galactic mass density. In this context, we discuss the possibility that infalling GCs drive the flattening of the galactic density profile, giving a possible alternative explanation to the so-called cusp/core problem. Moreover, we briefly discuss the link between GC infall process and the absence of massive black holes in the centre of dSphs.

  3. Thermodynamic evaluation of the solidification phase of molten core-concrete under estimated Fukushima Daiichi nuclear power plant accident conditions

    Science.gov (United States)

    Kitagaki, Toru; Yano, Kimihiko; Ogino, Hideki; Washiya, Tadahiro

    2017-04-01

    The solidification phases of molten core-concrete under the estimated molten core-concrete interaction (MCCI) conditions in the Fukushima Daiichi Nuclear Power Plant Unit 1 were predicted using the thermodynamic equilibrium calculation tool, FactSage 6.2, and the NUCLEA database in order to contribute toward the 1F decommissioning work and to understand the accident progression via the analytical results for the 1F MCCI products. We showed that most of the U and Zr in the molten core-concrete forms (U,Zr)O2 and (Zr,U)SiO4, and the formation of other phases with these elements is limited. However, the formation of (Zr,U)SiO4 requires a relatively long time because it involves a change in the crystal structure from fcc-(U,Zr)O2 to tet-(U,Zr)O2, followed by the formation of (Zr,U)SiO4 by reaction with SiO2. Therefore, the formation of (Zr,U)SiO4 is limited under quenching conditions. Other common phases are the oxide phases, CaAl2Si2O8, SiO2, and CaSiO3, and the metallic phases of the Fe-Si and Fe-Ni alloys. The solidification phenomenon of the crust under quenching conditions and that of the molten pool under thermodynamic equilibrium conditions in the 1F MCCI progression are discussed.

  4. Special Sm core complex functions in assembly of the U2 small nuclear ribonucleoprotein of Trypanosoma brucei.

    Science.gov (United States)

    Preusser, Christian; Palfi, Zsofia; Bindereif, Albrecht

    2009-08-01

    The processing of polycistronic pre-mRNAs in trypanosomes requires the spliceosomal small ribonucleoprotein complexes (snRNPs) U1, U2, U4/U6, U5, and SL, each of which contains a core of seven Sm proteins. Recently we reported the first evidence for a core variation in spliceosomal snRNPs; specifically, in the trypanosome U2 snRNP, two of the canonical Sm proteins, SmB and SmD3, are replaced by two U2-specific Sm proteins, Sm15K and Sm16.5K. Here we identify the U2-specific, nuclear-localized U2B'' protein from Trypanosoma brucei. U2B'' interacts with a second U2 snRNP protein, U2-40K (U2A'), which in turn contacts the U2-specific Sm16.5K/15K subcomplex. Together they form a high-affinity, U2-specific binding complex. This trypanosome-specific assembly differs from the mammalian system and provides a functional role for the Sm core variation found in the trypanosomal U2 snRNP.

  5. Analysis of three-phase power transformer laminated magnetic core designs

    Directory of Open Access Journals (Sweden)

    M.I. Levin

    2014-03-01

    Full Text Available Analysis and research into properties and parameters of different-type laminated magnetic cores of three-phase power transformers are conducted. Most of new laminated magnetic core designs are found to have significant shortcomings resulted from design and technological features of their manufacturing. These shortcomings cause increase in ohmic loss in the magnetic core, which eliminates advantages of the new core configurations and makes them uncompetitive as compared with the classical laminated magnetic core design.

  6. Transient Analysis of Air-Core Coils by Moment Method

    Science.gov (United States)

    Fujita, Akira; Kato, Shohei; Hirai, Takao; Okabe, Shigemitu

    In electric power system a threat of lighting surge is decreased by using ground wire and arrester, but the risk of failure of transformer is still high. Winding is the most familiar conductor configuration of electromagnetic field components such as transformer, resistors, reactance device etc. Therefore, it is important that we invest the lighting surge how to advance into winding, but the electromagnet coupling in a winding makes lighting surge analysis difficult. In this paper we present transient characteristics analysis of an air-core coils by moment method in frequency domain. We calculate the inductance from time response and impedance in low frequency, and compare them with the analytical equation which is based on Nagaoka factor.

  7. Developing engineering design core competences through analysis of industrial products

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp; Lenau, Torben Anker

    2011-01-01

    Most product development work carried out in industrial practice is characterised by being incremental, i.e. the industrial company has had a product in production and on the market for some time, and now time has come to design a new and upgraded variant. This type of redesign project requires...... that the engineering designers have core design competences to carry through an analysis of the existing product encompassing both a user-oriented side and a technical side, as well as to synthesise solution proposals for the new and upgraded product. The authors of this paper see an educational challenge in staging...... a course module, in which students develop knowledge, understanding and skills, which will prepare them for being able to participate in and contribute to redesign projects in industrial practice. In the course module Product Analysis and Redesign that has run for 8 years we have developed and refined...

  8. Nuclear analysis techniques as a component of thermoluminescence dating

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, J.R.; Hutton, J.T.; Habermehl, M.A. [Adelaide Univ., SA (Australia); Van Moort, J. [Tasmania Univ., Sandy Bay, TAS (Australia)

    1996-12-31

    In luminescence dating, an age is found by first measuring dose accumulated since the event being dated, then dividing by the annual dose rate. Analyses of minor and trace elements performed by nuclear techniques have long formed an essential component of dating. Results from some Australian sites are reported to illustrate the application of nuclear techniques of analysis in this context. In particular, a variety of methods for finding dose rates are compared, an example of a site where radioactive disequilibrium is significant and a brief summary is given of a problem which was not resolved by nuclear techniques. 5 refs., 2 tabs.

  9. SOVT analysis of the nuclear industry in Mexico; Analisis FODA de la industria nuclear en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez R, E.; Hernandez B, M. C., E-mail: edelmiraf@yahoo.com [Instituto Tecnologico de Toluca, Division de Estudios de Posgrado, Av. Instituto Tecnologico s/n, Ex-rancho La Virgen, 52140 Metepec, Estado de Mexico (Mexico)

    2011-11-15

    In this work the analysis of strengths, opportunities, vulnerabilities and threats (SOVT) of the nuclear industry in Mexico is presented. This industry presents among its strengths that Mexico is a highly electrified country and has a good established normative mark of nuclear security. Although the Secretaria de Energia in Mexico, with base to the exposed in the Programa Sectorial de Energia 2007-2012, is analyzing the convenience of the generation starting from this source, considering the strong technological dependence of the exterior and the limited federal budget dedicated to this field. As a result of the analysis of the SOVT matrix, were found a great number of strengths that threats, although the vulnerabilities list is major to the strengths, the opportunities list is the bigger. Therefore, the nuclear industry can be a sustainable industry, taking the necessary decisions and taking advantage of the detected opportunities. (Author)

  10. Analysis and prediction of leucine-rich nuclear export signals

    DEFF Research Database (Denmark)

    la Cour, Tanja; Kiemer, Lars; Mølgaard, Anne

    2004-01-01

    We present a thorough analysis of nuclear export signals and a prediction server, which we have made publicly available. The machine learning prediction method is a significant improvement over the generally used consensus patterns. Nuclear export signals (NESs) are extremely important regulators...... of the subcellular location of proteins. This regulation has an impact on transcription and other nuclear processes, which are fundamental to the viability of the cell. NESs are studied in relation to cancer, the cell cycle, cell differentiation and other important aspects of molecular biology. Our conclusion from...... this analysis is that the most important properties of NESs are accessibility and flexibility allowing relevant proteins to interact with the signal. Furthermore, we show that not only the known hydrophobic residues are important in defining a nuclear export signals. We employ both neural networks and hidden...

  11. Subchannel analysis of a small ultra-long cycle fast reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Kim, Ji Hyun; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2014-04-01

    Highlights: • The UCFR-100 is small-sized one of 60 years long-life nuclear reactors without refueling. • The design safety limits of the UCFR-100 are evaluated using MATRA-LMR. • The subchannel results are below the safety limits of general SFR design criteria. - Abstract: Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100 MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498 °C on average and 551 °C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498 °C on average and 548 °C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499 °C on average and 538 °C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria.

  12. Wavelet analysis of the nuclear phase space

    Energy Technology Data Exchange (ETDEWEB)

    Jouault, B.; Sebille, F.; Mota, V. de la

    1997-12-31

    The description of transport phenomena in nuclear matter is addressed in a new approach based on the mathematical theory of wavelets and the projection methods of statistical physics. The advantage of this framework is to offer the opportunity to use information concepts common to both the formulation of physical properties and the mathematical description. This paper focuses on two features, the extraction of relevant informations using the geometrical properties of the underlying phase space and the optimization of the theoretical and numerical treatments based on convenient choices of the representation spaces. (author). 34 refs.

  13. Nuclear power and the public: analysis of collected survey research

    Energy Technology Data Exchange (ETDEWEB)

    Melber, B.D.; Nealey, S.M.; Hammersla, J.; Rankin, W.L.

    1977-11-01

    This executive summary highlights the major findings of a comprehensive synthesis and analysis of over 100 existing surveys dealing with public attitudes toward nuclear power issues. Questions of immediate policy relevance to the nuclear debate are posed and answered on the basis of these major findings. For each issue area, those sections of the report in which more-detailed discussion and presentation of relevant data may be found are indicated.

  14. Status of XSUSA for Sampling Based Nuclear Data Uncertainty and Sensitivity Analysis

    Directory of Open Access Journals (Sweden)

    Pasichnyk I.

    2013-03-01

    Full Text Available In the present contribution, an overview of the sampling based XSUSA method for sensitivity and uncertainty analysis with respect to nuclear data is given. The focus is on recent developments and applications of XSUSA. These applications include calculations for critical assemblies, fuel assembly depletion calculations, and steadystate as well as transient reactor core calculations. The analyses are partially performed in the framework of international benchmark working groups (UACSA – Uncertainty Analyses for Criticality Safety Assessment, UAM – Uncertainty Analysis in Modelling. It is demonstrated that particularly for full-scale reactor calculations the influence of the nuclear data uncertainties on the results can be substantial. For instance, for the radial fission rate distributions of mixed UO2/MOX light water reactor cores, the 2σ uncertainties in the core centre and periphery can reach values exceeding 10%. For a fast transient, the resulting time behaviour of the reactor power was covered by a wide uncertainty band. Overall, the results confirm the necessity of adding systematic uncertainty analyses to best-estimate reactor calculations.

  15. Technical aspects of nuclear microprobe analysis of senile plaques from alzheimer patients

    Science.gov (United States)

    Larsson, N. P.-O.; Tapper, U. A. S.; Sturesson, K.; Odselius, R.; Brun, A.

    1990-04-01

    Alzheimer's disease, a common form of senile dementia, has been proposed to be caused by aluminium. One of the interesting structures to be studied, senile plaque cores in the brain, have centres of only about 10 μm. We have investigated the possibility of applying nuclear microprobes to sections containing senile plaques. An alternative staining procedure, TMToluidin blue staining using a spray technique, is also presented. An outline is given of a procedure for preparing senile plaque specimens for nuclear microprobe analysis. This includes a technique for accurate ion beam positioning, utilizing electron microscopy-grids. The subject may be of general interest since sample preparation is one of the most important aspects in microprobe analysis of biological matter.

  16. An out-of-core thermionic-converter system for nuclear space power.

    Science.gov (United States)

    Breitwieser, R.

    1972-01-01

    Reexamination of designs of nuclear thermionic space power systems with the converter outside the reactor in the perspective of recent advances in heat-transfer methods, materials, converter performance, and radiation design. The 40- to 70-kW(e) power range is treated. The configuration is found to meet the constraints of readily available launch vehicles. It allows for off-design operation including startup, shutdown, and possible emergency conditions; provides tolerance of failure by extensive use of modular, redundant elements; incorporates and uses heat pipes in a fashion that reduces the need for extensive in-pile testing of system components; and uses thermionic converters, nuclear fuel elements, and heat-transfer devices in a geometrical form adapted from existing incore thermionic system designs.

  17. An algorithm for multi-group two-dimensional neutron diffusion kinetics in nuclear reactor cores

    OpenAIRE

    Marcelo Schramm

    2016-01-01

    The objective of this thesis is to introduce a new methodology for two{dimensional multi{ group neutron diffusion kinetics in a reactor core. The presented methodology uses a polyno- mial approximation in a rectangular homogeneous domain with non{homogeneous boundary conditions. As it consists on a truncated Taylor series, its error estimates varies with the size of the rectangle. The coefficients are obtained mainly by their relations with the independent term, which is determined by the dif...

  18. Nuclear structure calculations in $^{20}$Ne with No-Core Configuration-Interaction model

    CERN Document Server

    Konieczka, Maciej

    2016-01-01

    Negative parity states in $^{20}$Ne and Gamow-Teller strength distribution for the ground-state beta-decay of $^{20}$Na are calculated for the very first time using recently developed No-Core Configuration-Interaction model. The approach is based on multi-reference density functional theory involving isospin and angular-momentum projections. Advantages and shortcomings of the method are briefly discussed.

  19. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  20. Neutron stars with hyperon cores: stellar radii and EOS near nuclear density

    CERN Document Server

    Fortin, M; Haensel, P; Bejger, M

    2014-01-01

    The existence of 2 Msun pulsars puts very strong constraints on the equation of state (EOS) of neutron stars (NSs) with hyperon cores, which can be satisfied only by special models of hadronic matter. The radius-mass relation for these models is so specific that it could be submitted to an observational test with forthcoming X-ray observatories. We want to study the impact of the presence of hyperon cores on the radius-mass relation for NS. We aim at finding how, and for which particular stellar mass range, a specific relation R(M), where M is gravitational mass, and R is radius, is associated with the presence of an hyperon core. We consider a large set of theoretical EOSs of dense matter, based on the relativistic mean-field (RMF) approximation, allowing for the presence of hyperons in NSs. We seek for correlations between R(M) and the stiffness of the EOS below the hyperon threshold, needed to pass the 2 Msun test. For NS masses 1.013km, which is due to a very stiff pre-hyperon segment of the EOS. At nucle...

  1. Ion beam analysis - development and application of nuclear reaction analysis methods, in particular at a nuclear microprobe

    Energy Technology Data Exchange (ETDEWEB)

    Sjoeland, K.A.

    1996-11-01

    This thesis treats the development of Ion Beam Analysis methods, principally for the analysis of light elements at a nuclear microprobe. The light elements in this context are defined as having an atomic number less than approx. 13. The work reported is to a large extent based on multiparameter methods. Several signals are recorded simultaneously, and the data can be effectively analyzed to reveal structures that can not be observed through one-parameter collection. The different techniques are combined in a new set-up at the Lund Nuclear Microprobe. The various detectors for reaction products are arranged in such a way that they can be used for the simultaneous analysis of hydrogen, lithium, boron and fluorine together with traditional PIXE analysis and Scanning Transmission Ion Microscopy as well as photon-tagged Nuclear Reaction Analysis. 48 refs.

  2. CIVIL LIABILITY FOR NUCLEAR DAMAGE: COMPARATIVE ANALYSIS OF INTERNATIONAL TREATIES

    Directory of Open Access Journals (Sweden)

    Laura Rimšaitė

    2013-06-01

    Full Text Available It was widely accepted that nuclear damage might be extensive and spread to other countries. International civil liability for nuclear damage is embodied by two major instruments: International Atomic Energy Agency (IAEA 1963 Vienna Convention on Civil liability for Nuclear Damage and Paris Convention of 1960 on third party liability (OECD with its amending protocols. Major problem arises because of lack of coherence and for this reason supplementary conventions and protocols has been adopted but sufficient results has not been achieved. International treaties on civil liability for nuclear damage are mostly based upon principles of operator’s exclusive, channeling, strict liability for nuclear damage, mandatory financial coverage, compensation without discrimination. These principles set ground for the appropriate compensation standard thus minimizing the difficulty level of complicated legal cross-actions and identifies certain subjects in individual cases who are liable also allows a concentration of the insurance capacity. Although Conventions sets similar principles, Europe remains in two different liability regimes which cover differences of liability amounts, scope of application, rules of jurisdiction conflicts. Problem of legal coherence at European Union level also arises because Member States are either parties to the Paris Convention or Vienna Convention at different speeds. This research paper provides an in-depth analysis of international legal framework development and impetus to create trans-boundary compensation mechanisms thus to foster development of European Union nuclear energy market and to provide higher protection for victims inside and outside the country where the incident has occurred. Purpose – provide comparative analysis of international treaties which regulate civil liability for nuclear damage in the context of European Union nuclear energy market development. Design/methodology - paper is based on document

  3. Analysis of the transient response of nuclear spins in GaAs with/without nuclear magnetic resonance

    Science.gov (United States)

    Rasly, Mahmoud; Lin, Zhichao; Yamamoto, Masafumi; Uemura, Tetsuya

    2016-05-01

    As an alternative to studying the steady-state responses of nuclear spins in solid state systems, working within a transient-state framework can reveal interesting phenomena. The response of nuclear spins in GaAs to a changing magnetic field was analyzed based on the time evolution of nuclear spin temperature. Simulation results well reproduced our experimental results for the transient oblique Hanle signals observed in an all-electrical spin injection device. The analysis showed that the so called dynamic nuclear polarization can be treated as a cooling tool for the nuclear spins: It works as a provider to exchange spin angular momentum between polarized electron spins and nuclear spins through the hyperfine interaction, leading to an increase in the nuclear polarization. In addition, a time-delay of the nuclear spin temperature with a fast sweep of the external magnetic field produces a possible transient state for the nuclear spin polarization. On the other hand, the nuclear magnetic resonance acts as a heating tool for a nuclear spin system. This causes the nuclear spin temperature to jump to infinity: i.e., the average nuclear spins along with the nuclear field vanish at resonant fields of 75As, 69Ga and 71Ga, showing an interesting step-dip structure in the oblique Hanle signals. These analyses provide a quantitative understanding of nuclear spin dynamics in semiconductors for application in future computation processing.

  4. Analysis of the transient response of nuclear spins in GaAs with/without nuclear magnetic resonance

    Directory of Open Access Journals (Sweden)

    Mahmoud Rasly

    2016-05-01

    Full Text Available As an alternative to studying the steady-state responses of nuclear spins in solid state systems, working within a transient-state framework can reveal interesting phenomena. The response of nuclear spins in GaAs to a changing magnetic field was analyzed based on the time evolution of nuclear spin temperature. Simulation results well reproduced our experimental results for the transient oblique Hanle signals observed in an all-electrical spin injection device. The analysis showed that the so called dynamic nuclear polarization can be treated as a cooling tool for the nuclear spins: It works as a provider to exchange spin angular momentum between polarized electron spins and nuclear spins through the hyperfine interaction, leading to an increase in the nuclear polarization. In addition, a time-delay of the nuclear spin temperature with a fast sweep of the external magnetic field produces a possible transient state for the nuclear spin polarization. On the other hand, the nuclear magnetic resonance acts as a heating tool for a nuclear spin system. This causes the nuclear spin temperature to jump to infinity: i.e., the average nuclear spins along with the nuclear field vanish at resonant fields of 75As, 69Ga and 71Ga, showing an interesting step-dip structure in the oblique Hanle signals. These analyses provide a quantitative understanding of nuclear spin dynamics in semiconductors for application in future computation processing.

  5. Core Stability in Athletes: A Critical Analysis of Current Guidelines.

    Science.gov (United States)

    Wirth, Klaus; Hartmann, Hagen; Mickel, Christoph; Szilvas, Elena; Keiner, Michael; Sander, Andre

    2017-03-01

    Over the last two decades, exercise of the core muscles has gained major interest in professional sports. Research has focused on injury prevention and increasing athletic performance. We analyzed the guidelines for so-called functional strength training for back pain prevention and found that programs were similar to those for back pain rehabilitation; even the arguments were identical. Surprisingly, most exercise specifications have neither been tested for their effectiveness nor compared with the load specifications normally used for strength training. Analysis of the scientific literature on core stability exercises shows that adaptations in the central nervous system (voluntary activation of trunk muscles) have been used to justify exercise guidelines. Adaptations of morphological structures, important for the stability of the trunk and therefore the athlete's health, have not been adequately addressed in experimental studies or in reviews. In this article, we explain why the guidelines created for back pain rehabilitation are insufficient for strength training in professional athletes. We critically analyze common concepts such as 'selective activation' and training on unstable surfaces.

  6. A Metropolis algorithm combined with Nelder-Mead Simplex applied to nuclear reactor core design

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F. [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil)], E-mail: wfsacco@iprj.uerj.br; Filho, Hermes Alves; Henderson, Nelio [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil); Oliveira, Cassiano R.E. de [Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

    2008-05-15

    A hybridization of the recently introduced Particle Collision Algorithm (PCA) and the Nelder-Mead Simplex algorithm is introduced and applied to a core design optimization problem which was previously attacked by other metaheuristics. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a three-enrichment-zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. The new metaheuristic performs better than the genetic algorithm, particle swarm optimization, and the Metropolis algorithms PCA and the Great Deluge Algorithm, thus demonstrating its potential for other applications.

  7. Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, F. [Saclay Center CEA, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

    2011-07-01

    The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

  8. Nuclear techniques of analysis in diamond synthesis and annealing

    Energy Technology Data Exchange (ETDEWEB)

    Jamieson, D. N.; Prawer, S.; Gonon, P.; Walker, R.; Dooley, S.; Bettiol, A.; Pearce, J. [Melbourne Univ., Parkville, VIC (Australia). School of Physics

    1996-12-31

    Nuclear techniques of analysis have played an important role in the study of synthetic and laser annealed diamond. These measurements have mainly used ion beam analysis with a focused MeV ion beam in a nuclear microprobe system. A variety of techniques have been employed. One of the most important is nuclear elastic scattering, sometimes called non-Rutherford scattering, which has been used to accurately characterise diamond films for thickness and composition. This is possible by the use of a database of measured scattering cross sections. Recently, this work has been extended and nuclear elastic scattering cross sections for both natural boron isotopes have been measured. For radiation damaged diamond, a focused laser annealing scheme has been developed which produces near complete regrowth of MeV phosphorus implanted diamonds. In the laser annealed regions, proton induced x-ray emission has been used to show that 50 % of the P atoms occupy lattice sites. This opens the way to produce n-type diamond for microelectronic device applications. All these analytical applications utilize a focused MeV microbeam which is ideally suited for diamond analysis. This presentation reviews these applications, as well as the technology of nuclear techniques of analysis for diamond with a focused beam. 9 refs., 6 figs.

  9. Fully encapsulated directional self-powered gamma ray detector for use in in-core nuclear reactor measurements

    Energy Technology Data Exchange (ETDEWEB)

    LeVert, F.E.; Cox, S.A.

    1979-01-01

    A study of a fully encapsulated directional self-powered gamma ray detector designed for localized in core measurements in a nuclear reactor was conducted. The detector consisted of a multilayer arrangement of a metal-dielectric-metal-dielectric-metal structure. The dielectric material was made of two plates of unequal thicknesses which were placed on opposite sides of the central metal plate. The direction discrimination exhibited by the detector was attributed to the combined effect of electron ranges, Photo-Compton electron generation rates, and the presence of E-fields in the unequal thicknesses of dielectric material. Results showing the response of the detector when it was placed in a gamma ray field with a known anisotropic component are presented.

  10. Constraining the symmetry energy content of nuclear matter from nuclear masses: a covariance analysis

    CERN Document Server

    Mondal, C; De, J N

    2015-01-01

    Elements of nuclear symmetry energy evaluated from different energy density functionals parametrized by fitting selective bulk properties of few representative nuclei are seen to vary widely. Those obtained from experimental data on nuclear masses across the periodic table, however, show that they are better constrained. A possible direction in reconciling this paradox may be gleaned from comparison of results obtained from use of the binding energies in the fitting protocol within a microscopic model with two sets of nuclei, one a representative standard set and another where very highly asymmetric nuclei are additionally included. A covariance analysis reveals that the additional fitting protocol reduces the uncertainties in the nuclear symmetry energy coefficient, its slope parameter as well as the neutron-skin thickness in $^{208}$Pb nucleus by $\\sim 50\\%$. The central values of these entities are also seen to be slightly reduced.

  11. Processing and geologic analysis of conventional cores from well ER-20-6 No. 1, Nevada Test Site

    Energy Technology Data Exchange (ETDEWEB)

    Prothro, L.B., Townsend, M.J.; Drellack, S.L. Jr. [and others

    1997-09-01

    In 1996, Well Cluster ER-20-6 was drilled on Pahute Mesa in Area 20, in the northwestern corner of the Nevada Test Site (NTS). The three wells of the cluster are located from 166 to 296 meters (m) (544 to 971 feet [ft]) southwest of the site of the underground nuclear test code-named BULLION, conducted in 1990 in Emplacement Hole U-20bd. The well cluster was planned to be the site of a forced-gradient experiment designed to investigate radionuclide transport in groundwater. To obtain additional information on the occurrence of radionuclides, nature of fractures, and lithology, a portion of Well ER-20-6 No. 1, the hole closest to the explosion cavity, was cored for later analysis. Bechtel Nevada (BN) geologists originally prepared the geologic interpretation of the Well Cluster ER-20-6 site and documented the geology of each well in the cluster. However, the cores from Well ER-20-6 No. 1 were not accessible at the time of that work. As the forced-gradient experiment and other radio nuclide migration studies associated with the well cluster progressed, it was deemed appropriate to open the cores, describe the geology, and re-package the core for long-term air-tight storage. This report documents and describes the processing, geologic analysis, and preservation of the conventional cores from Well ER20-6 No. 1.

  12. Exploring the nuclear pasta phase in core-collapse supernova matter.

    Science.gov (United States)

    Pais, Helena; Stone, Jirina R

    2012-10-12

    The core-collapse supernova phenomenon, one of the most explosive events in the Universe, presents a challenge to theoretical astrophysics. Of the large variety of forms of matter present in core-collapse supernova, we focus on the transitional region between homogeneous (uniform) and inhomogeneous (pasta) phases. A three-dimensional, finite temperature Skyrme-Hartree-Fock (3D-SHF)+BCS calculation yields, for the first time fully self-consistently, the critical density and temperature of both the onset of the pasta in inhomogeneous matter, consisting of neutron-rich heavy nuclei and a free neutron and electron gas, and its dissolution to a homogeneous neutron, proton, and electron liquid. We also identify density regions for different pasta formations between the two limits. We employ four different forms of the Skyrme interaction, SkM*, SLy4, NRAPR, and SQMC700 and find subtle variations in the low density and high density transitions into and out of the pasta phase. One new stable pasta shape has been identified, in addition to the classic ones, on the grid of densities and temperatures used in this work. Our results are critically compared to recent calculations of pasta formation in the quantum molecular dynamics approach and Thomas-Fermi and coexisting phase approximations to relativistic mean-field models.

  13. Comparison between different flux traps assembled in the core of the nuclear reactor IPEN/MB-01 by measuring of the thermal and epithermal neutron fluxes using activation foils

    Energy Technology Data Exchange (ETDEWEB)

    Mura, Luiz Ernesto Credidio; Bitelli, Ulysses d' Utra; Mura, Luis Felipe Liambos; Carluccio, Thiago; Andrade, Graciete Simoes de, E-mail: ubitelli@ipen.b, E-mail: gsasilva@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The production of radioisotopes is one of the most important applications of nuclear research reactors. This study investigated a method called Flux Trap, which is used to increase the yield of production of radioisotopes in nuclear reactors. The method consists in the rearrangement of the fuel rods to allow the increase of the thermal neutron flux in the irradiation region inside the reactor core, without changing the standard reactor power level. Various configurations were assembled with the objective of finding the configuration with the highest thermal neutron flux in the region of irradiation. The method of activation analysis was used to measure the thermal neutron flux and determine the most efficient reactor core configuration . It was found that there was an increase in the thermal neutron flux of 337% in the most efficient configuration, which demonstrates the effectiveness of the method. (author)

  14. Convergence Nanorobot Analysis for Radiation Therapy-Industrial Innovations in Nuclear Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Taeho [Yonsei Univ., Wonju (Korea, Republic of)

    2015-10-15

    The important step of the commercialization is the make the prototype nanorobot where lots of applications could be introduced for the industry. For the much more advanced operations of the nanorobot, it is needed to imagine the strategy for the operation in the non-regular shaped organs like the lung which shows the different feature following breaths. The biological stuffs are usually in the irregular shape and could be changed by the external force or the infected viruses. The biological substance could be made by the amorphous material which is used frequently in the industry. The antibody reaction is a particular matter which could be happen in the human body. So, the adaptations of the nanorobot could be increased for the practical purposed. Fig. 7 is the newly imagined convergence nuclear technology with nanorobotics for nuclear engineering fields in which many kinds of applications are imagined. Following the new applications of the nanorobot, it is possible to challenge for the difficult matters in the conventional nuclear industry. Fig. 8 shows the historic mistakes in commercialized nuclear power plants (NPPs) considering the nuclear reactor analysis and safety system induced by the accident. Firstly, the non-matched flux shapes made by the multiplications of Bessel function and cosine function by the cylindrical core shape, which is different from the spherical or rectangular core shape, couldn't describe the exact flux shape. Secondly, the safety system installed to start in the accident is the piping-based injection equipment. However, the safety injection systems have failed in three major sever accidents as Three Mile Island (TMI), Chernobyl, and Fukushima cases due to the significant piping failures.

  15. Mass Defect from Nuclear Physics to Mass Spectral Analysis

    Science.gov (United States)

    Pourshahian, Soheil

    2017-09-01

    Mass defect is associated with the binding energy of the nucleus. It is a fundamental property of the nucleus and the principle behind nuclear energy. Mass defect has also entered into the mass spectrometry terminology with the availability of high resolution mass spectrometry and has found application in mass spectral analysis. In this application, isobaric masses are differentiated and identified by their mass defect. What is the relationship between nuclear mass defect and mass defect used in mass spectral analysis, and are they the same? [Figure not available: see fulltext.

  16. The Development of Nufreq-N AN Analytical Model for the Stability Analysis of Nuclear Coupled Density-Wave Oscillations in Boiling Water Nuclear Reactors.

    Science.gov (United States)

    Park, Goon Cherl

    A state-of-the-art one-dimensional thermal-hydraulic model has been developed to be used for the linear analysis of nuclear-coupled density-wave oscillations in a boiling water nuclear reactor (BWR). The model accounts for phasic slip, distributed spacers, subcooled boiling, space/time -dependent power distributions and distributed heated wall dynamics. In addition to a parallel channel stability analysis, a detailed model was derived for the BWR loop analysis of both the natural and forced circulation modes of operation. In its final form, this model constitutes a multi -input, multi-output(MIMO) linear system, which features a general nodal neutron kinetics model. Kinetics parameters for use in the kinetics model have been obtained by utilizing self-consistent nodal data and power distributions. The stability characteristics of a typical BWR/4 has been investigated with the Nyquist criterion. The computer implementation of this model, NUFREQ -N, was used for the parametric study of a typical BWR/4 and comparisons were made with existing in-core and out -of-core data. Also, NUFREQ-N was used to analyze the expected stability characteristics of a typical BWR/4. The parametric results revealed important factors influencing BWR stability margin. It was found that NUFREQ -N generally agreed well with out-of-core data. This was especially true for the predicted power-to-flow transfer function, which is the most important transfer function in thermal-hydraulic stability analysis. In the stability analysis of a typical BWR/4 it was found that it is very important to use accurate models of thermal-hydraulic and neutron kinetic phenomena. Moreover, the accuracy of the nuclear input data is extremely important.

  17. Safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

    2002-04-01

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term nuclear R and D Program. In this report, key safety design features are described and safety analyses results for typical ATWS accidents in the KALIMER design with breakeven core are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the event categorization and acceptance criteria for the KALIMER safety analysis are described in chapter 2. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER system response to the events. They are categorized as Bounding Events (BEs) because of their low probability of occurrence. In chapter 4, the performance analysis results of the KALIMER containment dome are described along with the HCDA accident scenario and source terms. The major containment parameters of peak pressure and peak temperature have been calculated using the CONTAIN-LMR code. Radiological consequence has been evaluated by the MACCS code. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using SCHAMBETA code developed in the framework of the modified bethe-tait method. Work energy potentials based arising from the sodium expansion as well as the isentropic fuel expansion are then calculated to evaluate the structural integrity of the reactor vessel, reactor internals and primary coolant system of KALIMER.

  18. Next Generation Nuclear Plant GAP Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    Ball, Sydney J [ORNL; Burchell, Timothy D [ORNL; Corwin, William R [ORNL; Fisher, Stephen Eugene [ORNL; Forsberg, Charles W. [Massachusetts Institute of Technology (MIT); Morris, Robert Noel [ORNL; Moses, David Lewis [ORNL

    2008-12-01

    As a follow-up to the phenomena identification and ranking table (PIRT) studies conducted recently by NRC on next generation nuclear plant (NGNP) safety, a study was conducted to identify the significant 'gaps' between what is needed and what is already available to adequately assess NGNP safety characteristics. The PIRT studies focused on identifying important phenomena affecting NGNP plant behavior, while the gap study gives more attention to off-normal behavior, uncertainties, and event probabilities under both normal operation and postulated accident conditions. Hence, this process also involved incorporating more detailed evaluations of accident sequences and risk assessments. This study considers thermal-fluid and neutronic behavior under both normal and postulated accident conditions, fission product transport (FPT), high-temperature metals, and graphite behavior and their effects on safety. In addition, safety issues related to coupling process heat (hydrogen production) systems to the reactor are addressed, given the limited design information currently available. Recommendations for further study, including analytical methods development and experimental needs, are presented as appropriate in each of these areas.

  19. Development of Reactor Core for Nuclear Thermal Propulsion%核热推进堆芯方案的发展

    Institute of Scientific and Technical Information of China (English)

    解家春; 赵守智

    2012-01-01

    Nuclear thermal propulsion heats propellant with fission energy. It's specific impulse is double of chemical rockets. It could play an important role in space mission. During the research process about nuclear thermal propulsion in USA and Russia, many reactors were well developed. The details of the reactors core were described, the characteristics of design were indicated, and the trend of development was summarized.%核热推进利用核裂变能加热工质,比冲可达化学火箭的2倍多,在空间活动中有广阔的应用前景.在美国和俄罗斯的研究过程中,对多个核热推进堆芯方案进行了较深入的研究.本工作介绍了这些堆芯方案的情况,详细说明了其设计特点,并总结了堆芯方案的发展趋势.

  20. Computer code system for the R and D of nuclear fuel cycle with fast reactor. 5. Development and application of reactor analysis code system

    Energy Technology Data Exchange (ETDEWEB)

    Yokoyama, Kenji; Hazama, Taira; Chiba, Go; Ohki, Shigeo; Ishikawa, Makoto [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-12-01

    In the core design of fast reactors (FRs), it is very important to improve the prediction accuracy of the nuclear characteristics for both reducing cost and ensuring reliability of FR plants. A nuclear reactor analysis code system for FRs has been developed by the Japan Nuclear Cycle Development Institute (JNC). This paper describes the outline of the calculation models and methods in the system consisting of several analysis codes, such as the cell calculation code CASUP, the core calculation code TRITAC and the sensitivity analysis code SAGEP. Some examples of verification results and improvement of the design accuracy are also introduced based on the measurement data from critical assemblies, e.g, the JUPITER experiment (USA/Japan), FCA (Japan), MASURCA (France), and BFS (Russia). Furthermore, application fields and future plans, such as the development of new generation nuclear constants and applications to MA{center_dot}FP transmutation, are described. (author)

  1. Detailed 3-D nuclear analysis of ITER blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, T.D., E-mail: tdbohm@wisc.edu [University of Wisconsin-Madison, Madison, WI (United States); Sawan, M.E.; Marriott, E.P.; Wilson, P.P.H. [University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, M.; Bullock, J. [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-10-15

    In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.

  2. Deconvolution-based resolution enhancement of chemical ice core records obtained by continuous flow analysis

    DEFF Research Database (Denmark)

    Rasmussen, Sune Olander; Andersen, Katrine K.; Johnsen, Sigfus Johann;

    2005-01-01

    Continuous flow analysis (CFA) has become a popular measuring technique for obtaining high-resolution chemical ice core records due to an attractive combination of measuring speed and resolution. However, when analyzing the deeper sections of ice cores or cores from low-accumulation areas, there ...

  3. Technique for continuous high-resolution analysis of trace substances in firn and ice cores

    Energy Technology Data Exchange (ETDEWEB)

    Roethlisberger, R.; Bigler, M.; Hutterli, M.; Sommer, S.; Stauffer, B.; Junghans, H.G.; Wagenbach, D.

    2000-01-15

    The very successful application of a CFA (Continuous flow analysis) system in the GRIP project (Greenland Ice Core Project) for high-resolution ammonium, calcium, hydrogen peroxide, and formaldehyde measurements along a deep ice core led to further development of this analysis technique. The authors included methods for continuous analysis technique. The authors included methods for continuous analysis of sodium, nitrate, sulfate, and electrolytical conductivity, while the existing methods have been improved. The melting device has been optimized to allow the simultaneous analysis of eight components. Furthermore, a new melter was developed for analyzing firn cores. The system has been used in the frame of the European Project for Ice Coring in Antarctica (EPICA) for in-situ analysis of several firn cores from Dronning Maud Land, Antarctica, and for the new ice core drilled at Dome C, Antarctica.

  4. Nuclear Forensics: Scientific Analysis Supporting Law Enforcement and Nuclear Security Investigations.

    Science.gov (United States)

    Keegan, Elizabeth; Kristo, Michael J; Toole, Kaitlyn; Kips, Ruth; Young, Emma

    2016-02-02

    Nuclear forensic science, or "nuclear forensic", aims to answer questions about nuclear material found outside of regulatory control. In this Feature, we provide a general overview of nuclear forensics, selecting examples of key "nuclear forensic signatures" which have allowed investigators to determine the identity of unknown nuclear material in real investigations.

  5. Full core analysis of IRIS reactor by using MCNPX.

    Science.gov (United States)

    Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S

    2016-07-01

    This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code.

  6. Earth's core and inner-core resonances from analysis of VLBI nutation and superconducting gravimeter data

    Science.gov (United States)

    Rosat, S.; Lambert, S. B.; Gattano, C.; Calvo, M.

    2017-01-01

    Geophysical parameters of the deep Earth's interior can be evaluated through the resonance effects associated with the core and inner-core wobbles on the forced nutations of the Earth's figure axis, as observed by very long baseline interferometry (VLBI), or on the diurnal tidal waves, retrieved from the time-varying surface gravity recorded by superconducting gravimeters (SGs). In this paper, we inverse for the rotational mode parameters from both techniques to retrieve geophysical parameters of the deep Earth. We analyse surface gravity data from 15 SG stations and VLBI delays accumulated over the last 35 yr. We show existing correlations between several basic Earth parameters and then decide to inverse for the rotational modes parameters. We employ a Bayesian inversion based on the Metropolis-Hastings algorithm with a Markov-chain Monte Carlo method. We obtain estimates of the free core nutation resonant period and quality factor that are consistent for both techniques. We also attempt an inversion for the free inner-core nutation (FICN) resonant period from gravity data. The most probable solution gives a period close to the annual prograde term (or S1 tide). However the 95 per cent confidence interval extends the possible values between roughly 28 and 725 d for gravity, and from 362 to 414 d from nutation data, depending on the prior bounds. The precisions of the estimated long-period nutation and respective small diurnal tidal constituents are hence not accurate enough for a correct determination of the FICN complex frequency.

  7. Genome-wide analysis of core promoter elements from conserved human and mouse orthologous pairs

    OpenAIRE

    Jin, Victor X.; Singer, Gregory AC; Agosto-Pérez, Francisco J; Liyanarachchi, Sandya; Davuluri, Ramana V.

    2006-01-01

    Background The canonical core promoter elements consist of the TATA box, initiator (Inr), downstream core promoter element (DPE), TFIIB recognition element (BRE) and the newly-discovered motif 10 element (MTE). The motifs for these core promoter elements are highly degenerate, which tends to lead to a high false discovery rate when attempting to detect them in promoter sequences. Results In this study, we have performed the first analysis of these core promoter elements in orthologous mouse a...

  8. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  9. Nuclear data uncertainty analysis on a minor actinide burner for transmuting spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hangbok

    1998-08-01

    A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWt minor actinides burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities of the performance parameters were generated using depletion perturbation methods for the constrained close fuel cycle of the reactor. The uncertainty analysis was performed using the sensitivity and covariance data taken from ENDF-B/V and other published sources. The uncertainty analysis of a liquid metal reactor for burning minor actinide has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180 %, 97 %, and 46 %, respectively. An analysis was performed to prioritize the minor actinide reactions for reducing the uncertainties. (author). 41 refs., 17 tabs., 1 fig.

  10. Nuclear-Renewable Hybrid Energy System Market Analysis Plans

    Energy Technology Data Exchange (ETDEWEB)

    Ruth, Mark

    2016-06-09

    This presentation describes nuclear-renewable hybrid energy systems (N-R HESs), states their potential benefits, provides figures for the four tightly coupled N-R HESs that NREL is currently analyzing, and outlines the analysis process that is underway.

  11. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

    Science.gov (United States)

    Adam, Zachary R

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  12. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis

    Science.gov (United States)

    Adam, Zachary R.

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 105-106 years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  13. Nuclear risk analysis of the Ulysses mission

    Energy Technology Data Exchange (ETDEWEB)

    Bartram, B.W.; Vaughan, F.R. (NUS Corporation, 910 Clopper Road, Gaithersburg, Maryland 20877-0962 (USA)); Englehart, D.R.W. (Office of New Production Reactors, U.S. Department of Energy, Washington, D.C. 20585 (USA))

    1991-01-01

    The use of a radioisotope thermoelectric generator fueled with plutonium-238 dioxide on the Space Shuttle-launched Ulysses mission implies some level of risk due to potential accidents. This paper describes the method used to quantify risks in the Ulysses mission Final Safety Analysis Report prepared for the U.S. Department of Energy. The starting point for the analysis described herein is following input of source term probability distributions from the General Electric Company. A Monte Carlo technique is used to develop probability distributions of radiological consequences for a range of accident scenarios thoughout the mission. Factors affecting radiological consequences are identified, the probability distribution of the effect of each factor determined, and the functional relationship among all the factors established. The probability distributions of all the factor effects are then combined using a Monte Carlo technique. The results of the analysis are presented in terms of complementary cumulative distribution functions (CCDF) by mission sub-phase, phase, and the overall mission. The CCDFs show the total probability that consequences (calculated health effects) would be equal to or greater than a given value.

  14. Buckling and dynamic analysis of drill strings for core sampling

    Energy Technology Data Exchange (ETDEWEB)

    Ziada, H.H., Westinghouse Hanford

    1996-05-15

    This supporting document presents buckling and dynamic stability analyses of the drill strings used for core sampling. The results of the drill string analyses provide limiting operating axial loads and rotational speeds to prevent drill string failure, instability and drill bit overheating during core sampling. The recommended loads and speeds provide controls necessary for Tank Waste Remediation System (TWRS) programmatic field operations.

  15. Development and analysis of U-core switched reluctance machine

    DEFF Research Database (Denmark)

    Jæger, Rasmus; Nielsen, Simon Staal; Rasmussen, Peter Omand

    2016-01-01

    these disadvantages have been presented, but not all of them have been demonstrated practically. This paper presents a practical demonstration and assessment of a segmented U-core SRM, which copes with some of the disadvantages of the regular SRM. The U-core SRM has a segmented stator, with a short flux path...

  16. Self-Healing Many-Core Architecture: Analysis and Evaluation

    Directory of Open Access Journals (Sweden)

    Arezoo Kamran

    2016-01-01

    Full Text Available More pronounced aging effects, more frequent early-life failures, and incomplete testing and verification processes due to time-to-market pressure in new fabrication technologies impose reliability challenges on forthcoming systems. A promising solution to these reliability challenges is self-test and self-reconfiguration with no or limited external control. In this work a scalable self-test mechanism for periodic online testing of many-core processor has been proposed. This test mechanism facilitates autonomous detection and omission of faulty cores and makes graceful degradation of the many-core architecture possible. Several test components are incorporated in the many-core architecture that distribute test stimuli, suspend normal operation of individual processing cores, apply test, and detect faulty cores. Test is performed concurrently with the system normal operation without any noticeable downtime at the application level. Experimental results show that the proposed test architecture is extensively scalable in terms of hardware overhead and performance overhead that makes it applicable to many-cores with more than a thousand processing cores.

  17. Study of core support barrel vibration monitoring using ex-core neutron noise analysis and fuzzy logic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Robby; Song, Seon Ho [Nuclear and Quantum Engineering Department, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kang, Hyun Gook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-03-15

    The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

  18. Data analysis techniques for nuclear and particle physicists

    CERN Document Server

    Pruneau, Claude

    2017-01-01

    This is an advanced data analysis textbook for scientists specializing in the areas of particle physics, nuclear physics, and related subfields. As a practical guide for robust, comprehensive data analysis, it focuses on realistic techniques to explain instrumental effects. The topics are relevant for engineers, scientists, and astroscientists working in the fields of geophysics, chemistry, and the physical sciences. The book serves as a reference for more senior scientists while being eminently accessible to advanced undergraduate and graduate students.

  19. Nuclear Magnetic Resonance Strategies for Metabolic Analysis.

    Science.gov (United States)

    Heude, Clement; Nath, Jay; Carrigan, John Bosco; Ludwig, Christian

    2017-01-01

    NMR spectroscopy is a powerful tool for metabolomic studies, offering highly reproducible and quantitative analyses. This burgeoning field of NMR metabolomics has been greatly aided by the development of modern spectrometers and software, allowing high-throughput analysis with near real-time feedback. Whilst one-dimensional proton (1D-(1)H) NMR analysis is best described and remains most widely used, a plethora of alternative NMR techniques are now available that offer additional chemical and structural information and resolve many of the limitations of conventional 1D-(1)H NMR such as spectral overlay. In this book chapter, we review the principal concepts of practical NMR spectroscopy, from common sample preparation protocols to the benefits and theoretical concepts underpinning the commonly used pulse sequences. Finally, as a case study to highlight the utility of NMR as a method for metabolomic investigation, we have detailed how NMR has been used to gain valuable insight into the metabolism occurring in kidneys prior to transplantation and the potential implications of this.

  20. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  1. SAMPSON Parallel Computation for Sensitivity Analysis of TEPCO's Fukushima Daiichi Nuclear Power Plant Accident

    Science.gov (United States)

    Pellegrini, M.; Bautista Gomez, L.; Maruyama, N.; Naitoh, M.; Matsuoka, S.; Cappello, F.

    2014-06-01

    On March 11th 2011 a high magnitude earthquake and consequent tsunami struck the east coast of Japan, resulting in a nuclear accident unprecedented in time and extents. After scram started at all power stations affected by the earthquake, diesel generators began operation as designed until tsunami waves reached the power plants located on the east coast. This had a catastrophic impact on the availability of plant safety systems at TEPCO's Fukushima Daiichi, leading to the condition of station black-out from unit 1 to 3. In this article the accident scenario is studied with the SAMPSON code. SAMPSON is a severe accident computer code composed of hierarchical modules to account for the diverse physics involved in the various phases of the accident evolution. A preliminary parallelization analysis of the code was performed using state-of-the-art tools and we demonstrate how this work can be beneficial to the nuclear safety analysis. This paper shows that inter-module parallelization can reduce the time to solution by more than 20%. Furthermore, the parallel code was applied to a sensitivity study for the alternative water injection into TEPCO's Fukushima Daiichi unit 3. Results show that the core melting progression is extremely sensitive to the amount and timing of water injection, resulting in a high probability of partial core melting for unit 3.

  2. Progress on Radiochemical Analysis for Nuclear Waste Management in Decommissioning

    DEFF Research Database (Denmark)

    Hou, Xiaolin; Qiao, Jixin; Shi, Keliang

    With the increaed numbers of nuclear facilities have been closed and are being or are going to be decommissioned, it is required to characterise the produced nuclear waste for its treatment by identification of the radionuclides and qualitatively determine them. Of the radionuclides related...... separation of radionuclides. In order to improve and maintain the Nodic competence in analysis of radionculides in waste samples, a NKS B project on this topic was launched in 2009. During the first phase of the NKS-B RadWaste project (2009-2010), a good achivement has been reached on establishment...... of collaboration, identifing the requirements from the Nordic nuclear industries and optimizing and development of some analytical methods (Hou et al. NKS-222, 2010). In the year 2011, this project (NKS-B RadWaste2011) continued. The major achievements of this project in 2011 include: (1) development of a method...

  3. In-field analysis and assessment of nuclear material

    Energy Technology Data Exchange (ETDEWEB)

    Morgado, R.E.; Myers, W.S.; Olivares, J.A.; Phillips, J.R.; York, R.L.

    1996-05-01

    Los Alamos National Laboratory has actively developed and implemented a number of instruments to monitor, detect, and analyze nuclear materials in the field. Many of these technologies, developed under existing US Department of Energy programs, can also be used to effectively interdict nuclear materials smuggled across or within national borders. In particular, two instruments are suitable for immediate implementation: the NAVI-2, a hand-held gamma-ray and neutron system for the detection and rapid identification of radioactive materials, and the portable mass spectrometer for the rapid analysis of minute quantities of radioactive materials. Both instruments provide not only critical information about the characteristics of the nuclear material for law-enforcement agencies and national authorities but also supply health and safety information for personnel handling the suspect materials.

  4. Analysis of nuclear activity of ten polar ring galaxies

    CERN Document Server

    Freitas-Lemes, P; Dors, O L; Faúndez-Abans, M

    2013-01-01

    The accumulation of mass from the interaction process that forms the polar ring galaxies is a factor that favors the conditions necessary to trigger nonthermal nuclear activities.. This fact encouraged the chemical analysis of ten polar ring galaxies. In order to verify the presence of an active nucleus in these galaxias, we built diagnostic diagrams using lines H{\\beta}, [OIII], [HI], H{\\alpha}, [NII], and [SII] and classified the type of nuclear activity. For galaxies that do not show shock, the parameters N2 and O3N2 were also determined. From this sample, we identified seven galaxies with an active nucleus and three that behave as HII regions. One galaxy with an active nucleus was classified as Seyfert. Although our data do not provide a statistically significant sample, we can speculate that polar ring galaxies are a setting conducive to trigger non-thermal nuclear activities.

  5. Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed.

  6. Homogeneous protein analysis by magnetic core-shell nanorod probes

    KAUST Repository

    Schrittwieser, Stefan

    2016-03-29

    Studying protein interactions is of vital importance both to fundamental biology research and to medical applications. Here, we report on the experimental proof of a universally applicable label-free homogeneous platform for rapid protein analysis. It is based on optically detecting changes in the rotational dynamics of magnetically agitated core-shell nanorods upon their specific interaction with proteins. By adjusting the excitation frequency, we are able to optimize the measurement signal for each analyte protein size. In addition, due to the locking of the optical signal to the magnetic excitation frequency, background signals are suppressed, thus allowing exclusive studies of processes at the nanoprobe surface only. We study target proteins (soluble domain of the human epidermal growth factor receptor 2 - sHER2) specifically binding to antibodies (trastuzumab) immobilized on the surface of our nanoprobes and demonstrate direct deduction of their respective sizes. Additionally, we examine the dependence of our measurement signal on the concentration of the analyte protein, and deduce a minimally detectable sHER2 concentration of 440 pM. For our homogeneous measurement platform, good dispersion stability of the applied nanoprobes under physiological conditions is of vital importance. To that end, we support our measurement data by theoretical modeling of the total particle-particle interaction energies. The successful implementation of our platform offers scope for applications in biomarker-based diagnostics as well as for answering basic biology questions.

  7. Compact Short-Pulsed Electron Linac Based Neutron Sources for Precise Nuclear Material Analysis

    Science.gov (United States)

    Uesaka, M.; Tagi, K.; Matsuyama, D.; Fujiwara, T.; Dobashi, K.; Yamamoto, M.; Harada, H.

    2015-10-01

    An X-band (11.424GHz) electron linac as a neutron source for nuclear data study for the melted fuel debris analysis and nuclear security in Fukushima is under development. Originally we developed the linac for Compton scattering X-ray source. Quantitative material analysis and forensics for nuclear security will start several years later after the safe settlement of the accident is established. For the purpose, we should now accumulate more precise nuclear data of U, Pu, etc., especially in epithermal (0.1-10 eV) neutrons. Therefore, we have decided to modify and install the linac in the core space of the experimental nuclear reactor "Yayoi" which is now under the decommission procedure. Due to the compactness of the X-band linac, an electron gun, accelerating tube and other components can be installed in a small space in the core. First we plan to perform the time-of-flight (TOF) transmission measurement for study of total cross sections of the nuclei for 0.1-10 eV energy neutrons. Therefore, if we adopt a TOF line of less than 10m, the o-pulse length of generated neutrons should be shorter than 100 ns. Electronenergy, o-pulse length, power, and neutron yield are ~30 MeV, 100 ns - 1 micros, ~0.4 kW, and ~1011 n/s (~103 n/cm2/s at samples), respectively. Optimization of the design of a neutron target (Ta, W, 238U), TOF line and neutron detector (Ce:LiCAF) of high sensitivity and fast response is underway. We are upgrading the electron gun and a buncher to realize higher current and beam power with a reasonable beam size in order to avoid damage of the neutron target. Although the neutron flux is limited in case of the X-band electron linac based source, we take advantage of its short pulse aspect and availability for nuclear data measurement with a short TOF system. First, we form a tentative configuration in the current experimental room for Compton scattering in 2014. Then, after the decommissioning has been finished, we move it to the "Yayoi" room and perform

  8. Definition of a core module for the nuclear retrograde response to altered organellar gene expression identifies GLK overexpressors as gun mutants.

    Science.gov (United States)

    Leister, Dario; Kleine, Tatjana

    2016-07-01

    Retrograde signaling can be triggered by changes in organellar gene expression (OGE) induced by inhibitors such as lincomycin (LIN) or mutations that perturb OGE. Thus, an insufficiency of the organelle-targeted prolyl-tRNA synthetase PRORS1 in Arabidopsis thaliana activates retrograde signaling and reduces the expression of nuclear genes for photosynthetic proteins. Recently, we showed that mTERF6, a member of the so-called mitochondrial transcription termination factor (mTERF) family, is involved in the formation of chloroplast (cp) isoleucine-tRNA. To obtain further insights into its functions, co-expression analysis of MTERF6, PRORS1 and two other genes for organellar aminoacyl-tRNA synthetases was conducted. The results suggest a prominent role of mTERF6 in aminoacylation activity, light signaling and seed storage. Analysis of changes in whole-genome transcriptomes in the mterf6-1 mutant showed that levels of nuclear transcripts for cp OGE proteins were particularly affected. Comparison of the mterf6-1 transcriptome with that of prors1-2 showed that reduced aminoacylation of proline (prors1-2) and isoleucine (mterf6-1) tRNAs alters retrograde signaling in similar ways. Database analyses indicate that comparable gene expression changes are provoked by treatment with LIN, norflurazon or high light. A core OGE response module was defined by identifying genes that were differentially expressed under at least four of six conditions relevant to OGE signaling. Based on this module, overexpressors of the Golden2-like transcription factors GLK1 and GLK2 were identified as genomes uncoupled mutants.

  9. An analysis of international nuclear fuel supply options

    Science.gov (United States)

    Taylor, J'tia Patrice

    As the global demand for energy grows, many nations are considering developing or increasing nuclear capacity as a viable, long-term power source. To assess the possible expansion of nuclear power and the intricate relationships---which cover the range of economics, security, and material supply and demand---between established and aspirant nuclear generating entities requires models and system analysis tools that integrate all aspects of the nuclear enterprise. Computational tools and methods now exist across diverse research areas, such as operations research and nuclear engineering, to develop such a tool. This dissertation aims to develop methodologies and employ and expand on existing sources to develop a multipurpose tool to analyze international nuclear fuel supply options. The dissertation is comprised of two distinct components: the development of the Material, Economics, and Proliferation Assessment Tool (MEPAT), and analysis of fuel cycle scenarios using the tool. Development of MEPAT is aimed for unrestricted distribution and therefore uses publicly available and open-source codes in its development when possible. MEPAT is built using the Powersim Studio platform that is widely used in systems analysis. MEPAT development is divided into three modules focusing on: material movement; nonproliferation; and economics. The material movement module tracks material quantity in each process of the fuel cycle and in each nuclear program with respect to ownership, location and composition. The material movement module builds on techniques employed by fuel cycle models such as the Verifiable Fuel Cycle Simulation (VISION) code developed at the Idaho National Laboratory under the Advanced Fuel Cycle Initiative (AFCI) for the analysis of domestic fuel cycle. Material movement parameters such as lending and reactor preference, as well as fuel cycle parameters such as process times and material factors are user-specified through a Microsoft Excel(c) data spreadsheet

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-08-01

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

  11. Knowledge Economy Core Journals: Identification through LISTA Database Analysis.

    Science.gov (United States)

    Nouri, Rasool; Karimi, Saeed; Ashrafi-rizi, Hassan; Nouri, Azadeh

    2013-03-01

    Knowledge economy has become increasingly broad over the years and identification of core journals in this field can be useful for librarians in journal selection process and also for researchers to select their studies and finding Appropriate Journal for publishing their articles. Present research attempts to determine core journals of Knowledge Economy indexed in LISTA (Library and Information Science and Technology). The research method was bibliometric and research population include the journals indexed in LISTA (From the start until the beginning of 2011) with at least one article a bout "knowledge economy". For data collection, keywords about "knowledge economy"-were extracted from the literature in this area-have searched in LISTA by using title, keyword and abstract fields and also taking advantage of LISTA thesaurus. By using this search strategy, 1608 articles from 390 journals were retrieved. The retrieved records import in to the excel sheet and after that the journals were grouped and the Bradford's coefficient was measured for each group. Finally the average of the Bradford's coefficients were calculated and core journals with subject area of "Knowledge economy" were determined by using Bradford's formula. By using Bradford's scattering law, 15 journals with the highest publication rates were identified as "Knowledge economy" core journals indexed in LISTA. In this list "Library and Information update" with 64 articles was at the top. "ASLIB Proceedings" and "Serials" with 51 and 40 articles are next in rank. Also 41 journals were identified as beyond core that "Library Hi Tech" with 20 articles was at the top. Increased importance of knowledge economy has led to growth of production of articles in this subject area. So the evaluation of journals for ranking these journals becomes a very challenging task for librarians and generating core journal list can provide a useful tool for journal selection and also quick and easy access to information. Core

  12. Knowledge Economy Core Journals: Identification through LISTA Database Analysis

    OpenAIRE

    Nouri, Rasool; Karimi, Saeed; Ashrafi-rizi, Hassan; Nouri, Azadeh

    2013-01-01

    Background Knowledge economy has become increasingly broad over the years and identification of core journals in this field can be useful for librarians in journal selection process and also for researchers to select their studies and finding Appropriate Journal for publishing their articles. Present research attempts to determine core journals of Knowledge Economy indexed in LISTA (Library and Information Science and Technology). Methods The research method was bibliometric and research popu...

  13. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  14. Operational modal analysis of flow-induced vibration of nuclear fuel rods in a turbulent axial flow

    Energy Technology Data Exchange (ETDEWEB)

    De Pauw, B., E-mail: bdepauw@vub.ac.be [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium); Vrije Universiteit Brussel (VUB), Department of Mechanical Engineering (AVRG), Brussels (Belgium); Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, Mol (Belgium); Weijtjens, W.; Vanlanduit, S. [Vrije Universiteit Brussel (VUB), Department of Mechanical Engineering (AVRG), Brussels (Belgium); Van Tichelen, K. [Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, Mol (Belgium); Berghmans, F. [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium)

    2015-04-01

    Highlights: • We describe an analysis technique to evaluate nuclear fuel pins. • We test a single fuel pin mockup subjected to turbulent axial flow. • Our analysis is based on operational modal analysis (OMA). • The accuracy and precision of our method is higher compared to traditional methods. • We demonstrate the possible onset of a fluid-elastic instability. - Abstract: Flow-induced vibration of nuclear reactor fuel pins can result in mechanical noise and lead to failure of the reactor's fuel assembly. This problem can be exacerbated in the new generation of liquid heavy metal fast reactors that use a much denser and more viscous coolant in the reactor core. An investigation of the flow-induced vibration in these particular conditions is therefore essential. In this paper, we describe an analysis technique to evaluate flow-induced vibration of nuclear reactor fuel pins subjected to a turbulent axial flow of heavy metal. We deal with a single fuel pin mockup designed for the lead–bismuth eutectic (LBE) cooled MYRRHA reactor which is subjected to similar flow conditions as in the reactor core. Our analysis is based on operational modal analysis (OMA) techniques. We show that the accuracy and precision of our OMA technique is higher compared to traditional methods and that it allows evaluating the evolution of modal parameters in operational conditions. We also demonstrate the possible onset of a fluid-elastic instability by tracking the modal parameters with increasing flow velocity.

  15. Ultra-High-Density Molecular Core and Warped Nuclear Disk in the Deep Potential of Radio Lobe Galaxy NGC 3079

    Science.gov (United States)

    Sofue, Y.; Koda, J.; Kohno, K.; Okumura, S. K.; Honma, M.; Kawamura, A.; Irwin, Judith A.

    2001-02-01

    We have performed high-resolution synthesis observations of the 12CO (J=1-0) line emission from the radio lobe edge-on spiral galaxy NGC 3079 using a seven-element millimeter-wave interferometer at the Nobeyama Radio Observatory, which consisted of the 45 m telescope and six-element array. The nuclear molecular disk (NMD) of 750 pc radius is found to be inclined by 20° from the optical disk, and the NMD has spiral arms. An ultra-high-density core (UHC) of molecular gas was found at the nucleus. The gaseous mass of the UHC within 125 pc radius is as large as ~3×108 Msolar, an order of magnitude more massive than that in the same area of the Galactic center, and the mean density is as high as ~3×103H2 cm-3. A position-velocity diagram along the major axis indicates that the rotation curve already starts at a finite velocity exceeding 300 km s-1 from the nucleus. The surface mass density in the central region is estimated to be as high as ~105 Msolar pc-2, producing a very deep gravitational potential. We argue that the very large differential rotation in such a deep potential will keep the UHC gravitationally stable during the current star formation.

  16. The threat of nuclear terrorism: from analysis to precautionary measures

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, M

    2003-07-01

    Facing the nuclear terrorism risk, this document analyzes the nature of the threat of nuclear terrorism, the risk of attack on nuclear installations, the limited protection of nuclear installations against aircraft crashes, the case of nuclear reprocessing plants, the case of nuclear transport and proposes measures which should be taken without endangering the foundations of democracy. (A.L.B.)

  17. Nuclear spins, magnetic moments and quadrupole moments of Cu isotopes from N = 28 to N = 46: probes for core polarization effects

    CERN Document Server

    Vingerhoets, P; Avgoulea, M; Billowes, J; Bissell, M L; Blaum, K; Brown, B A; Cheal, B; De Rydt, M; Forest, D H; Geppert, Ch; Honma, M; Kowalska, M; Kramer, J; Krieger, A; Mane, E; Neugart, R; Neyens, G; Nortershauser, W; Otsuka, T; Schug, M; Stroke, H H; Tungate, G; Yordanov, D T

    2010-01-01

    Measurements of the ground-state nuclear spins, magnetic and quadrupole moments of the copper isotopes from 61Cu up to 75Cu are reported. The experiments were performed at the ISOLDE facility, using the technique of collinear laser spectroscopy. The trend in the magnetic moments between the N=28 and N=50 shell closures is reasonably reproduced by large-scale shell-model calculations starting from a 56Ni core. The quadrupole moments reveal a strong polarization of the underlying Ni core when the neutron shell is opened, which is however strongly reduced at N=40 due to the parity change between the $pf$ and $g$ orbits. No enhanced core polarization is seen beyond N=40. Deviations between measured and calculated moments are attributed to the softness of the 56Ni core and weakening of the Z=28 and N=28 shell gaps.

  18. TREAT Transient Analysis Benchmarking for the HEU Core

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used to determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.

  19. Complex signal amplitude analysis for complete fusion nuclear reaction products

    CERN Document Server

    Tsyganov, Yu S

    2015-01-01

    A complex analysis has been performed on the energy amplitude signals corresponding to events of Z=117 element measured in the 249Bk+48Ca complete fusion nuclear reaction. These signals were detected with PIPS position sensitive detector. The significant values of pulse height defect both for recoils (ER) and fission fragments (FF) were measured. Comparison with the computer simulations and empirical formulae has been performed both for ER and FF signals.

  20. Analysis and Design of ITER 1 MV Core Snubber

    Institute of Scientific and Technical Information of China (English)

    王海田; 李格

    2012-01-01

    The core snubber, as a passive protection device, can suppress arc current and absorb stored energy in stray capacitance during the electrical breakdown in accelerating electrodes of ITER NBI. In order to design the core snubber of ITER, the control parameters of the arc peak current have been firstly analyzed by the Fink-Baker-Owren (FBO) method, which are used for designing the DIIID 100 kV snubber. The B-H curve can be derived from the measured voltage and current waveforms, and the hysteresis loss of the core snubber can be derived using the revised parallelogram method. The core snubber can be a simplified representation as an equivalent parallel resistance and inductance, which has been neglected by the FBO method. A simulation code including the parallel equivalent resistance and inductance has been set up. The simulation and experiments result in dramatically large arc shorting currents due to the parallel inductance effect. The case shows that the core snubber utilizing the FBO method gives more compact design.

  1. Nuclear analysis of ITER Test Blanket Module Port Plug

    Energy Technology Data Exchange (ETDEWEB)

    Villari, Rosaria, E-mail: rosaria.villari@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Kim, Byoung Yoon; Barabash, Vladimir; Giancarli, Luciano; Levesy, Bruno; Loughlin, Michael; Merola, Mario [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Moro, Fabio [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Pascal, Romain [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Petrizzi, Luigino [European Commission, DG Research & Innovation G5, CDMA 00/030, B-1049 Brussels (Belgium); Polunovsky, Eduard; Van Der Laan, Jaap G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France)

    2015-10-15

    Highlights: • 3D nuclear analysis of the ITER TBM Port Plug (PP). • Calculations of neutron fluxes, nuclear heating, damage and He-production in TBM PP components. • Shutdown dose rate assessment with Advanced D1S method considering different configurations. • Potential design improvements to reduce the shutdown dose rate in the port interspace. - Abstract: Nuclear analyses have been performed for the ITER Test Blanket Module Port Plug (TBM PP) using the MCNP-5 Monte Carlo Code. A detailed 3D model of the TBM Port Plug with dummy TBM has been integrated into the ITER MCNP model (B-lite v.3). Neutron fluxes, nuclear heating, helium production and neutron damage have been calculated in all the TBM PP components. Global shutdown dose rate calculations have also been performed with Advanced D1S method for different configurations of the TBM PP system. This paper presents the results of these analyses and discusses potential design improvements aiming to further reduce the shutdown dose rate in the port interspace.

  2. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

    2014-10-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  3. Comparative analysis of radiation characteristics from various types of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kryuchkov, E.F.; Opalovsky, V.A.; Tikhomirov, G.V. [Moscow Engineering Physics Institute (State University) (Russian Federation)

    2003-07-01

    At the present time, in purposes of the most effective utilization of nuclear materials, new advanced fuel cycles are under development. These cycles imply application of uranium-plutonium, uranium-thorium and some other types of nuclear fuel. However, it is obvious that the parameters of new nuclear fuel (NF) types will be quite different from those for traditional NF types. These differences can affect significantly the conditions for storage, transportation and reprocessing of spent nuclear fuel (SNF). So, it is necessary to carry out a comparative analysis of radiation characteristics for various NF types at different stages of nuclear fuel cycle (NFC). The present paper addresses radiation properties of the following NF types: UO{sub 2}, UO{sub 2}-PuO{sub 2}, ThO{sub 2}-PaO{sub 2}-UO{sub 2}. Numerical studies have been carried out to determine radiation properties of these NF types at the following NFC stages: radiation properties of NF directly before and after irradiation in the reactor core, after different cooling time, radiation properties of uranium and plutonium fractions after chemical separation, radiation properties of NF re-fabricated for recycle, radiation properties of NF after the second and third recycles. The computer code package SCALE is used for evaluating the radiation properties of different SNF types. Finally, the following major conclusions can be made: 1) Correct description of SNF radiation and dosimetric properties requires available benchmark data on contents of heavy nuclides in SNF; 2) ThO{sub 2}-PaO{sub 2}-UO{sub 2} fuel demonstrates an important feature: internal transmutation of minor actinides provided the ultra-high fuel burn-up is achieved.

  4. Elucidating the domain architecture and functions of non-core RAG1: The capacity of a non-core zinc-binding domain to function in nuclear import and nucleic acid binding

    Directory of Open Access Journals (Sweden)

    Zhao Shuying

    2011-05-01

    Full Text Available Abstract Background The repertoire of the antigen-binding receptors originates from the rearrangement of immunoglobulin and T-cell receptor genetic loci in a process known as V(DJ recombination. The initial site-specific DNA cleavage steps of this process are catalyzed by the lymphoid specific proteins RAG1 and RAG2. The majority of studies on RAG1 and RAG2 have focused on the minimal, core regions required for catalytic activity. Though not absolutely required, non-core regions of RAG1 and RAG2 have been shown to influence the efficiency and fidelity of the recombination reaction. Results Using a partial proteolysis approach in combination with bioinformatics analyses, we identified the domain boundaries of a structural domain that is present in the 380-residue N-terminal non-core region of RAG1. We term this domain the Central Non-core Domain (CND; residues 87-217. Conclusions We show how the CND alone, and in combination with other regions of non-core RAG1, functions in nuclear localization, zinc coordination, and interactions with nucleic acid. Together, these results demonstrate the multiple roles that the non-core region can play in the function of the full length protein.

  5. Analysis of the In-core Quadrant Power Tilt affected by Burned Fuel Shuffles of WEC Type NPPs in Republic of Korea

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seung-beom; Jeon, Jeong-pyo; Song, Han-seung; Seong, Ki-bong; Lim, Chae-joon [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2016-10-15

    This paper presents the designed Burned Fuel Shuffles (BFS) and the related results of measured In-core Quadrant Power Tilt (IQPT) in recent cycles of WEC (Westinghouse Electric Company) type NPPs (Nuclear Power Plants) in Republic of Korea. And the IQPT sensitivity results affected by BFS are also analyzed. Excessive core Quadrant Power Tilt (QPT) causes unreliability about designed power distribution and increases peaking factors in the affected core quadrants. The peaking factors are under surveillance during the cycle for the safe operation and the Quadrant Power Tilt Ratio (QPTR) is covered by the Technical Specifications. Possible causes for QPT include manufacturing tolerance, asymmetric core configurations, operating conditions, and so forth, but the actual cause of specific core tilts frequently cannot be definitively identified. But nuclear designer continuously try to minimize the QPT by the general control of burned fuel distribution in a reload core. In this study, the general guidelines of BFSP for effective mitigation of IQPT were introduced by references and the actual states of designed BFSP were analyzed for WEC type plant operating in the Republic of Korea. Results revealed that the BFSP was applied within appropriate level, which keeps IQPT below the level of guideline during the operations. Also, the correlation between BFSP of category 1/3 and IQPT were quantitatively confirmed by the sensitivity analysis concerned with the change of BFSP.

  6. Preliminary Uncertainty Analysis for SMART Digital Core Protection and Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Bon Seung; In, Wang Kee; Hwang, Dae Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The Korea Atomic Energy Research Institute (KAERI) developed on-line digital core protection and monitoring systems, called SCOPS and SCOMS as a part of SMART plant protection and monitoring system. SCOPS simplified the protection system by directly connecting the four RSPT signals to each core protection channel and eliminated the control element assembly calculator (CEAC) hardware. SCOMS adopted DPCM3D method in synthesizing core power distribution instead of Fourier expansion method being used in conventional PWRs. The DPCM3D method produces a synthetic 3-D power distribution by coupling a neutronics code and measured in-core detector signals. The overall uncertainty analysis methodology which is used statistically combining uncertainty components of SMART core protection and monitoring system was developed. In this paper, preliminary overall uncertainty factors for SCOPS/SCOMS of SMART initial core were evaluated by applying newly developed uncertainty analysis method

  7. Fuel cycle analysis of once-through nuclear systems.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  8. Fuel cycle analysis of once-through nuclear systems.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  9. Magnetic, Structural, and Particle Size Analysis of Single- and Multi-Core Magnetic Nanoparticles

    DEFF Research Database (Denmark)

    Ludwig, Frank; Kazakova, Olga; Barquin, Luis Fernandez

    2014-01-01

    We have measured and analyzed three different commercial magnetic nanoparticle systems, both multi-core and single-core in nature, with the particle (core) size ranging from 20 to 100 nm. Complementary analysis methods and same characterization techniques were carried out in different labs...... and the results are compared with each other. The presented results primarily focus on determining the particle size—both the hydrodynamic size and the individual magnetic core size—as well as magnetic and structural properties. The used analysis methods include transmission electron microscopy, static...... and dynamic magnetization measurements, and Mössbauer spectroscopy. We show that particle (hydrodynamic and core) size parameters can be determined from different analysis techniques and the individual analysis results agree reasonably well. However, in order to compare size parameters precisely determined...

  10. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  11. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, Tim, E-mail: tdbohm@wisc.edu [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, Michael; Bullock, James [Formerly, Fusion Technology, Sandia National Laboratories, Albuquerque, NM (United States)

    2015-10-15

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  12. Vulnerability Analysis Considerations for the Transportation of Special Nuclear Material

    Energy Technology Data Exchange (ETDEWEB)

    Nicholson, Lary G.; Purvis, James W.

    1999-07-21

    The vulnerability analysis methodology developed for fixed nuclear material sites has proven to be extremely effective in assessing associated transportation issues. The basic methods and techniques used are directly applicable to conducting a transportation vulnerability analysis. The purpose of this paper is to illustrate that the same physical protection elements (detection, delay, and response) are present, although the response force plays a dominant role in preventing the theft or sabotage of material. Transportation systems are continuously exposed to the general public whereas the fixed site location by its very nature restricts general public access.

  13. Radiochemical analysis for nuclear waste management in decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Hou, X. (Technical Univ. of Denmark, Risoe National Lab. for Sustainable Energy. Radiation Research Div., Roskilde (Denmark))

    2010-07-15

    The NKS-B RadWaste project was launched from June 2009. The on-going decommissioning activities in Nordic countries and current requirements and problems on the radiochemical analysis of decommissioning waste were discussed and overviewed. The radiochemical analytical methods used for determination of various radionuclides in nuclear waste are reviewed, a book was written by the project partners Jukka Lehto and Xiaolin Hou on the chemistry and analysis of radionuclide to be published in 2010. A summary of the methods developed in Nordic laboratories is described in this report. The progresses on the development and optimization of analytical method in the Nordic labs under this project are presented. (author)

  14. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor, Rev. 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Hoon; Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun

    2005-12-15

    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  15. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun

    2005-07-15

    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  16. Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

    OpenAIRE

    Shen, Xiuzhong; Nakajima, Ken; Unesaki, Hironobu; Mishima, Kaichiro

    2013-01-01

    The purpose of this study is to realize the full core conversion from the use of High Enriched Uranium (HEU) fuels to the use of Low Enriched Uranium (LEU) fuels in Kyoto University Research Reactor (KUR). Although the conversion of nuclear energy sources is required to keep the safety margins and reactor reliability based on KUR HEU core, the uranium density (3.2 gU/cm3) and enrichment (20%) of LEU fuel (U3Si2–AL) are quite different from the uranium density (0.58 gU/cm3) and enrichment (93%...

  17. INTEGRATION OF FACILITY MODELING CAPABILITIES FOR NUCLEAR NONPROLIFERATION ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.; Hamm, L.; Garcia, H.; Burr, T.; Coles, G.; Edmunds, T.; Garrett, A.; Krebs, J.; Kress, R.; Lamberti, V.; Schoenwald, D.; Tzanos, C.; Ward, R.

    2011-07-18

    Developing automated methods for data collection and analysis that can facilitate nuclear nonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facility modeling that can integrate and interpret observations collected from monitored facilities in order to ascertain their functional details will be a critical element of these methods. Although improvements are continually sought, existing facility modeling tools can characterize all aspects of reactor operations and the majority of nuclear fuel cycle processing steps, and include algorithms for data processing and interpretation. Assessing nonproliferation status is challenging because observations can come from many sources, including local and remote sensors that monitor facility operations, as well as open sources that provide specific business information about the monitored facilities, and can be of many different types. Although many current facility models are capable of analyzing large amounts of information, they have not been integrated in an analyst-friendly manner. This paper addresses some of these facility modeling capabilities and illustrates how they could be integrated and utilized for nonproliferation analysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facility modeling tools. After considering a representative sampling of key facility modeling capabilities, the proposed integration framework is illustrated with several examples.

  18. Seismological analysis of the fourth North Korean nuclear test

    Science.gov (United States)

    Hartmann, Gernot; Gestermann, Nicolai; Ceranna, Lars

    2016-04-01

    The Democratic People's Republic of Korea has conducted its fourth underground nuclear explosions on 06.01.2016 at 01:30 (UTC). The explosion was clearly detected and located by the seismic network of the International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). Additional seismic stations of international earthquake monitoring networks at regional distances, which are not part of the IMS, are used to precisely estimate the epicenter of the event in the North Hamgyong province (41.38°N / 129.05°E). It is located in the area of the North Korean Punggye-ri nuclear test site, where the verified nuclear tests from 2006, 2009, and 2013 were conducted as well. The analysis of the recorded seismic signals provides the evidence, that the event was originated by an explosive source. The amplitudes as well as the spectral characteristics of the signals were examined. Furthermore, the similarity of the signals with those from the three former nuclear tests suggests very similar source type. The seismograms at the 8,200 km distant IMS station GERES in Germany, for example, show the same P phase signal for all four explosions, differing in the amplitude only. The comparison of the measured amplitudes results in the increasing magnitude with the chronology of the explosions from 2006 (mb 4.2), 2009 (mb 4.8) until 2013 (mb 5.1), whereas the explosion in 2016 had approximately the same magnitude as that one three years before. Derived from the magnitude, a yield of 14 kt TNT equivalents was estimated for both explosions in 2013 and 2016; in 2006 and 2009 yields were 0.7 kt and 5.4 kt, respectively. However, a large inherent uncertainty for these values has to be taken into account. The estimation of the absolute yield of the explosions depends very much on the local geological situation and the degree of decoupling of the explosive from the surrounding rock. Due to the missing corresponding information, reliable magnitude-yield estimation for the

  19. Radionuclide analysis on bamboos following the Fukushima nuclear accident.

    Directory of Open Access Journals (Sweden)

    Takumi Higaki

    Full Text Available In response to contamination from the recent Fukushima nuclear accident, we conducted radionuclide analysis on bamboos sampled from six sites within a 25 to 980 km radius of the Fukushima Daiichi nuclear power plant. Maximum activity concentrations of radiocesium (134Cs and (137Cs in samples from Fukushima city, 65 km away from the Fukushima Daiichi plant, were in excess of 71 and 79 kBq/kg, dry weight (DW, respectively. In Kashiwa city, 195 km away from the Fukushima Daiichi, the sample concentrations were in excess of 3.4 and 4.3 kBq/kg DW, respectively. In Toyohashi city, 440 km away from the Fukushima Daiichi, the concentrations were below the measurable limits of up to 4.5 Bq/kg DW. In the radiocesium contaminated samples, the radiocesium activity was higher in mature and fallen leaves than in young leaves, branches and culms.

  20. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide...

  1. Radiation exposure analysis of female nuclear medicine radiation workers

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ju Young [Dept. of Biomedical Engineering Graduate School, Chungbuk National University, Cheongju (Korea, Republic of); Park, Hoon Hee [Dept. of Radiological Technologist, Shingu College, Sungnam (Korea, Republic of)

    2016-06-15

    In this study, radiation workers who work in nuclear medicine department were analyzed to find the cause of differences of radiation exposure from General Characteristic, Knowledge, Recognition and Conduct, especially females working on nuclear medicine radiation, in order to pave the way for positive defense against radiation exposure. The subjects were 106 radiation workers who were divided into two groups of sixty-four males and forty-two females answered questions about their General Characteristic, Knowledge, Recognition, Conduct, and radiation exposure dose which was measured by TLD (Thermo Luminescence Dosimeter). The results of the analysis revealed that as the higher score of knowledge and conduct was shown, the radiation exposure decreased in female groups, and as the higher score of conduct was shown, the radiation exposure decreased in male groups. In the correlation analysis of female groups, the non-experienced in pregnancy showed decreasing amount of radiation exposure as the score of knowledge and conduct was higher and the experienced in pregnancy showed decreasing amount of radiation exposure as the score of recognition and conduct was higher. In the regression analysis on related factors of radiation exposure dose of nuclear medicine radiation workers, the gender caused the meaningful result and the amount of radiation exposure of female groups compared to male groups. In the regression analysis on related factors of radiation exposure dose of female groups, the factor of conduct showed a meaningful result and the amount of radiation exposure of the experienced in pregnancy was lower compared to the non-experienced. The conclusion of this study revealed that radiation exposure of female groups was lower than that of male groups. Therefore, male groups need to more actively defend themselves against radiation exposure. Among the female groups, the experienced in pregnancy who have an active defense tendency showed a lower radiation exposure. Thus

  2. Performance modeling and analysis of parallel Gaussian elimination on multi-core computers

    Directory of Open Access Journals (Sweden)

    Fadi N. Sibai

    2014-01-01

    Full Text Available Gaussian elimination is used in many applications and in particular in the solution of systems of linear equations. This paper presents mathematical performance models and analysis of four parallel Gaussian Elimination methods (precisely the Original method and the new Meet in the Middle –MiM– algorithms and their variants with SIMD vectorization on multi-core systems. Analytical performance models of the four methods are formulated and presented followed by evaluations of these models with modern multi-core systems’ operation latencies. Our results reveal that the four methods generally exhibit good performance scaling with increasing matrix size and number of cores. SIMD vectorization only makes a large difference in performance for low number of cores. For a large matrix size (n ⩾ 16 K, the performance difference between the MiM and Original methods falls from 16× with four cores to 4× with 16 K cores. The efficiencies of all four methods are low with 1 K cores or more stressing a major problem of multi-core systems where the network-on-chip and memory latencies are too high in relation to basic arithmetic operations. Thus Gaussian Elimination can greatly benefit from the resources of multi-core systems, but higher performance gains can be achieved if multi-core systems can be designed with lower memory operation, synchronization, and interconnect communication latencies, requirements of utmost importance and challenge in the exascale computing age.

  3. Core influence on the frequency response analysis (FRA of power transformers through the finite element method

    Directory of Open Access Journals (Sweden)

    D. L. Alvarez

    2015-11-01

    Full Text Available In this paper the influence of core parameters in Frequency Response Analysis is analyzed through the equivalent circuit impedance matrix of the transformer winding; the parameters of the circuit have been computed using the Finite Element Method. In order to appreciate the behavior of the iron core in comparison to the air core, the frequency dependence of resonances is calculated to show how the air core only influences the results at low frequencies. The core is modeled using a complex permeability, and the parameters of conductivity and permeability are varied to show their influence in the resonances, which turned out to be negligible. In order to explain this behavior, the eigenvalues of the inverse impedance matrix are calculated showing that they are similar for different values of conductivity and permeability. Finally, the magnetic flux inside and outside the core and its influence in the frequency response is studied.

  4. Supermode analysis of the 18-core photonic crystal fiber laser

    Institute of Scientific and Technical Information of China (English)

    王远; 姚建铨; 郑一博; 温午麒; 陆颖; 王鹏

    2012-01-01

    The modal of 18-core photonic crystal fiber laser is discussed and calculated.And corresponding far-field distribution of the supermodes is given by Fresnel diffraction integral.For improving beam quality,the mode selection method based on the Talbot effect is introduced.The reflection coefficients are calculated,and the result shows that an in-phase supermode can be locked better at a large propagation distance.

  5. Analysis of White Dwarfs with Strange-Matter Cores

    CERN Document Server

    Mathews, G J; O'Gorman, B; Lan, N Q; Zech, W; Otsuki, K; Weber, F

    2006-01-01

    We summarize masses and radii for a number of white dwarfs as deduced from a combination of proper motion studies, Hipparcos parallax distances, effective temperatures, and binary or spectroscopic masses. A puzzling feature of these data is that some stars appear to have radii which are significantly smaller than that expected for a standard electron-degenerate white-dwarf equations of state. We construct a projection of white-dwarf radii for fixed effective mass and conclude that there is at least marginal evidence for bimodality in the radius distribution forwhite dwarfs. We argue that if such compact white dwarfs exist it is unlikely that they contain an iron core. We propose an alternative of strange-quark matter within the white-dwarf core. We also discuss the impact of the so-called color-flavor locked (CFL) state in strange-matter core associated with color superconductivity. We show that the data exhibit several features consistent with the expected mass-radius relation of strange dwarfs. We identify ...

  6. A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Painter, Chad L.; Geelhood, Ken J.; Wootan, David W.; Meriwether, George H.; Cuta, Judith M.; Adkins, Harold E.; Matson, Dean W.; Abrego, Celestino P.

    2007-12-01

    Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling, core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.

  7. Analysis algorithm for digital data used in nuclear spectroscopy

    CERN Document Server

    AUTHOR|(CDS)2085950; Sin, Mihaela

    Data obtained from digital acquisition systems used in nuclear spectroscopy experiments must be converted by a dedicated algorithm in or- der to extract the physical quantities of interest. I will report here the de- velopment of an algorithm capable to read digital data, discriminate between random and true signals and convert the results into a format readable by a special data analysis program package used to interpret nuclear spectra and to create coincident matrices. The algorithm can be used in any nuclear spectroscopy experimental setup provided that digital acquisition modules are involved. In particular it was used to treat data obtained from the IS441 experiment at ISOLDE where the beta decay of 80Zn was investigated as part of ultra-fast timing studies of neutron rich Zn nuclei. The results obtained for the half-lives of 80Zn and 80Ga were in very good agreement with previous measurements. This fact proved unquestionably that the conversion algorithm works. Another remarkable result was the improve...

  8. Analysis of Proliferation Resistance of Nuclear Fuel Cycle Systems

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Hong Lae; Ko, Won Il; Kim, Ho Dong

    2009-11-15

    Proliferation resistance (PR) has been evaluated for the five nuclear fuel cycle systems, potentially deployable in Korea in the future, using the fourteen proliferation resistance attributes suggested in the TOPS report. Unidimensional Utility Theory (UUT) was used in the calculation of utility value for each of the fourteen proliferation resistance attributes, and Multi-Attribute Utility Theory (MAUT), a decision tool with multiple objectives, was used in the evaluation of the proliferation resistance of each nuclear fuel cycle system. Analytic Hierarchy Process (AHP) and Expert Elicitation (EE) were utilized in the derivation of weighting factors for the fourteen proliferation resistance attributes. Among the five nuclear fuel cycle systems evaluated, the once-through fuel cycle system showed the highest level of proliferation resistance, and Pyroprocessing-SFR fuel cycle system showed the similar level of proliferation resistance with the DUPIC fuel cycle system, which has two time higher level of proliferation resistance compared to that of the thermal MOX fuel cycle system. Sensitivity analysis was also carried out to make up for the uncertainty associated with the derivation of weighting factors for the fourteen proliferation resistance attributes.

  9. In situ structural analysis of the human nuclear pore complex.

    Science.gov (United States)

    von Appen, Alexander; Kosinski, Jan; Sparks, Lenore; Ori, Alessandro; DiGuilio, Amanda L; Vollmer, Benjamin; Mackmull, Marie-Therese; Banterle, Niccolo; Parca, Luca; Kastritis, Panagiotis; Buczak, Katarzyna; Mosalaganti, Shyamal; Hagen, Wim; Andres-Pons, Amparo; Lemke, Edward A; Bork, Peer; Antonin, Wolfram; Glavy, Joseph S; Bui, Khanh Huy; Beck, Martin

    2015-10-01

    Nuclear pore complexes are fundamental components of all eukaryotic cells that mediate nucleocytoplasmic exchange. Determining their 110-megadalton structure imposes a formidable challenge and requires in situ structural biology approaches. Of approximately 30 nucleoporins (Nups), 15 are structured and form the Y and inner-ring complexes. These two major scaffolding modules assemble in multiple copies into an eight-fold rotationally symmetric structure that fuses the inner and outer nuclear membranes to form a central channel of ~60 nm in diameter. The scaffold is decorated with transport-channel Nups that often contain phenylalanine-repeat sequences and mediate the interaction with cargo complexes. Although the architectural arrangement of parts of the Y complex has been elucidated, it is unclear how exactly it oligomerizes in situ. Here we combine cryo-electron tomography with mass spectrometry, biochemical analysis, perturbation experiments and structural modelling to generate, to our knowledge, the most comprehensive architectural model of the human nuclear pore complex to date. Our data suggest previously unknown protein interfaces across Y complexes and to inner-ring complex members. We show that the transport-channel Nup358 (also known as Ranbp2) has a previously unanticipated role in Y-complex oligomerization. Our findings blur the established boundaries between scaffold and transport-channel Nups. We conclude that, similar to coated vesicles, several copies of the same structural building block--although compositionally identical--engage in different local sets of interactions and conformations.

  10. Evaluation Indicators for Analysis of Nuclear Fuel Cycle Sustainability

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Ko, Won Il; Chang, Hong Lae

    2008-01-15

    In this report, an attempt was made to derive indicators for the evaluation of the sustainability of the nuclear fuel cycle, using the methodologies developed by the INPRO, OECD/NEA and Gen-IV. In deriving the indicators, the three main elements of the sustainability, i.e., economics, environmental impact, and social aspect, as well as the technological aspect of the nuclear fuel cycle, considering the importance of the safety, were selected as the main criteria. An evaluation indicator for each criterion was determined, and the contents and evaluation method of each indicator were proposed. In addition, a questionnaire survey was carried out for the objectivity of the selection of the indicators in which participated some experts of the Korea Energy Technology and Emergency Management Institute (KETEMI) . Although the proposed indicators do not satisfy the characteristics and requirements of general indicators, it is presumed that they can be used in the analysis of the sustainability of the nuclear fuel cycle because those indicators incorporate various expert judgment and public opinions. On the other hand, the weighting factor of each indicator should be complemented in the future, using the AHP method and expert advice/consultations.

  11. Preliminary radiation criteria and nuclear analysis for ETF

    Energy Technology Data Exchange (ETDEWEB)

    Engholm, B.A.

    1980-09-01

    Preliminary biological and materials radiation dose criteria for the Engineering Test Facility are described and tabulated. In keeping with the ETF Mission Statement, a key biological dose criterion is a 24-hour shutdown dose rate of 2 mrem/hr on the surface of the outboard bulk shield. Materials dose criteria, which primarily govern the inboard shield design, include 10/sup 9/ rads exposure limit to epoxy insulation, 3 x 10/sup -4/ dpa damage to the TF coil copper stabilizer, and a total nuclear heating rate of 5 kW in the inboard TF coils. Nuclear analysis performed during FY 80 was directed primarily at the inboard and outboard bulk shielding, and at radiation streaming in the neutral beam drift ducts. Inboard and outboard shield thicknesses to achieve the biological and materials radiation criteria are 75 cm inboard and 125 cm outboard, the configuration consisting of alternating layers of stainless steel and borated water. The outboard shield also includes a 5 cm layer of lead. NBI duct streaming analyses performed by ORNL and LASL will play a key role in the design of the duct and NBI shielding in FY 81. The NBI aluminum cryopanel nuclear heating rate during the heating cycle is about 1 milliwatt/cm/sup 3/, which is far less than the permissible limit.

  12. Description and Analysis of Core Samples: The Lunar Experience

    Science.gov (United States)

    McKay, David S.; Allton, Judith H.

    1997-01-01

    Although no samples yet have been returned from a comet, extensive experience from sampling another solar system body, the Moon, does exist. While, in overall structure, composition, and physical properties the Moon bears little resemblance to what is expected for a comet, sampling the Moon has provided some basic lessons in how to do things which may be equally applicable to cometary samples. In particular, an extensive series of core samples has been taken on the Moon, and coring is the best way to sample a comet in three dimensions. Data from cores taken at 24 Apollo collection stations and 3 Luna sites have been used to provide insight into the evolution of the lunar regolith. It is now well understood that this regolith is very complex and reflects gardening (stirring of grains by micrometeorites), erosion (from impacts and solar wind sputtering), maturation (exposure on the bare lunar surface to solar winds ions and micrometeorite impacts) and comminution of coarse grains into finer grains, blanket deposition of coarse-grained layers, and other processes. All of these processes have been documented in cores. While a cometary regolith should not be expected to parallel in detail the lunar regolith, it is possible that the upper part of a cometary regolith may include textural, mineralogical, and chemical features which reflect the original accretion of the comet, including a form of gardening. Differences in relative velocities and gravitational attraction no doubt made this accretionary gardening qualitatively much different than the lunar version. Furthermore, at least some comets, depending on their orbits, have been subjected to impacts of the uppermost surface by small projectiles at some time in their history. Consequently, a more recent post-accretional gardening may have occurred. Finally, for comets which approach the sun, large scale erosion may have occurred driven by gas loss. The uppermost material of these comets may reflect some of the features

  13. Thermohydraulic Design Analysis Modeling for Korea Advanced NUclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    NTR engines have continued as a main stream based on the mature technology. The typical core design of the NERVA derived engines uses hexagonal shaped fuel elements with circular cooling channels and structural tie-tube elements for supporting the fuel elements, housing moderator and regeneratively cooling the moderator. The state-of-the-art NTR designs mostly use a fast or epithermal neutron spectrum core utilizing a HEU fuel to make a high power reactor with small and simple core geometry. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The KANUTER-HEU and -LEU are the innovative and futuristic NTR engines to reduce the reactor size and to implement a LEU fuel in the reactor by using thermal neutron spectrum. The KANUTERs have some features in the reactor design such as the integrated fuel element and the regeneratively cooling channels to increase room for moderator and heat transfer in the core, and ensuing rocket performance. To study feasible design points in terms of thermo-hydraulics and to estimate rocket performance of the KANUTERs, the NSES is under development. The model of the NSES currently focuses on thermo-hydraulic analysis of the peculiar and complex EHTGR design during the propulsion mode in steady-state. The results indicate comparable performance for future applications, even though it uses the heavier LEU fuel. In future, the NSES will be modified to obtain temperature distribution of the entire reactor components and then more extensive design analysis of neutronics, thermohydraulics and their coupling will be conducted to validate design feasibility and to optimize the reactor design enhancing the rocket performance.

  14. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  15. Developments and needs in nuclear analysis of fusion technology

    Energy Technology Data Exchange (ETDEWEB)

    Pampin, R., E-mail: raul.pampin@f4e.europa.eu [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Davis, A. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Izquierdo, J. [F4E Fusion For Energy, Josep Pla 2, Torres Diagonal Litoral B3, Barcelona 08019 (Spain); Leichtle, D. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz Platz 1, D-76344 Karlsruhe (Germany); Loughlin, M.J. [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul lez Durance (France); Sanz, J. [UNED, Departamento de Ingenieria Energetica, Juan del Rosal 12, 28040 Madrid (Spain); Turner, A. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Villari, R. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Wilson, P.P.H. [University of Wisconsin, Nuclear Engineering Department, Madison, WI (United States)

    2013-10-15

    Highlights: • Complex fusion nuclear analyses require detailed models, sophisticated acceleration and coupling of cumbersome tools. • Progress on development of tools and methods to meet specific needs of fusion nuclear analysis reported. • Advances in production of reference models and in preparation and QA of acceleration and coupling algorithms shown. • Evaluation and adaptation studies of alternative transport codes presented. • Discussion made of the importance of efforts in these and other areas, considering some of the more pressing needs. -- Abstract: Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case of fusion technology in current experiments, ITER, next-step devices and power plant studies. Calculations are intricate and computer-intensive, typically requiring detailed geometry models, sophisticated acceleration algorithms, high-performance parallel computations, and coupling of large and complex transport and activation codes and databases. This paper reports progress on some key areas in the development of tools and methods to meet the specific needs of fusion nuclear analyses. In particular, advances in the production and modernisation of reference models, in the preparation and quality assurance of acceleration algorithms and coupling schemes, and in the evaluation and adaptation of alternative transport codes are presented. Emphasis is given to ITER-relevant activities, which are the main driver of advances in the field. Discussion is made of the importance of efforts in these and other areas, considering some of the more pressing needs and requirements. In some cases, they call for a more efficient and coordinated use of the scarce resources available.

  16. Human Factors Considerations in New Nuclear Power Plants: Detailed Analysis.

    Energy Technology Data Exchange (ETDEWEB)

    OHara,J.; Higgins, J.; Brown, W.; Fink, R.

    2008-02-14

    This Nuclear Regulatory Commission (NRC) sponsored study has identified human-performance issues in new and advanced nuclear power plants. To identify the issues, current industry developments and trends were evaluated in the areas of reactor technology, instrumentation and control technology, human-system integration technology, and human factors engineering (HFE) methods and tools. The issues were organized into seven high-level HFE topic areas: Role of Personnel and Automation, Staffing and Training, Normal Operations Management, Disturbance and Emergency Management, Maintenance and Change Management, Plant Design and Construction, and HFE Methods and Tools. The issues where then prioritized into four categories using a 'Phenomena Identification and Ranking Table' methodology based on evaluations provided by 14 independent subject matter experts. The subject matter experts were knowledgeable in a variety of disciplines. Vendors, utilities, research organizations and regulators all participated. Twenty issues were categorized into the top priority category. This Brookhaven National Laboratory (BNL) technical report provides the detailed methodology, issue analysis, and results. A summary of the results of this study can be found in NUREG/CR-6947. The research performed for this project has identified a large number of human-performance issues for new control stations and new nuclear power plant designs. The information gathered in this project can serve as input to the development of a long-term strategy and plan for addressing human performance in these areas through regulatory research. Addressing human-performance issues will provide the technical basis from which regulatory review guidance can be developed to meet these challenges. The availability of this review guidance will help set clear expectations for how the NRC staff will evaluate new designs, reduce regulatory uncertainty, and provide a well-defined path to new nuclear power plant

  17. Development of RCM analysis software for Korean nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ho; Choi, Kwang Hee; Jeong, Hyeong Jong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A software called KEPCO RCM workstation (KRCM) has been developed to optimize the maintenance strategies of Korean nuclear power plants. The program modules of the KRCM were designed in a manner that combines EPRI methodologies and KEPRI analysis technique. The KRCM is being applied to the three pilot system, chemical and volume control system, main steam system, and compressed air system of Yonggwang Units 1 and 2. In addition, the KRCM can be utilized as a tool to meet a part of the requirements of maintenance rule (MR) imposed by U.S. NRC. 3 refs., 4 figs. (Author)

  18. Nuclear analysis of the IFMIF European lithium target assembly system

    Energy Technology Data Exchange (ETDEWEB)

    Frisoni, M., E-mail: manuela.frisoni@enea.it [ENEA Bologna, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Bernardi, D.; Miccichè, G.; Serra, M. [ENEA CR Brasimone, Bacino del Brasimone 40032, Camugnano (Italy)

    2014-10-15

    Highlights: •Coupled n–γ transport calculations performed for the ENEA target assembly system. •The MCNP5 1.6 Monte Carlo code was used integrated with the McDeLicious-11 code. •A complete mapping of nuclear responses provided for the thermomechanical analysis. •Neutron activation calculations were performed for backplate, frame, nozzle and target chamber. •Results confirm that the design of the system needs remote handling tools. -- Abstract: In the framework of the Engineering Validation and Engineering Design Activities (EVEDA) phase of the International Fusion Materials Irradiation Facility (IFMIF) project, ENEA was in charge of the design of the European version of the target assembly (TA) system which employs a removable bayonet backplate (BP) concept. With the aim of assessing the nuclear behaviour of the system and supplying the necessary input data to the thermomechanical analysis, coupled neutron-gamma transport calculations have been carried out for the whole TA + BP system, using the MCNP5 1.6 Monte Carlo transport code integrated with the McDeLicious-11 neutron source code provided by KIT. Neutron activation calculations have been performed by means of the EASY-2010 activation system in order to provide radioactive inventories useful for thermomechanical analysis and safety purposes. This paper summarizes the results obtained by the neutronic and activation calculations for the most irradiated components of the TA, such as backplate, frame, nozzle and target chamber.

  19. Numerical analysis of a nuclear fuel element for nuclear thermal propulsion

    Science.gov (United States)

    Wang, Ten-See; Schutzenhofer, Luke

    1991-01-01

    A computational fluid dynamics model with porosity and permeability formulations in the transport equations has been developed to study the concept of nuclear thermal propulsion through the analysis of a pulsed irradiation of a particle bed element (PIPE). The numerical model is a time-accurate pressure-based formulation. An adaptive upwind scheme is employed for spatial discretization. The upwind scheme is based on second- and fourth-order central differencing with adaptive artificial dissipation. Multiblocked porosity regions have been formulated to model the cold frit, particle bed, and hot frit. Multiblocked permeability regions have been formulated to describe the flow shaping effect from the thickness-varying cold frit. Computational results for several zero-power density PIPEs and an elevated-particle-temperature PIPE are presented. The implications of the computational results are discussed.

  20. Analysis in environmental radioactivity around Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yong Woo; Han, Man Jung; Cho, Seong Won; Cho, Hong Jun; Oh, Hyeon Kyun; Lee, Jeong Min; Chang, Jae Sook [KORTIC, Taejon (Korea, Republic of)

    2003-12-15

    Twelve kinds of environmental samples such as soil, seawater, underground water, etc. around Nuclear Power Plants(NPPs) were collected. Tritium chemical analysis was tried for the samples of rain water, pine-needle, air, seawater, underground water, chinese cabbage, again of rice and milk sampled around NPPs, and surface seawater and rain water sampled over the country. Strontium in the soil that were sampled at 60 point of district in Korea were analyzed. Tritium were analyzed in 21 samples of surface seawater around the Korea peninsular that were supplied form KFRDI(National Fisheries Research and Development Institute). Sampling and chemical analysis environmental samples around Kori, Woolsung, Youngkwang, Wooljin NPPs and Taeduk science town for tritium and strontium analysis was managed according to plans. Succeed to KINS after all samples were tried.

  1. Grain-size analysis of sediment cores collected in 2009 offshore from Palos Verdes, California

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This part of the data release includes grain-size analysis of sediment cores collected in 2009 offshore of Palos Verdes, California. It is one of seven files...

  2. Fukushima Daiichi Unit 1 Uncertainty Analysis-Exploration of Core Melt Progression Uncertain Parameters-Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brooks, Dusty Marie [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    Sandia National Laboratories (SNL) has conducted an uncertainty analysi s (UA) on the Fukushima Daiichi unit (1F1) accident progression wit h the MELCOR code. Volume I of the 1F1 UA discusses the physical modeling details and time history results of the UA. Volume II of the 1F1 UA discusses the statistical viewpoint. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). The goal of this work was to perform a focused evaluation of uncertainty in core damage progression behavior and its effect on key figures - of - merit (e.g., hydrogen production, fraction of intact fuel, vessel lower head failure) and in doing so assess the applicability of traditional sensitivity analysis techniques .

  3. 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.

  4. Coupled-mode analysis for single-helix chiral fiber gratings with small core-offset

    Institute of Scientific and Technical Information of China (English)

    Li Yang; Linlin Xue; Jue Su; Jingren Qian

    2011-01-01

    Using conventional coupled-mode theory,a set of coupled-mode equations are formulated for single-helix chiral fiber long-period gratings.A helical-core fiber is analyzed as an example.The analysis is simple in mathematical form and intuitive in physical concept.Based on the analysis,the polarization independence of mode coupling in special fiber gratings is revealed.The transmission characteristics of helical-core fibers are also simulated and discussed.

  5. Nuclear Forensic Science: Analysis of Nuclear Material Out of Regulatory Control

    Science.gov (United States)

    Kristo, Michael J.; Gaffney, Amy M.; Marks, Naomi; Knight, Kim; Cassata, William S.; Hutcheon, Ian D.

    2016-06-01

    Nuclear forensic science seeks to identify the origin of nuclear materials found outside regulatory control. It is increasingly recognized as an integral part of a robust nuclear security program. This review highlights areas of active, evolving research in nuclear forensics, with a focus on analytical techniques commonly employed in Earth and planetary sciences. Applications of nuclear forensics to uranium ore concentrates (UOCs) are discussed first. UOCs have become an attractive target for nuclear forensic researchers because of the richness in impurities compared to materials produced later in the fuel cycle. The development of chronometric methods for age dating nuclear materials is then discussed, with an emphasis on improvements in accuracy that have been gained from measurements of multiple radioisotopic systems. Finally, papers that report on casework are reviewed, to provide a window into current scientific practice.

  6. High-order harmonic generation from polyatomic molecules including nuclear motion and a nuclear modes analysis

    DEFF Research Database (Denmark)

    Madsen, Christian Bruun; Abu-Samha, Mahmoud; Madsen, Lars Bojer

    2010-01-01

    We present a generic approach for treating the effect of nuclear motion in high-order harmonic generation from polyatomic molecules. Our procedure relies on a separation of nuclear and electron dynamics where we account for the electronic part using the Lewenstein model and nuclear motion enters...... as a nuclear correlation function. We express the nuclear correlation function in terms of Franck-Condon factors, which allows us to decompose nuclear motion into modes and identify the modes that are dominant in the high-order harmonic generation process. We show results for the isotopes CH4 and CD4...... and thereby provide direct theoretical support for a recent experiment [S. Baker et al., Science 312, 424 (2006)] that uses high-order harmonic generation to probe the ultrafast structural nuclear rearrangement of ionized methane....

  7. 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.

  8. Turbopump Design and Analysis Approach for Nuclear Thermal Rockets

    Science.gov (United States)

    Chen, Shu-cheng S.; Veres, Joseph P.; Fittje, James E.

    2006-01-01

    A rocket propulsion system, whether it is a chemical rocket or a nuclear thermal rocket, is fairly complex in detail but rather simple in principle. Among all the interacting parts, three components stand out: they are pumps and turbines (turbopumps), and the thrust chamber. To obtain an understanding of the overall rocket propulsion system characteristics, one starts from analyzing the interactions among these three components. It is therefore of utmost importance to be able to satisfactorily characterize the turbopump, level by level, at all phases of a vehicle design cycle. Here at NASA Glenn Research Center, as the starting phase of a rocket engine design, specifically a Nuclear Thermal Rocket Engine design, we adopted the approach of using a high level system cycle analysis code (NESS) to obtain an initial analysis of the operational characteristics of a turbopump required in the propulsion system. A set of turbopump design codes (PumpDes and TurbDes) were then executed to obtain sizing and performance characteristics of the turbopump that were consistent with the mission requirements. A set of turbopump analyses codes (PUMPA and TURBA) were applied to obtain the full performance map for each of the turbopump components; a two dimensional layout of the turbopump based on these mean line analyses was also generated. Adequacy of the turbopump conceptual design will later be determined by further analyses and evaluation. In this paper, descriptions and discussions of the aforementioned approach are provided and future outlooks are discussed.

  9. Turbopump Design and Analysis Approach for Nuclear Thermal Rockets

    Science.gov (United States)

    Chen, Shu-Cheng S.; Veres, Joseph P.; Fittje, James E.

    2006-01-01

    A rocket propulsion system, whether it is a chemical rocket or a nuclear thermal rocket, is fairly complex in detail but rather simple in principle. Among all the interacting parts, three components stand out: they are pumps & turbines (turbopumps), and the thrust chamber. To obtain an understanding of the overall rocket propulsion system characteristics, one starts from analyzing the interactions among these three components. It is therefore of utmost importance to be able to satisfactorily characterize the turbopump, level by level, at all phases of a vehicle design cycle. Here at the NASA Glenn Research Center, as the starting phase of a rocket engine design, specifically a Nuclear Thermal Rocket Engine design, we adopted the approach of using a high level system cycle analysis code (NESS) to obtain an initial analysis of the operational characteristics of a turbopump required in the propulsion system. A set of turbopump design codes (PumpDes and TurbDes) were then executed to obtain sizing and performance parameters of the turbopump that were consistent with the mission requirements. A set of turbopump analyses codes (PUMPA and TURBA) were applied to obtain the full performance map for each of the turbopump components; a two dimensional layout of the turbopump based on these mean line analyses was also generated. Adequacy of the turbopump conceptual design will later be determined by further analyses and evaluation. In this paper, descriptions and discussions of the aforementioned approach are provided and future outlooks are discussed.

  10. Multi-Detector Analysis System for Spent Nuclear Fuel Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Reber, Edward Lawrence; Aryaeinejad, Rahmat; Cole, Jerald Donald; Drigert, Mark William; Jewell, James Keith; Egger, Ann Elizabeth; Cordes, Gail Adele

    1999-09-01

    The Spent Nuclear Fuel (SNF) Non-Destructive Analysis (NDA) program at INEEL is developing a system to characterize SNF for fissile mass, radiation source term, and fissile isotopic content. The system is based on the integration of the Fission Assay Tomography System (FATS) and the Gamma-Neutron Analysis Technique (GNAT) developed under programs supported by the DOE Office of Non-proliferation and National Security. Both FATS and GNAT were developed as separate systems to provide information on the location of special nuclear material in weapons configuration (FATS role), and to measure isotopic ratios of fissile material to determine if the material was from a weapon (GNAT role). FATS is capable of not only determining the presence and location of fissile material but also the quantity of fissile material present to within 50%. GNAT determines the ratios of the fissile and fissionable material by coincidence methods that allow the two prompt (immediately) produced fission fragments to be identified. Therefore, from the combination of FATS and GNAT, MDAS is able to measure the fissile material, radiation source term, and fissile isotopics content.

  11. Inner/Outer nuclear membrane fusion in nuclear pore assembly: biochemical demonstration and molecular analysis.

    Science.gov (United States)

    Fichtman, Boris; Ramos, Corinne; Rasala, Beth; Harel, Amnon; Forbes, Douglass J

    2010-12-01

    Nuclear pore complexes (NPCs) are large proteinaceous channels embedded in double nuclear membranes, which carry out nucleocytoplasmic exchange. The mechanism of nuclear pore assembly involves a unique challenge, as it requires creation of a long-lived membrane-lined channel connecting the inner and outer nuclear membranes. This stabilized membrane channel has little evolutionary precedent. Here we mapped inner/outer nuclear membrane fusion in NPC assembly biochemically by using novel assembly intermediates and membrane fusion inhibitors. Incubation of a Xenopus in vitro nuclear assembly system at 14°C revealed an early pore intermediate where nucleoporin subunits POM121 and the Nup107-160 complex were organized in a punctate pattern on the inner nuclear membrane. With time, this intermediate progressed to diffusion channel formation and finally to complete nuclear pore assembly. Correct channel formation was blocked by the hemifusion inhibitor lysophosphatidylcholine (LPC), but not if a complementary-shaped lipid, oleic acid (OA), was simultaneously added, as determined with a novel fluorescent dextran-quenching assay. Importantly, recruitment of the bulk of FG nucleoporins, characteristic of mature nuclear pores, was not observed before diffusion channel formation and was prevented by LPC or OA, but not by LPC+OA. These results map the crucial inner/outer nuclear membrane fusion event of NPC assembly downstream of POM121/Nup107-160 complex interaction and upstream or at the time of FG nucleoporin recruitment.

  12. Petrophysical analysis of limestone rocks by nuclear logging and 3D high-resolution X-ray computed microtomography

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, M.F.S. [Nuclear Instrumentation Laboratory, COPPE-PEN, UFRJ, P.O. Box 68509, 21941-972 Rio de Janeiro, RJ (Brazil); Lima, I., E-mail: inaya@lin.ufrj.br [Nuclear Instrumentation Laboratory, COPPE-PEN, UFRJ, P.O. Box 68509, 21941-972 Rio de Janeiro, RJ (Brazil); Department of Mechanical Engineering and Energy, IPRJ-UERJ, Nova Friburgo, RJ (Brazil); Ferrucio, P.L.; Abreu, C.J.; Borghi, L. [Geology Department, Geosciences Institute, Rio de Janeiro Federal University, Rio de Janeiro, RJ (Brazil); Lopes, R.T. [Nuclear Instrumentation Laboratory, COPPE-PEN, UFRJ, P.O. Box 68509, 21941-972 Rio de Janeiro, RJ (Brazil)

    2011-10-01

    This study presents the pore-space system analysis of the 2-ITAB-1-RJ well cores, which were drilled in the Sao Jose do Itaborai Basin, in the state of Rio de Janeiro, Brasil. The analysis presented herein has been developed based on two techniques: nuclear logging and 3D high-resolution X-ray computed microtomography. Nuclear logging has been proven to be the technique that provides better quality and more quantitative information about the porosity using radioactive sources. The Density Gamma Probe and the Neutron Sonde used in this work provide qualitative information about bulk density variations and compensated porosity of the geological formation. The samples obtained from the well cores were analyzed by microtomography. The use of this technique in sedimentary rocks allows quantitative evaluation of pore system and generates high-resolution 3D images ({approx}microns order). The images and data obtained by microtomography were integrated with the response obtained by nuclear logging. The results obtained by these two techniques allow the understanding of the pore-size distribution and connectivity, as well as the porosity values. Both techniques are important and they complement each other.

  13. Qualitative diagnosis for transients analysis on nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lorre, J.P.; Dorlet, E.; Evrard, J.M.

    1995-12-31

    One of the major aims of an intelligent monitoring system, is the supervision task which assist the operator in understanding what occurs on a process. Failures hypotheses must be located and the inferring process must be explained. This paper demonstrate a second generation expert system (SEXTANT) decided to the transients analysis on PWR nuclear reactors. This system detects failures by simulating the process with a numerical model. A diagnosis module uses an even graph built from a causal graph model of the plant to generate hypotheses, and a numerical model to validate these hypotheses. Hypotheses are stored into scenarios which are concurrent possible interpretations of the process evolution. The approach is illustrated by an application for the analysis of the house load operation on a pressurized water reactor. (authors). 9 refs., 10 figs.

  14. Sensitivity analysis for reliable design verification of nuclear turbosets

    Energy Technology Data Exchange (ETDEWEB)

    Zentner, Irmela, E-mail: irmela.zentner@edf.f [Lamsid-Laboratory for Mechanics of Aging Industrial Structures, UMR CNRS/EDF, 1, avenue Du General de Gaulle, 92141 Clamart (France); EDF R and D-Structural Mechanics and Acoustics Department, 1, avenue Du General de Gaulle, 92141 Clamart (France); Tarantola, Stefano [Joint Research Centre of the European Commission-Institute for Protection and Security of the Citizen, T.P. 361, 21027 Ispra (Italy); Rocquigny, E. de [Ecole Centrale Paris-Applied Mathematics and Systems Department (MAS), Grande Voie des Vignes, 92 295 Chatenay-Malabry (France)

    2011-03-15

    In this paper, we present an application of sensitivity analysis for design verification of nuclear turbosets. Before the acquisition of a turbogenerator, energy power operators perform independent design assessment in order to assure safe operating conditions of the new machine in its environment. Variables of interest are related to the vibration behaviour of the machine: its eigenfrequencies and dynamic sensitivity to unbalance. In the framework of design verification, epistemic uncertainties are preponderant. This lack of knowledge is due to inexistent or imprecise information about the design as well as to interaction of the rotating machinery with supporting and sub-structures. Sensitivity analysis enables the analyst to rank sources of uncertainty with respect to their importance and, possibly, to screen out insignificant sources of uncertainty. Further studies, if necessary, can then focus on predominant parameters. In particular, the constructor can be asked for detailed information only about the most significant parameters.

  15. A probabilistic safety analysis of incidents in nuclear research reactors.

    Science.gov (United States)

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  16. Characterization of Nuclear Fuel using Multivariate Statistical Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Robel, M; Robel, M; Robel, M; Kristo, M J; Kristo, M J

    2007-11-27

    Various combinations of reactor type and fuel composition have been characterized using principle components analysis (PCA) of the concentrations of 9 U and Pu isotopes in the 10 fuel as a function of burnup. The use of PCA allows the reduction of the 9-dimensional data (isotopic concentrations) into a 3-dimensional approximation, giving a visual representation of the changes in nuclear fuel composition with burnup. Real-world variation in the concentrations of {sup 234}U and {sup 236}U in the fresh (unirradiated) fuel was accounted for. The effects of reprocessing were also simulated. The results suggest that, 15 even after reprocessing, Pu isotopes can be used to determine both the type of reactor and the initial fuel composition with good discrimination. Finally, partial least squares discriminant analysis (PSLDA) was investigated as a substitute for PCA. Our results suggest that PLSDA is a better tool for this application where separation between known classes is most important.

  17. Development of Few Group Cross Section Calculation Model for Core Nuclear Design Code CYCAS%堆芯核设计程序CYCAS少群截面模型开发

    Institute of Scientific and Technical Information of China (English)

    杨伟焱; 汤春桃; 毕光文; 杨波

    2016-01-01

    少群截面模型为堆芯三维扩散计算提供实时的节块均匀少群截面,是堆芯计算程序的关键模型之一.CYCAS程序是上海核工程研究设计院最新开发的堆芯三维核设计程序.本文在详细解析影响节块截面的各种因素的基础上,提出应用于CYCAS程序的少群截面的模型.该模型采用能谱修正方法处理由于能谱变化所引入的二次效应,采用微观燃耗修正方法处理燃耗历史效应.单组件和A P1000核电厂的数值验证计算表明,该模型具有很高的计算精度.%The few group cross section calculation model generates node homogeneous few group cross section for core 3D diffusion calculation ,w hich is one of the key models of core calculation code .CYCAS is the new core 3D nuclear design code developed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI) .A new model based on detail analysis of the factors affecting node cross section was developed for CYCAS .In the model ,the energy spectrum correction method was used to process the second order effect introduced by energy spectrum change , and the micro-depletion correction method was utilized to treat depletion history effect .The numerical results of unit assembly and AP1000 core validate the high accuracy of the new model within CYCAS .

  18. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  19. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  20. Comparative bacterial proteomics: analysis of the core genome concept.

    Directory of Open Access Journals (Sweden)

    Stephen J Callister

    Full Text Available While comparative bacterial genomic studies commonly predict a set of genes indicative of common ancestry, experimental validation of the existence of this core genome requires extensive measurement and is typically not undertaken. Enabled by an extensive proteome database developed over six years, we have experimentally verified the expression of proteins predicted from genomic ortholog comparisons among 17 environmental and pathogenic bacteria. More exclusive relationships were observed among the expressed protein content of phenotypically related bacteria, which is indicative of the specific lifestyles associated with these organisms. Although genomic studies can establish relative orthologous relationships among a set of bacteria and propose a set of ancestral genes, our proteomics study establishes expressed lifestyle differences among conserved genes and proposes a set of expressed ancestral traits.

  1. Comparative Bacterial Proteomics: Analysis of the Core Genome Concept

    Energy Technology Data Exchange (ETDEWEB)

    Callister, Stephen J.; McCue, Lee Ann; Turse, Josh E.; Monroe, Matthew E.; Auberry, Kenneth J.; Smith, Richard D.; Adkins, Joshua N.; Lipton, Mary S.

    2008-02-06

    Comparative bacterial genomic studies commonly predict a set of genes indicative of common ancestry. Experimental validation of the existence of this core genome requires extensive measurement and is not typically undertaken. Enabled by an extensive proteome database development over a six year period, we experimentally verified the expression of proteins predicted from genomic ortholog comparisons among 17 environmental and pathogenic bacteria. More exclusive relationships were observed among the expressed protein content of phenotypically related bacteria, which is indicative of the specific lifestyles associated with these organisms. While genomic studies establish relative orthologous relationships among a set of bacteria and propose a set of ancestral genes, our proteomics study establishes expressed lifestyle differences among conserved genes and proposes a set of expressed ancestral traits.

  2. Dendritic Structure Analysis of CMSX-4 Cored Turbine Blades Roots

    Directory of Open Access Journals (Sweden)

    Krawczyk J.

    2016-06-01

    Full Text Available The microstructure of as-cast cored turbine blades roots, made of the single-crystal CMSX-4 nickel-based superalloy was investigated. Analysed blades were obtained by directional solidification technique in the industrial ALD Bridgman induction furnace. The investigations of the microstructure of blades roots were performed using SEM and X-ray techniques including diffraction topography with the use of Auleytner method. Characteristic shapes of dendrites with various arrangement were observed on the SEM images taken from the cross-sections, made transversely to the main blades axis. The differences in quality of the structure in particular areas of blades roots were revealed. Based on the results, the influence of cooling bores on blades root structure was analysed and the changes in the distribution and geometry of cooling bores were proposed.

  3. Comparative Bacterial Proteomics: Analysis of the Core Genome Concept

    Science.gov (United States)

    Callister, Stephen J.; McCue, Lee Ann; Turse, Joshua E.; Monroe, Matthew E.; Auberry, Kenneth J.; Smith, Richard D.; Adkins, Joshua N.; Lipton, Mary S.

    2008-01-01

    While comparative bacterial genomic studies commonly predict a set of genes indicative of common ancestry, experimental validation of the existence of this core genome requires extensive measurement and is typically not undertaken. Enabled by an extensive proteome database developed over six years, we have experimentally verified the expression of proteins predicted from genomic ortholog comparisons among 17 environmental and pathogenic bacteria. More exclusive relationships were observed among the expressed protein content of phenotypically related bacteria, which is indicative of the specific lifestyles associated with these organisms. Although genomic studies can establish relative orthologous relationships among a set of bacteria and propose a set of ancestral genes, our proteomics study establishes expressed lifestyle differences among conserved genes and proposes a set of expressed ancestral traits. PMID:18253490

  4. Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

    Science.gov (United States)

    Shirvan, Koroush

    Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

  5. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  6. Thermal analysis of HTS air-core transformer used in voltage compensation type active SFCL

    Science.gov (United States)

    Song, M.; Tang, Y.; Li, J.; Zhou, Y.; Chen, L.; Ren, L.

    2010-11-01

    The three-phase voltage compensation type active superconducting fault current limiter (SFCL) is composed of three HTS air-core transformers and a three-phase four-wire Pulse Width Modulation (PWM) converter. The primary winding of the each phase HTS air-core transformer is in series with the main system, and the second winding is connected with the PWM converter. The single-phase conduction-cooled HTS air-core transformer is consisting of four double-pancakes wound by the Bi2223/Ag tape. In this paper, according to the electromagnetic analysis on the single-phase HTS air-core transformer, its AC loss corresponding to different operation modes is calculated. Furthermore, the thermal behaviors are studied by the time-stepping numerical simulations. On the basis of the simulation results, the related problems with the HTS air-core transformer's thermal stability are discussed.

  7. Two dimensional dynamic analysis of sandwich plates with gradient foam cores

    Energy Technology Data Exchange (ETDEWEB)

    Mu, Lin; Xiao, Deng Bao; Zhao, Guiping [State Key Laboratory for Mechanical structure Strength and Vibration, School of AerospaceXi' an Jiaotong University, Xi' an (China); Cho, Chong Du [Dept. of Mechanical Engineering, Inha University, Inchon (Korea, Republic of)

    2016-09-15

    Present investigation is concerned about dynamic response of composite sandwich plates with the functionally gradient foam cores under time-dependent impulse. The analysis is based on a model of the gradient sandwich plate, in which the face sheets and the core adopt the Kirchhoff theory and a [2, 1]-order theory, respectively. The material properties of the gradient foam core vary continuously along the thickness direction. The gradient plate model is validated with the finite element code ABAQUS®. And the results show that the proposed model can predict well the free vibration of composite sandwich plates with gradient foam cores. The influences of gradient foam cores on the natural frequency, deflection and energy absorbing of the sandwich plates are also investigated.

  8. Improvement of Axial Reflector Cross Section Generation Model for PWR Core Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Cheon Bo; Lee, Kyung Hoon; Cho, Jin Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This paper covers the study for improvement of axial reflector XS generation model. In the next section, the improved 1D core model is represented in detail. Reflector XS generated by the improved model is compared to that of the conventional model in the third section. Nuclear design parameters generated by these two XS sets are also covered in that section. Significant of this study is discussed in the last section. Two-step procedure has been regarded as the most practical approach for reactor core designs because it offers core design parameters quite rapidly within acceptable range. Thus this approach is adopted for SMART (System-integrated Modular Advanced Reac- Tor) core design in KAERI with the DeCART2D1.1/ MASTER4.0 (hereafter noted as DeCART2D/ MASTER) code system. Within the framework of the two-step procedure based SMART core design, various researches have been studied to improve the core design reliability and efficiency. One of them is improvement of reflector cross section (XS) generation models. While the conventional FA/reflector two-node model used for most core designs to generate reflector XS cannot consider the actual configuration of fuel rods that intersect at right angles to axial reflectors, the revised model reflects the axial fuel configuration by introducing the radially simplified core model. The significance of the model revision is evaluated by observing HGC generated by DeCART2D, reflector XS, and core design parameters generated by adopting the two models. And it is verified that about 30 ppm CBC error can be reduced and maximum Fq error decreases from about 6 % to 2.5 % by applying the revised model. Error of AO and axial power shapes are also reduced significantly. Therefore it can be concluded that the simplified 1D core model improves the accuracy of the axial reflector XS and leads to the two-step procedure reliability enhancement. Since it is hard for core designs to be free from the two-step approach, it is necessary to find

  9. The Activities of the European Consortium on Nuclear Data Development and Analysis for Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physic and Reactor Technology, 76344 Eggenstein-Leopoldshafen (Germany); Avrigeanu, M.; Avrigeanu, V. [Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), RO-077125 Magurele (Romania); Cabellos, O. [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Kodeli, I. [Jozef Stefan Institute (JSI), Jamova 39, 1000 Ljubljana (Slovenia); Koning, A. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 LE Petten (Netherlands); Konobeyev, A.Yu. [Karlsruhe Institute of Technology, Institute for Neutron Physic and Reactor Technology, 76344 Eggenstein-Leopoldshafen (Germany); Leeb, H. [Technische Universitaet Wien, Atominstitut, Wiedner Hauptstrasse 8–10, 1040 Wien (Austria); Rochman, D. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 LE Petten (Netherlands); Pereslavtsev, P. [Karlsruhe Institute of Technology, Institute for Neutron Physic and Reactor Technology, 76344 Eggenstein-Leopoldshafen (Germany); Sauvan, P. [Universidad Nacional de Educacion a Distancia, C. Juan del Rosal, 12, 28040 Madrid (Spain); Sublet, J.-C. [Euratom/CCFE Fusion Association, Culham Science Centre, OX14 3DB (United Kingdom); Trkov, A. [Jozef Stefan Institute (JSI), Jamova 39, 1000 Ljubljana (Slovenia); Dupont, E. [OECD Nuclear Energy Agency, Paris (France); Leichtle, D.; Izquierdo, J. [Fusion for Energy, Barcelona (Spain)

    2014-06-15

    This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations.

  10. 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.

  11. Identifying and Tracking Individual Updraft Cores using Cluster Analysis: A TWP-ICE case study

    Science.gov (United States)

    Li, X.; Tao, W.; Collis, S. M.; Varble, A.

    2013-12-01

    Cumulus parameterizations in GCMs depend strongly on the vertical velocity structures of convective updraft cores, or plumes. There hasn't been an accurate way of identifying these cores. The majority of previous studies treat the updraft as a single grid column entity, thus missing many intrinsic characteristics, e.g., the size, strength and spatial orientation of an individual core, its life cycle, and the time variations of the entrainment/detrainment rates associated with its life cycle. In this study, we attempt to apply an innovative algorithm based on the centroid-based k-means cluster analysis to improve our understanding of convection and its associated updraft cores. Both 3-D Doppler radar retrievals and cloud-resolving model simulations of a TWP-ICE campaign case during the monsoon period will be used to test and improve this algorithm. This will provide for more in-depth comparisons between CRM simulations and observations that were not possible previously using the traditional piecewise analysis with each updraft column. The first step is to identify the strongest cores (maximum velocity >10 m/s), since they are well defined and produce definite answers when the cluster analysis algorithm is applied. The preliminary results show that the radar retrieved updraft cores are smaller in size and with the maximum velocity located uniformly at higher levels compared with model simulations. Overall, the model simulations produce much stronger cores compared with the radar retrievals. Within the model simulations, the bulk microphysical scheme simulation produces stronger cores than the spectral bin microphysical scheme. Planned researches include using high temporal-resolution simulations to further track the life cycle of individual updraft cores and study their characteristics.

  12. Large-Scale Weibull Analysis of H-451 Nuclear- Grade Graphite Specimen Rupture Data

    Science.gov (United States)

    Nemeth, Noel N.; Walker, Andrew; Baker, Eric H.; Murthy, Pappu L.; Bratton, Robert L.

    2012-01-01

    A Weibull analysis was performed of the strength distribution and size effects for 2000 specimens of H-451 nuclear-grade graphite. The data, generated elsewhere, measured the tensile and four-point-flexure room-temperature rupture strength of specimens excised from a single extruded graphite log. Strength variation was compared with specimen location, size, and orientation relative to the parent body. In our study, data were progressively and extensively pooled into larger data sets to discriminate overall trends from local variations and to investigate the strength distribution. The CARES/Life and WeibPar codes were used to investigate issues regarding the size effect, Weibull parameter consistency, and nonlinear stress-strain response. Overall, the Weibull distribution described the behavior of the pooled data very well. However, the issue regarding the smaller-than-expected size effect remained. This exercise illustrated that a conservative approach using a two-parameter Weibull distribution is best for designing graphite components with low probability of failure for the in-core structures in the proposed Generation IV (Gen IV) high-temperature gas-cooled nuclear reactors. This exercise also demonstrated the continuing need to better understand the mechanisms driving stochastic strength response. Extensive appendixes are provided with this report to show all aspects of the rupture data and analytical results.

  13. Seismic margin analysis technique for nuclear power plant structures

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jeong Moon; Choi, In Kil

    2001-04-01

    In general, the Seismic Probabilistic Risk Assessment (SPRA) and the Seismic Margin Assessment(SAM) are used for the evaluation of realistic seismic capacity of nuclear power plant structures. Seismic PRA is a systematic process to evaluate the seismic safety of nuclear power plant. In our country, SPRA has been used to perform the probabilistic safety assessment for the earthquake event. SMA is a simple and cost effective manner to quantify the seismic margin of individual structural elements. This study was performed to improve the reliability of SMA results and to confirm the assessment procedure. To achieve this goal, review for the current status of the techniques and procedures was performed. Two methodologies, CDFM (Conservative Deterministic Failure Margin) sponsored by NRC and FA (Fragility Analysis) sponsored by EPRI, were developed for the seismic margin review of NPP structures. FA method was originally developed for Seismic PRA. CDFM approach is more amenable to use by experienced design engineers including utility staff design engineers. In this study, detailed review on the procedures of CDFM and FA methodology was performed.

  14. Summary of Prometheus Radiation Shielding Nuclear Design Analysis

    Energy Technology Data Exchange (ETDEWEB)

    J. Stephens

    2006-01-13

    This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL & Bettis) shielding nuclear design analyses done for the project.

  15. the international politics of nuclear weapons: a constructivist analysis

    African Journals Online (AJOL)

    JK

    interest including India, Pakistan, China, North Korea and the US; and .... and conventions against the development, stockpiling and use of nuclear weapons. ...... process, the Agency discovered illicit nuclear procurement networks (UN News.

  16. Deconvolution-based resolution enhancement of chemical ice core records obtained by continuous flow analysis

    DEFF Research Database (Denmark)

    Rasmussen, Sune Olander; Andersen, Katrine K.; Johnsen, Sigfus Johann

    2005-01-01

    Continuous flow analysis (CFA) has become a popular measuring technique for obtaining high-resolution chemical ice core records due to an attractive combination of measuring speed and resolution. However, when analyzing the deeper sections of ice cores or cores from low-accumulation areas......, there is still need for further improvement of the resolution. Here a method for resolution enhancement of CFA data is presented. It is demonstrated that it is possible to improve the resolution of CFA data by restoring some of the detail that was lost in the measuring process, thus improving the usefulness...

  17. Using Profile Analysis via Multidimensional Scaling (PAMS) to identify core profiles from the WMS-III.

    Science.gov (United States)

    Frisby, Craig L; Kim, Se-Kang

    2008-03-01

    Profile Analysis via Multidimensional Scaling (PAMS) is a procedure for extracting latent core profiles in a multitest data set. The PAMS procedure offers several advantages compared with other profile analysis procedures. Most notably, PAMS estimates individual profile weights that reflect the degree to which an individual's observed profile approximates the shape and scatter of latent core profiles. The PAMS procedure was applied to index scores of nonreplicated participants from the standardization sample (N = 1,033) for the Wechsler Memory Scale--Third Edition (D. Tulsky, J. Zhu, & M. F. Ledbetter, 2002). PAMS extracted discrepant visual memory and auditory memory versus working memory core profiles for the complete 16- to 89-year-old sample and discrepant working memory and auditory memory versus working memory core profiles for the 75- to 89-year-old cohort. Implications for use of PAMS in future research are discussed.

  18. Coherent Network Analysis of Gravitational Waves from Three-Dimensional Core-Collapse Supernova Models

    CERN Document Server

    Hayama, Kazuhiro; Kotake, Kei; Takiwaki, Tomoya

    2015-01-01

    Using predictions from three-dimensional (3D) hydrodynamics simulations of core-collapse supernovae (CCSNe), we present a coherent network analysis to detection, reconstruction, and the source localization of the gravitational-wave (GW) signals. By combining with the GW spectrogram analysis, we show that several important hydrodynamics features imprinted in the original waveforms persist in the waveforms of the reconstructed signals. The characteristic excess in the GW spectrograms originates not only from rotating core-collapse and bounce, the subsequent ring down of the proto-neutron star (PNS) as previously identified, but also from the formation of magnetohydrodynamics jets and non-axisymmetric instabilities in the vicinity of the PNS. Regarding the GW signals emitted near at the rotating core bounce, the horizon distance, which we set by a SNR exceeding 8, extends up to $\\sim$ 18 kpc for the most rapidly rotating 3D model among the employed waveform libraries. Following the rotating core bounce, the domi...

  19. Nuclear knowledge-management. A core competence of VGB; Uebergreifendes Wissensmanagement fuer Kernkraftwerke. Eine VGB-Kernkompetenz

    Energy Technology Data Exchange (ETDEWEB)

    Pamme, Hartmut [RWE Power AG, Essen (Germany). Steuerung Kernkraftwerke

    2009-07-01

    It is a well established expectation that utilities/operators of nuclear power plants communicate their own operational situation and are able to comment promptly on any findings and events in the international nuclear scene. In order to gain synergies on knowledge management, utilities have been using VGB as common platform for many years. The paper describes the generic expectations concerning knowledge management towards an association like VGB. It is analysed which elements and peculiarities of modern knowledge management are already established within VGB in the nuclear field. (orig.)

  20. Sensitivity analysis of parameters important to nuclear criticality safety of Castor X/28F spent nuclear fuel cask

    Energy Technology Data Exchange (ETDEWEB)

    Leotlela, Mosebetsi J. [Witwatersrand Univ., Johannesburg (South Africa). School of Physics; Koeberg Operating Unit, Johannesburg (South Africa). Regulations and Licensing; Malgas, Isaac [Koeberg Nuclear Power Station, Duinefontein (South Africa). Nuclear Engineering Analysis; Taviv, Eugene [ASARA consultants (PTY) LTD, Johannesburg (South Africa)

    2015-11-15

    In nuclear criticality safety analysis it is essential to ascertain how various components of the nuclear system will perform under certain conditions they may be subjected to, particularly if the components of the system are likely to be affected by environmental factors such as temperature, radiation or material composition. It is therefore prudent that a sensitivity analysis is performed to determine and quantify the response of the output to variation in any of the input parameters. In a fissile system, the output parameter of importance is the k{sub eff}. Therefore, in attempting to prevent reactivity-induced accidents, it is important for the criticality safety analyst to have a quantified degree of response for the neutron multiplication factor to perturbation in a given input parameter. This article will present the results of the perturbation of the parameters that are important to nuclear criticality safety analysis and their respective correlation equations for deriving the sensitivity coefficients.

  1. ASTEC adaptation for PHWR limited core damage accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, P., E-mail: pmajum@barc.gov.in [Bhabha Atomic Research Centre, Reactor Safety Division, Mumbai 400085 (India); Chatterjee, B.; Lele, H.G. [Bhabha Atomic Research Centre, Reactor Safety Division, Mumbai 400085 (India); Guillard, G.; Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, 13115 Saint-Paul-lez-Durance (France)

    2014-06-01

    Under limited core damage accidents (LCDAs) of Pressurized Heavy Water Reactor (PHWR), coolable geometry of the channel might be retained thanks to the presence of moderator heat sink. Indeed, the pressure tube is amenable to creep deformation at high temperature due to internal pressure and fuel bundles weight. Partial or complete circumferential contact between pressure tube and calandria tube aids heat dissipation to the moderator. A new module has been developed by Bhabha Atomic Research Centre (BARC) for simulating this phenomenon which is specific to horizontal-type of reactors. It requires additional calculation of pressure tube sagging/ballooning and temperature field in the circumferential direction. The module is well validated with available experimental results concerning pressure tube deformation and the associated heat transfer in the area of contact. It is then used in analysing typical LCDAs scenarios in Indian PHWR under low and medium internal pressure conditions. This module is implemented in the ASTEC IRSN-GRS severe accident code version under development and will thus be available in the next major version V2.1.

  2. Thermal buckling analysis of truss-core sandwich plates

    Institute of Scientific and Technical Information of China (English)

    陈继伟; 刘咏泉; 刘伟; 苏先樾

    2013-01-01

    Truss-core sandwich plates have received much attention in virtue of the high values of strength-to-weight and stiffness-to-weight as well as the great ability of impulse-resistance recently. It is necessary to study the stability of sandwich panels under the influence of the thermal load. However, the sandwich plates are such complex three-dimensional (3D) systems that direct analytical solutions do not exist, and the finite element method (FEM) cannot represent the relationship between structural parameters and mechanical properties well. In this paper, an equivalent homogeneous continuous plate is idealized by obtaining the effective bending and transverse shear stiffness based on the characteristics of periodically distributed unit cells. The first order shear deformation theory for plates is used to derive the stability equation. The buckling temperature of a simply supported sandwich plate is given and verified by the FEM. The effect of related parameters on mechanical properties is investigated. The geometric parameters of the unit cell are optimized to attain the maximum buckling temperature. It is shown that the optimized sandwich plate can improve the resistance to thermal buckling significantly.

  3. Measurement of solid state nuclear tracks in apatite by thermal analysis method

    Institute of Scientific and Technical Information of China (English)

    HE ShaoRong; YANG TongSuo; LI TianXiang; LU BaiZuo; JI ShuLi; HENG ShuYun

    2009-01-01

    A new measurement method of thermal analysis for solid state nuclear tracks is proposed. The an-nealing heat emitted by the unit mass of solid state nuclear tracks in heavy particles of the sample is determined via micro-thermal analysis method. Hence, the number of solid state nuclear tracks in the unit mass of sample is determined. In particular, this paper introduces the method and its significance to measure the number of a-particles nuclear tracks in apatite by measuring the annealing heat of a-particles nuclear tracks. In addition, the mechanism of the measurement and potential applications are discussed.

  4. Distribution of radionuclides in a marine sediment core off the waterspout of the nuclear power plants in Daya Bay, northeastern South China Sea.

    Science.gov (United States)

    Zhou, Peng; Li, Dongmei; Li, Haitao; Fang, Hongda; Huang, Chuguang; Zhang, Yusheng; Zhang, Hongbiao; Zhao, Li; Zhou, Junjie; Wang, Hua; Yang, Jie

    2015-07-01

    A sediment core was collected and dated using (210)Pbex dating method off the waterspout of nuclear power base of Daya Bay, northeastern South China Sea. The γ-emitting radionuclides were analyzed using HPGe γ spectrometry, gross alpha and beta radioactivity as well as other geochemical indicators were deliberated to assess the impact of nuclear power plants (NPP) operation and to study the past environment changes. It suggested that NPP provided no new radioactivity source to sediment based on the low specific activity of (137)Cs. Two broad peaks of TOC, TC and LOI accorded well with the commercial operations of Daya Bay NPP (1994.2 and 1994.5) and LNPP Phase I (2002.5 and 2003.3), implying that the mass input of cooling water from NPP may result into a substantial change in the ecological environment and Daya Bay has been severely impacted by human activities.

  5. Modeling of Plutonium Ionization Probabilities for Use in Nuclear Forensic Analysis by Resonance Ionization Mass Spectrometry

    Science.gov (United States)

    2016-12-01

    material and chemical composition within the sample. This data can then be included in analysis by law enforcement and intelligence agencies to...PLUTONIUM IONIZATION PROBABILITIES FOR USE IN NUCLEAR FORENSIC ANALYSIS BY RESONANCE IONIZATION MASS SPECTROMETRY by Steven F. Hutchinson...IONIZATION PROBABILITIES FOR USE IN NUCLEAR FORENSIC ANALYSIS BY RESONANCE IONIZATION MASS SPECTROMETRY 5. FUNDING NUMBERS 6. AUTHOR(S) Steven F

  6. Feasibility of probing solid state nuclear tracks by thermal analysis method

    Institute of Scientific and Technical Information of China (English)

    YANG TongSuo; ZHOU Bing; YANG XinXin; HE ShaoRong; HENG ShuYun; YUAN SunSheng

    2007-01-01

    The feasibility of probing solid state nuclear tracks by thermal analysis method is discussed both theoretically and experimentally. Comparison is made between the thermal analysis method and the optical microscope method, and it is demonstrated that this thermal analysis method is applicable to probing solid state nuclear tracks.

  7. Steady state thermal hydraulic analysis of LMR core using COBRA-K code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol

    1997-02-01

    A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.

  8. Chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade uranyl nitrate solutions

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

    The standard covers analytical procedures to determine compliance of nuclear-grade uranyl nitrate solution to specifications. The following methods are described in detail: uranium by ferrous sulfate reduction-potassium dichromate titrimetry and by ignition gravimetry; specific gravity by pycnometry; free acid by oxalate complexation; thorium by the Arsenazo(III) (photometric) method; chromium by the diphenylcarbazide (photometric) method; molybdenum by the thiocyanate (photometric) method; halogens separation by steam distillation; fluorine by specific ion electrode; halogen distillate analysis: chloride, bromide and iodide by amperometric microtitrimetry; bromine by the fluorescein (photometric) method; sulfate sulfur by (photometric) turbidimetry; phosphorus by the molybdenum blue (photometric) method; silicon by the molybdenum blue (photometric) method; carbon by persulfate oxidation-acid titrimetry; nonvolatile impurities by spectrography; volatile impurities by rotating-disk spark spectrography; boron by emission spectrography; impurity elements by spark source mass spectrography; isotopic composition by multiple filament surface-ionization mass spectrometry; uranium-232 by alpha spectrometry; total alpha activity by direct alpha counting; fission product activity by beta and gamma counting; entrained organic matter by infrared spectrophotometry. (JMT)

  9. Helium mobility in SON68 borosilicate nuclear glass: A nuclear reaction analysis approach

    Energy Technology Data Exchange (ETDEWEB)

    Bès, R., E-mail: rene.bes@cnrs-orleans.fr [CNRS, UPR3079 CEMHTI, 1D Avenue de la Recherche Scientifique, 45071 Orléans cedex 2 (France); Sauvage, T. [CNRS, UPR3079 CEMHTI, 1D Avenue de la Recherche Scientifique, 45071 Orléans cedex 2 (France); Université d’Orléans, Faculté des Sciences, Avenue du Parc Floral, BP 6749, 45067 Orléans cedex 2 (France); Peuget, S. [CEA/DEN/VRH/DTCD/SECM/LMPA Marcoule (France); Haussy, J. [CEA, DAM, DIF, F-91297 Arpajon (France); Chamssedine, F. [Université Libanaise, Faculté des Sciences V, Nabatiyeh (Lebanon); Oliviero, E. [CSNSM, CNRS/IN2P3 and Université Paris-Sud, Bât. 104-108, F-91405 Orsay (France); Fares, T. [CEA/DEN/VRH/DTCD/SECM/LMPA Marcoule (France); Vincent, L. [Institut d’Electronique Fondamentale, CNRS and Université Paris-Sud, UMR 8622, F-91405 Orsay (France)

    2013-11-15

    The {sup 3}He behavior in the non active R7T7 type borosilicate glass called SON68 has been investigated using the implantation method to introduce helium in the material. Nuclear Reaction Analysis (NRA) was performed to follow the helium concentration depth profile evolution as a function of annealing time and temperature. In addition, in situ Transmission Electron Microscopy (TEM) has been implemented to study the formation of helium bubbles during both implantation and annealing processes. Numerical modeling with two different approaches is proposed and discussed to investigate the helium mobility mechanisms. Our study reveals for helium incorporation by implantation at low temperature the presence of several helium populations with disparate diffusivities. The most mobile helium fraction would be attributed to atomic diffusion. The corresponding activation energy value (0.61 eV) extracted from Arrhenius graphs is in good agreement with literature data. The results also highlight that the damages associated to helium sursaturation are the source of small helium clusters formation, with a reduced mobility instead of the atomic mobility measured by the infusion technique. Small cavities that support this assumption have been observed by TEM at low temperature.

  10. Analysis of core samples from the BPXA-DOE-USGS Mount Elbert gas hydrate stratigraphic test well: Insights into core disturbance and handling

    Energy Technology Data Exchange (ETDEWEB)

    Kneafsey, Timothy J.; Lu, Hailong; Winters, William; Boswell, Ray; Hunter, Robert; Collett, Timothy S.

    2009-09-01

    Collecting and preserving undamaged core samples containing gas hydrates from depth is difficult because of the pressure and temperature changes encountered upon retrieval. Hydrate-bearing core samples were collected at the BPXA-DOE-USGS Mount Elbert Gas Hydrate Stratigraphic Test Well in February 2007. Coring was performed while using a custom oil-based drilling mud, and the cores were retrieved by a wireline. The samples were characterized and subsampled at the surface under ambient winter arctic conditions. Samples thought to be hydrate bearing were preserved either by immersion in liquid nitrogen (LN), or by storage under methane pressure at ambient arctic conditions, and later depressurized and immersed in LN. Eleven core samples from hydrate-bearing zones were scanned using x-ray computed tomography to examine core structure and homogeneity. Features observed include radial fractures, spalling-type fractures, and reduced density near the periphery. These features were induced during sample collection, handling, and preservation. Isotopic analysis of the methane from hydrate in an initially LN-preserved core and a pressure-preserved core indicate that secondary hydrate formation occurred throughout the pressurized core, whereas none occurred in the LN-preserved core, however no hydrate was found near the periphery of the LN-preserved core. To replicate some aspects of the preservation methods, natural and laboratory-made saturated porous media samples were frozen in a variety of ways, with radial fractures observed in some LN-frozen sands, and needle-like ice crystals forming in slowly frozen clay-rich sediments. Suggestions for hydrate-bearing core preservation are presented.

  11. Analysis of biological materials using a nuclear microprobe

    Science.gov (United States)

    Mulware, Stephen Juma

    The use of nuclear microprobe techniques including: Particle induced x-ray emission (PIXE) and Rutherford backscattering spectrometry (RBS) for elemental analysis and quantitative elemental imaging of biological samples is especially useful in biological and biomedical research because of its high sensitivity for physiologically important trace elements or toxic heavy metals. The nuclear microprobe of the Ion Beam Modification and Analysis Laboratory (IBMAL) has been used to study the enhancement in metal uptake of two different plants. The roots of corn (Zea mays) have been analyzed to study the enhancement of iron uptake by adding Fe (II) or Fe(III) of different concentrations to the germinating medium of the seeds. The Fe uptake enhancement effect produced by lacing the germinating medium with carbon nanotubes has also been investigated. The aim of this investigation is to ensure not only high crop yield but also Fe-rich food products especially from calcareous soil which covers 30% of world's agricultural land. The result will help reduce iron deficiency anemia, which has been identified as the leading nutritional disorder especially in developing countries by the World Health Organization. For the second plant, Mexican marigold (Tagetes erecta ), the effect of an arbuscular mycorrhizal fungi (Glomus intraradices ) for the improvement of lead phytoremediation of lead contaminated soil has been investigated. Phytoremediation provides an environmentally safe technique of removing toxic heavy metals (like lead), which can find their way into human food, from lands contaminated by human activities like mining or by natural disasters like earthquakes. The roots of Mexican marigold have been analyzed to study the role of arbuscular mycorrhizal fungi in enhancement of lead uptake from the contaminated rhizosphere.

  12. A High-Resolution Continuous Flow Analysis System for Polar Ice Cores

    DEFF Research Database (Denmark)

    Dallmayr, Remi; Goto-Azuma, Kumiko; Kjær, Helle Astrid;

    2016-01-01

    of Polar Research (NIPR) in Tokyo. The system allows the continuous analysis of stable water isotopes and electrical conductivity, as well as the collection of discrete samples from both inner and outer parts of the core. This CFA system was designed to have sufficiently high temporal resolution to detect......In recent decades, the development of continuous flow analysis (CFA) technology for ice core analysis has enabled greater sample throughput and greater depth resolution compared with the classic discrete sampling technique. We developed the first Japanese CFA system at the National Institute...

  13. 太阳中心核反应链中密度的时间变化%Time Variation of Density in the Core Nuclear Reaction Chain

    Institute of Scientific and Technical Information of China (English)

    杜九林; 葛德彪

    2000-01-01

    研究太阳中心的核反应对太阳中微子问题的解释至关重要,也是当前太阳振动测量中最为困难的问题.通过对p-pI核反应链中密度的时间变化研究发现,在从0.090 2R⊙到0.150 7R⊙的中心区域内,3He的粒子数密度恰好以5 min左右的周期随时间振荡.这类振荡可以引起太阳核能产生的周期变化 ,从而改变了太阳中微子的产生率.%The research into the central core nuclear re acti ons of the Sun plays a vital role in understanding the solar neutrino problem an d also is the most difficult problem in the helioseismic measurements. This pape r studies the time-dependent variation of density in the core p-pI nuclear re action chain. It is very coincidentally discovered that the particle number d ensity of 3He oscillates with a time period of the order of 5 minute i n the core layer ex tending from about 0.090 2 R⊙ to 0.150 7 R⊙. This oscillation can lead t o the similar time variation of energy generation of the core and therefore chan ge the production of the 8B and 7Be solar neutrino fluxes.

  14. Big data analysis of public acceptance of nuclear power in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Seung Kook [Policy Research Division, Korea Atomic Energy Research Institute (KAERI), Daejeon (Korea, Republic of)

    2017-06-15

    Public acceptance of nuclear power is important for the government, the major stakeholder of the industry, because consensus is required to drive actions. It is therefore no coincidence that the governments of nations operating nuclear reactors are endeavoring to enhance public acceptance of nuclear power, as better acceptance allows stable power generation and peaceful processing of nuclear wastes produced from nuclear reactors. Past research, however, has been limited to epistemological measurements using methods such as the Likert scale. In this research, we propose big data analysis as an attractive alternative and attempt to identify the attitudes of the public on nuclear power. Specifically, we used common big data analyses to analyze consumer opinions via SNS (Social Networking Services), using keyword analysis and opinion analysis. The keyword analysis identified the attitudes of the public toward nuclear power. The public felt positive toward nuclear power when Korea successfully exported nuclear reactors to the United Arab Emirates. With the Fukushima accident in 2011 and certain supplier scandals in 2012, however, the image of nuclear power was degraded and the negative image continues. It is recommended that the government focus on developing useful businesses and use cases of nuclear power in order to improve public acceptance.

  15. Analysis of Current Global Nuclear Safety and Security Cooperation

    Institute of Scientific and Technical Information of China (English)

    Liu; Chong

    2014-01-01

    Last year, global nuclear security and safety cooperation achieved some progress. In terms of nuclear safety, too many flaws are exposed by the current severe situation of the Fukushima in Japan’s new nuclear safety regulation system, and sound the alarm for East Asia countries accelerating the regional nuclear safety cooperation. In terms of nuclear security, since the Seoul Summit in March 2012, global nuclear security cooperation has achieved new successes. IAEA has and would play the central role in pushing forward the international framework and strengthening nuclear security globally. However, there are still some obstacles to overcome in the future, which need international society to enhance communication and common understanding, especially high-level consultations.

  16. Using Profile Analysis via Multidimensional Scaling (PAMS) to Identify Core Profiles from the WMS-III

    Science.gov (United States)

    Frisby, Craig L.; Kim, Se-Kang

    2008-01-01

    Profile Analysis via Multidimensional Scaling (PAMS) is a procedure for extracting latent core profiles in a multitest data set. The PAMS procedure offers several advantages compared with other profile analysis procedures. Most notably, PAMS estimates individual profile weights that reflect the degree to which an individual's observed profile…

  17. A Methodology for Loading the Advanced Test Reactor Driver Core for Experiment Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cowherd, Wilson M.; Nielsen, Joseph W.; Choe, Dong O.

    2016-11-01

    In support of experiments in the ATR, a new methodology was devised for loading the ATR Driver Core. This methodology will replace the existing methodology used by the INL Neutronic Analysis group to analyze experiments. Studied in this paper was the as-run analysis for ATR Cycle 152B, specifically comparing measured lobe powers and eigenvalue calculations.

  18. David Coleman and the Technologisation of the Common Core: A Critical Discourse Analysis

    Science.gov (United States)

    Johnson, Lindy L.

    2014-01-01

    Drawing on sociocultural perspectives and New Literacies Studies this study uses Critical Discourse Analysis (CDA) as a tool to closely analyse one way the Common Core State Standards in the United States are being produced, disseminated and consumed. The analysis focuses on a section of the CCSS, a model lesson given by one of the primary…

  19. Correction of Ex-core nuclear instrumentation measurement in mechanical shim mode%机械补偿模式下堆外核测测量结果修正

    Institute of Scientific and Technical Information of China (English)

    王瑞; 周赟

    2013-01-01

    Mechanical Shim(MSHIM) mode is used in AP1000 nuclear power plant to control core reactivity and power distribution during load maneuver operations.In MSHIM mode,the motion of MSHIM control rods changes the neutron flux distribution in core,and consequently changes the core power and its distribution.The relationship between the core power and the ex-core Power Range(PR) detector Axial Flux Difference(AFD) is affected significantly,and the current methodology that uses the ex-core PR detector AFD as the input makes too many errors in reactor protection set-point calculation.In this article,the process of measuring axial flux difference by the PR detectors is introduced,and the analysis of effect on AFD in MSHIM mode is discussed.At last,the mathematical modeling of the core power difference and the ex-core PR detector AFD is given to find a method to reduce the error.%AP1000采用机械补偿模式在负荷跟踪过程中对堆芯反应性和功率分布进行控制,在机械补偿模式下,由于机械补偿控制棒的移动,堆芯中子注量率分布发生变化,从而改变了堆芯功率及其分布,作为反应堆保护重要参数的堆外核测测量结果无法满足停堆保护的要求。本文主要介绍了堆外核测堆芯轴向功率偏差的测量方法及其在确定反应堆超温和超功率停堆保护整定值中的作用,分析了采用堆外核测功率量程探测器测量轴向功率偏差在机械补偿模式下受到的影响,提出一种修正测量结果的方法。

  20. Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

    Science.gov (United States)

    Butler, C.; Albright, D.

    2007-01-01

    Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

  1. Applications of nuclear microprobe analysis to dermatological research

    Science.gov (United States)

    Pallon, Jan; Forslind, Bo; Werner-Linde, Ylva; Yang, C.; Utui, R. J.; Elfman, M.; Malmqvist, K. G.; Kristiansson, P.; Sjöland, K. A.

    1997-07-01

    The elemental distributions over epidermal skin cross sections as revealed by nuclear microprobe analysis on cryo-sections from human skin provides new insight into the physiology of skin. Recently interest has been focused on the end stage of epidermal differentiation, the programmed cell death, occurring in the uppermost layer of the viable epidermis, the stratum granulosum. Calcium is one of the important messengers that controls the events of this programmed cell death, which shares a number of characteristics with apoptosis. We have previously shown that the Ca-gradient over normal skin cross sections is compatible with the finding from cell culture of epidermal cells which need a minimum level of 0.1 mM (Ca 2+) to develop a normal stratum corneum. To gain more information from the large number of data assessed during the actual analysis we have applied multivariate statistical analysis to the complete dataset obtained at NMP analyses. This statistical method reveals covariation of several elements and in addition provides a means to interpret the quantitative data in a meaningful biological context.

  2. First in-core simultaneous measurements of nuclear heating and thermal neutron flux obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lepeltier, Valerie; Bubendorff, Jacques; Carcreff, Hubert [Nuclear studies and reactor irradiation Service, CEA Saclay 91191 Gif sur Yvette (France); Salmon, Laurent [Thermalhydraulics and Fluid Mechanics Section, CEA Saclay 91191 Gif sur Yvette, (France)

    2015-07-01

    Nuclear heating inside a MTR reactor has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. This development required preliminary modelling and irradiation of mock-ups of the calorimetric probe in the ex-core area, where nuclear heating rate does not exceed 2 W.g{sup -1}. The calorimeter working modes, the different measurement procedures allowed with such a new probe, the main modeling and experimental results and expected advantages of this new technique have been already presented. However, these first in-core measurements were not performed beyond 6 W.g{sup -1}, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at the 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 10{sup 14} n.cm{sup -2}.s{sup -1} and nuclear heating up to 12 W.g{sup -1}. A comprehensive measurement campaign carried out from 2013 to 2015 inside all accessible irradiation locations of the core, allowed to qualify definitively this new device, not only in terms of measurement ability but also in terms of reliability. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a systematic comparison is made between results obtained by

  3. Exploring ice core drilling chips from a cold Alpine glacier for cosmogenic radionuclide (10Be) analysis

    Science.gov (United States)

    Zipf, Lars; Merchel, Silke; Bohleber, Pascal; Rugel, Georg; Scharf, Andreas

    Ice cores offer unique multi-proxy paleoclimate records, but provide only very limited sample material, which has to be carefully distributed for various proxy analyses. Beryllium-10, for example, is analysed in polar ice cores to investigate past changes of the geomagnetic field, solar activity, and the aerosol cycle, as well as to more accurately date the material. This paper explores the suitability of a drilling by-product, the so-called drilling chips, for 10Be-analysis. An ice core recently drilled at a cold Alpine glacier is used to directly compare 10Be-data from ice core samples with corresponding drilling chips. Both sample types have been spiked with 9Be-carrier and identically treated to chemically isolate beryllium. The resulting BeO has been investigated by accelerator mass spectrometry (AMS) for 10Be/9Be-ratios to calculate 10Be-concentrations in the ice. As a promising first result, four out of five sample-combinations (ice core and drilling chips) agree within 2-sigma uncertainty range. However, further studies are needed in order to fully demonstrate the potential of drilling chips for 10Be-analysis in alpine and shallow polar ice cores.

  4. Prompt Gamma Activation Analysis of the Nyírlugos obsidian core depot find

    Directory of Open Access Journals (Sweden)

    Zsolt Kasztovszky

    2014-03-01

    Full Text Available The Nyírlugos obsidian core depot find is one of the most important lithic assemblages in the collection of the Hungarian National Museum (HNM. The original set comprised 12 giant obsidian cores, of which 11 are currently on the permanent archaeological exhibition of the HNM. One of the cores is known to be inDebrecen. The first publication attributed the hoard, on the strength of giant (flint blades known from the Early and Middle Copper Age Tiszapolgár and Bodrogkeresztúr cultures, to the Copper Age. In the light of recent finds it is more likely to belong to the Middle Neolithic period. The source area was defined as Tokaj Mts., about100 kmto the NW from Nyírlugos. The size and beauty of the exceptional pieces exclude any invasive analysis. Using Prompt Gamma Activation Analysis (PGAA, we can measure major chemical components and some key trace elements of stone artefacts with adequate accuracy to successfully determine provenance of obsidian. Recent methodological development also facilitated the study of relatively large objects like the Nyírlugos cores. The cores were individually measured by PGAA. The results show that the cores originate from the Carpathian 1 sources, most probably the Viničky variety (C1b. The study of the hoard as a batch is an important contribution to the assessment of prehistoric trade and allows us to reconsider the so-called Carpathian, especially Carpathian 1 (Slovakian sources.

  5. Seismic design and analysis of nuclear power plant structures

    Institute of Scientific and Technical Information of China (English)

    Pentti Varpasuo

    2013-01-01

    The seismic design and analysis of nuclear power plant (NPP) begin with the seismic hazard assessment and design ground motion development for the site.The following steps are needed for the seismic hazard assessment and design ground motion development:a.the development of regional seismo-tectonic model with seismic source areas within 500 km radius centered to the site; b.the development of strong motion prediction equations;c.logic three development for taking into account uncertainties and seismic hazard quantification; d.the development of uniform hazard response spectra for ground motion at the site; e.simulation of acceleration time histories compatible with uniform hazard response spectra.The following phase two in seismic design of NPP structures is the analysis of structural response for the design ground motion.This second phase of the process consists of the following steps:a.development of structural models of the plant buildings; b.development of the soil model underneath the plant buildings for soil-structure interaction response analysis; c.determination of in-structure response spectra for the plant buildings for the equipment response analysis.In the third phase of the seismic design and analysis the equipment is analyzed on the basis of in-structure response spectra.For this purpose the structural models of the mechanical components and piping in the plant are set up.In large 3D-structural models used today the heaviest equipment of the primary coolant circuit is included in the structural model of the reactor building.In the fourth phase the electrical equipment and automation and control equipment are seismically qualified with the aid of the in-structure spectra developed in the phase two using large three-axial shaking tables.For this purpose the smoothed envelope spectra for calculated in-structure spectra are constructed and acceleration time is fitted to these smoothed envelope spectra.

  6. Analysis of ferroresonance in a neutral grounding system with nonlinear core loss

    Institute of Scientific and Technical Information of China (English)

    Hui Meng; Zhang Yan-Bin; Liu Chong-Xin

    2009-01-01

    The chaotic behaviour exhibited by a typical ferroresonant circuit in a neutral grounding system is investigated in this paper. In most earlier ferroresonance studies the core loss of the power transformer was neglected or represented by a linear resistance. However, this is not always true. In this paper the core loss of the power transformer is modelled by a third order series in voltage and the magnetization characteristics of the transformer are modelled by an 11th order two-term polynomial. Extensive simulations are carried out to analyse the effect of nonlinear core loss on transformer ferroresonance. A detailed analysis of simulation results demonstrates that, with the nonlinear core loss model used, the onset of chaos appears at a larger source voltage and the transient duration is shorter.

  7. Gap analysis: a method to assess core competency development in the curriculum.

    Science.gov (United States)

    Fater, Kerry H

    2013-01-01

    To determine the extent to which safety and quality improvement core competency development occurs in an undergraduate nursing program. Rapid change and increased complexity of health care environments demands that health care professionals are adequately prepared to provide high quality, safe care. A gap analysis compared the present state of competency development to a desirable (ideal) state. The core competencies, Nurse of the Future Nursing Core Competencies, reflect the ideal state and represent minimal expectations for entry into practice from pre-licensure programs. Findings from the gap analysis suggest significant strengths in numerous competency domains, deficiencies in two competency domains, and areas of redundancy in the curriculum. Gap analysis provides valuable data to direct curriculum revision. Opportunities for competency development were identified, and strategies were created jointly with the practice partner, thereby enhancing relevant knowledge, attitudes, and skills nurses need for clinical practice currently and in the future.

  8. Case Study for Effectiveness Analysis on Nuclear Regulatory Infrastructure Support for Emerging Nuclear Energy Countries

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. E.; Byeon, M. J.; Yoo, J. W.; Lee, J. M.; Lim, J. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    The donor countries need to make decisions on various steps such as whether to fully accept newcomers’ requests, the depth of support, and how the supportive action will be carried out. Such is not an easy task due to limited time, resources, manpower, etc. Thus, creating an infrastructure to support emerging nuclear energy countries is needed. This paper suggests the resource portfolio concept used in business management and aims to analyze the validity of supporting the new entrants’ development of regulatory infrastructure as a case study. This study tries to develop a very simple Excel-based tool for assessing the supporting strategy quantitatively and screening the activities that is projected to be less effective and attractive. There are many countries, so called newcomers, which have expressed interests in developing their own nuclear power program. It has been recognized by the international community that every country considering embarking upon their own nuclear power program should establish their nuclear safety infrastructure to sustain a high level of nuclear safety. The newcomers have requested for considerable assistance from the IAEA and they already have bilateral cooperation programs with the advanced countries with matured nuclear regulatory programs. Currently, the regulatory bodies that provide support are confronted with two responsibilities as follows; the primary objective of the regulatory bodies is to ensure that the operator fulfills the responsibility to protect human health.

  9. Expert judgment in analysis of human and organizational behaviour at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Reiman, L. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland). Dept. of Nuclear Safety

    1994-12-01

    Probabilistic safety assessment (PSA) of a nuclear power plant includes an assessment of the probability of each event sequence that can lead to a reactor core damage and of their consequences. Despite increasing maturity of PSA methods, there are still several problems in their use. These include the assessment of human reliability and the impact of organizational factors on plant safety. The assessment of both these issues is based on expert judgment. Therefore, the use of expert judgment in analysis of human and organizational behaviour was studied theoretically and in practical case studies in this thesis. Human errors were analysed in two case studies. In the first study cognitive actions of control room operators were analysed. For this purpose methods were developed for the qualitative and quantitative phases of the analysis. Errors of test and maintenance personnel were analysed in the second case study. Especially the dependence of errors between sequential tasks performed in redundant subsystems of a safety system was studied. A method to assess organizational behaviour was developed and applied in the third case study. The three case studies demonstrated that expert judgment can be used in the analysis of human reliability and organizational behaviour taking into account the observations made and the remarks presented in the study. However, significant uncertainties are related with expert judgment. Recommendations are presented concerning the use of different methods. Also, some insights are presented into how reliance on expert judgment could be reduced. (241 refs., 20 figs., 36 tabs.).

  10. Characterization of a Continuous Wave Laser for Resonance Ionization Mass Spectroscopy Analysis in Nuclear Forensics

    Science.gov (United States)

    2015-06-01

    OF A CONTINUOUS WAVE LASER FOR RESONANCE IONIZATION MASS SPECTROSCOPY ANALYSIS IN NUCLEAR FORENSICS by Sunny G. Lau June 2015 Thesis...IONIZATION MASS SPECTROSCOPY ANALYSIS IN NUCLEAR FORENSICS 5. FUNDING NUMBERS 6. AUTHOR(S) Sunny G. Lau 7. PERFORMING ORGANIZATION NAME(S) AND...200 words) The application of resonance ionization mass spectroscopy (RIMS) to nuclear forensics involves the use of lasers to selectively ionize

  11. The ATWS analysis of one control rod withdraw out of the HTR-10GT core in addition with bypass valve failure

    Energy Technology Data Exchange (ETDEWEB)

    Lang, Minggang, E-mail: langmg@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Room 207, Building B, Nengkelou, Beijing (China); Dong, Yujie [Institute of Nuclear and New Energy Technology, Tsinghua University, Room 503, Building A, Nengkelou, Beijing (China)

    2014-05-01

    The 10 MW high temperature gas cooled test reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700 °C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750 °C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basic accidents of HTR10-GT must be analyzed according to China's nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was greater than 800 °C, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism, in spite of the failure of bypass valve. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1283.3 °C. It was a litter than 1230 °C – the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation will be raised to 1620 °C, so the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT

  12. Influence of multigroup nuclear data uncertainties on the reactor core physics calculation%多群核数据不确定性对堆芯物理计算的影响

    Institute of Scientific and Technical Information of China (English)

    潘昕怿; 兰兵; 张春明; 靖剑平; 攸国顺

    2016-01-01

    Background: The uncertainty of nuclear data is one of the key factors resulting in the uncertainty of reactor physics calculation. Purpose: The influence of multigroup nuclear data uncertainties on the reactor core physics calculation was studied in this paper. Methods:The stochastic sampling modular SAMP based on covariance matrix of nuclear data was developed, and the hybrid method and stochastic sampling method were realized using SCALE (Standardized Computer Analyses for Licensing Evaluation) software package. The two methods were validated using 3×3 hypothetical core and then applied to the first cycle of Almaraz pressurized-water reactor (PWR) in the IAEA (International Atomic Energy Agency) fuel management benchmark. Results: Results of the two methods are in good agreement. The uncertainty of core effective multiplication factor is about 0.5%, and the maximum uncertainties of the radial and axial power are about 1.9% and 0.45% respectively in Almaraz PWR. Conclusion:The two-step method and stochastic sampling method can both be used for the uncertainty analysis of reactor core calculation.%核数据不确定性是造成反应堆物理计算结果不确定性的重要因素之一。基于所需抽样核数据的协方差矩阵开发了随机抽样模块(Stochastic Sampling, SAMP),在此基础上利用SCALE (Standardized Computer Analyses for Licensing Evaluation)软件包实现了混合法和随机抽样法两种不确定性分析方法,以研究多群核数据不确定性对堆芯物理计算的影响。以3×3假想堆芯为对象,对两种方法进行了验证,然后应用于国际原子能机构(International Atomic Energy Agency, IAEA)燃料管理基准题中的Almaraz核电厂首循环堆芯。分析结果表明,两种方法结果符合良好,Almaraz核电厂堆芯kef 不确定性约为0.5%,堆芯径向和轴向功率的最大不确定性分别为1.9%和0.45%。

  13. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository - Volume 3: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, L.L.; Wilson, J.R. (INEEL); Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K. (SNL); Rath, J.S. (New Mexico Engineering Research Institute)

    1998-10-01

    The United States Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

  14. Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

    Science.gov (United States)

    Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

    1972-01-01

    A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

  15. Geoantineutrino Spectrum, 3He/4He - ratio radial distribution and Slow Nuclear Burning on the Boundary of the Liquid and Solid Phases of the Earth's core

    CERN Document Server

    Rusov, V D; Vaschenko, V N; Tarasov, V A; Zelentsova, T N; Bolshakov, V N; Litvinov, D A; Kosenko, S I; Byegunova, O A

    2006-01-01

    The problem of the geoantineutrino deficit and the experimental results of the interaction of uranium dioxide and carbide with iron-nickel and silica-alumina melts at high pressure (5-10 Gpa) and temperature (1600-22000 C) have motivated us to consider the possible consequences of the assumption made by V.Anisichkin and coauthors that there is an actinid shell on boundary of liquid and solid phases of the Earth's core. We have shown that the activation of a natural nuclear reactor operating as the solitary waves of nuclear burning in 238U- and/or 232Th-medium (in particular, the neutron-fission progressive wave of Feoktistov and/or Teller-Ishikawa-Wood) can be such a physical consequence. The simplified model of the kinetics of accumulation and burnup in U-Pu fuel cycle of Feoktistov is developed. The results of the numerical simulation of neutron-fission wave in two-phase UO2/Fe medium on a surface of the Earth's solid core are presented. The georeactor model of 3He origin and the 3He/4He-ratio distribution ...

  16. The nuclear starburst in Arp 299-A: From the 5.0 GHz VLBI radio light-curves to its core-collapse supernova rate

    CERN Document Server

    Bondi, M; Herrero-Illana, R; Alberdi, A

    2012-01-01

    The nuclear region of the Luminous Infra-red Galaxy Arp 299-A hosts a recent ($\\simeq 10$ Myr), intense burst of massive star formation which is expected to lead to numerous core-collapse supernovae (CCSNe). Previous VLBI observations, carried out with the EVN at 5.0 GHz and with the VLBA at 2.3 and 8.4 GHz, resulted in the detection of a large number of compact, bright, non-thermal sources in a region $\\lsim$150 pc in size. We aim at establishing the nature of all non-thermal, compact components in Arp 299-A, as well as estimating its core-collapse supernova rate. We use multi-epoch European VLBI Network (EVN) observations taken at 5.0 GHz to image with milliarcsecond resolution the compact radio sources in the nuclear region of Arp 299-A. We also use one single-epoch 5.0 GHz Multi-Element Radio Linked Interferometer Network (MERLIN) observation to image the extended emission in which the compact radio sources --traced by our EVN observations-- are embedded. Twenty-six compact sources are detected, 8 of them...

  17. Nuclear Mass Visualization and Analysis at nuclearmasses.org

    Science.gov (United States)

    Smith, Michael S.; Lingerfelt, Eric J.; Nesaraja, Caroline D.; Koura, Hiroyuki; Kondev, Filip G.

    2010-08-01

    Nuclear masses play an important role in a wide variety of studies, motivating a tremendous surge in new precision mass measurements and global mass models. To handle this information, and to address inadequacies of existing text-based disseminations of nuclear masses, we have built a suite of codes-online at nuclearmasses.org-to manage, visualize, access, manipulate, share, compare, and analyze nuclear mass datasets. The services at this site are described, along with plans for future development.

  18. Economic Analysis of Different Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Won Il Ko

    2012-01-01

    Full Text Available An economic analysis has been performed to compare four nuclear fuel cycle options: a once-through cycle (OT, DUPIC recycling, thermal recycling using MOX fuel in a pressurized water reactor (PWR-MOX, and sodium fast reactor recycling employing pyroprocessing (Pyro-SFR. This comparison was made to suggest an economic competitive fuel cycle for the Republic of Korea. The fuel cycle cost (FCC has been calculated based on the equilibrium material flows integrated with the unit cost of the fuel cycle components. The levelized fuel cycle costs (LFCC have been derived in terms of mills/kWh for a fair comparison among the FCCs, and the results are as follows: OT 7.35 mills/kWh, DUPIC 9.06 mills/kWh, PUREX-MOX 8.94 mills/kWh, and Pyro-SFR 7.70 mills/kWh. Due to unavoidable uncertainties, a cost range has been applied to each unit cost, and an uncertainty study has been performed accordingly. A sensitivity analysis has also been carried out to obtain the break-even uranium price (215$/kgU for the Pyro-SFR against the OT, which demonstrates that the deployment of the Pyro-SFR may be economical in the foreseeable future. The influence of pyrotechniques on the LFCC has also been studied to determine at which level the potential advantages of Pyro-SFR can be realized.

  19. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    Science.gov (United States)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-04-01

    The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. This approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.

  20. Sensitivity analysis of FDS 6 results for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Alvear, Daniel; Puente, Eduardo; Abreu, Orlando [Cantabria Univ., Santander (Spain). Group GIDAI - Fire Safety-Research and Technology; Peco, Julian [Consejo de Seguridad Nuclear, Madrid (Spain)

    2015-12-15

    The Spanish standard ''Instruction IS-30, Rev. 1'' (February 21, 2013) allows the new approaches of risk informed performance based design (PBD) The Spanish standard ''Instruction IS-30, rev. 1'' (February 21, 2013) for demonstrating the safe shutdown capability in case of fire in nuclear power plants. In this sense, fire computer models have become an interesting tool to study real fire scenarios. Such models use a set of input parameters that define the features of the physical domain, material, radiation, turbulence, etc. This paper analyses the impact of the grid size and different sub-models of the fire simulation code FDS, version 6 with the objective to evaluate and define their relative weight in the final simulation results. For the grid size analysis, two different scale scenarios were selected, the bench scale test PENLIGHT and a large-scale test similar to Appendix B of NUREG - 1934 (17 m x 10 m x 4.6 m, with an ignition source of 2 MW and 16 cable trays). For the sub-model analysis, the PRS-INT4 real scale configuration of the INTEGRAL experimental campaign of the international OECD PRISME Project has been used. The results offer relevant data for users and show the critical parameters that must be selected properly to guarantee the quality of the simulations.

  1. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-04-01

    Abstract—The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of reprocessing streams in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type. Locally weighted PLS models were fitted on-the-fly to estimate continuous fuel characteristics. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE. This automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters and material diversions.

  2. UPVapor: Cofrentes nuclear power plant production results analysis software

    Energy Technology Data Exchange (ETDEWEB)

    Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain); Palomo, M. J. [ISIRYM, Universidad Politecnica de Valencia, Camino de Vera s/n, Valencia (Spain); Baraza, A. [Iberdrola Generacion S. A., Central Nuclear Cofrentes, Carretera Almansa Requena s/n, 04662 Cofrentes, Valencia (Spain); Vaquer, J., E-mail: m.curiel@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

    2010-10-15

    UPVapor software version 02 has been developed for the Cofrentes nuclear power plant Data Analysis Department (Spain). It is an analysis graphical environment in which users have available all the plant variables registered in the process computer system (SIEC). In order to perform this, UPVapor software has many advanced graphic tools for work simplicity, as well as a friendly environment easy to use and with many configuration possibilities. Plant variables are classified in the same way that they are in SIEC computer and these values are taken from it through the network of Iberdrola. UPVapor can generate two different types of graphics: evolution graphs and X Y graphs. The first ones analyse the evolution up to twenty plant variables in a user's defined time period and according to historic plant files. Many tools are available: cursors, graphic configuration, mobile means, non valid data visualization ... Moreover, a particular analysis configuration can be saved, as a pre selection, giving the possibility of charging pre selection directly and developing quick monitoring of a group of preselected plant variables. In X Y graphs, it is possible to analyse a variable value against another variable in a defined time. As an option, users can filter previous data depending on a variable certain range, with the possibility of programming up to five filters. As well as the other graph, X Y graph has many configurations, saving and printing options. With UPVapor software, data analysts can save a valuable time during daily work and, as it is of easy utilization, it permits to other users to perform their own analysis without ask the analysts to develop. Besides, it can be used from any work centre with access to network framework. (Author)

  3. Stereological analysis of nuclear volume in recurrent meningiomas

    DEFF Research Database (Denmark)

    Madsen, C; Schrøder, H D

    1994-01-01

    A stereological estimation of nuclear volume in recurrent and non-recurrent meningiomas was made. The aim was to investigate whether this method could discriminate between these two groups. We found that the mean nuclear volumes in recurrent meningiomas were all larger at debut than in any...... of the control tumors. The mean nuclear volume of the individual recurrent tumors appeared to change with time, showing a tendency to diminish. A relationship between large nuclear volume at presentation and number of or time interval between recurrences was not found. We conclude that measurement of mean...

  4. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  5. 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.

  6. The Dynamic Monte Carlo Method for Transient Analysis of Nuclear Reactors

    NARCIS (Netherlands)

    Sjenitzer, B.L.

    2013-01-01

    In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing co

  7. Review of Overall Safety Manual for space nuclear systems. An evaluation of a nuclear safety analysis methodology for plutonium-fueled space nuclear systems

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, J.; Inhaber, H.

    1984-02-01

    As part of its duties in connection with space missions involving nuclear power sources, the Office of Nuclear Safety (ONS) of the Office of Assistant Secretary for Environmental Protection, Safety, and Emergency Preparedness has been assigned the task of reviewing the Overall Safety Manual (OSM) (memo from B.J. Rock to J.R. Maher, December 1, 1982). The OSM, dated July 1981 and in four volumes, was prepared by NUS Corporation, Rockville, Maryland, for the US Department of Energy. The OSM provides many of the technical models and much of the data which are used by (1) space launch contractors in safety analysis reports and (2) the broader Interagency Nuclear Safety Review Panel (INSRP) safety evaluation reports. If fhs interaction between the OSM, contractors, and INSRP is to work effectively, the OSM must be accurate, comprehensive, understandable, and usable.

  8. Analysis of Random-Loading HTR-PROTEUS Cores with Continuous Energy Monte Carlo Code Based on A Statistical Geometry Model

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Isao; Miyamaru, Hiroyuki [Division of Electrical, Electronic and Information Engineering, Osaka University, Yamada-oka 2-1, Suita, Osaka, 565-0871 (Japan)

    2008-07-01

    Spherical elements have remarkable features in various applications in the nuclear engineering field. In 1990's, by the project of HTR-PROTEUS at PSI various pebble bed reactor experiments were conducted including cores with a lot of spherical fuel elements loaded randomly. In this study, criticality experiments of the random-loading HTR-PROTEUS cores were analyzed by MCNP-BALL, which could deal with a random arrangement of spherical fuel elements exactly with a statistical geometry model. As a result of analysis, the calculated effective multiplication factors were in fairly good agreement with the measurements within about 0.5%DELTAk/k. In comparison with other numerical analysis, our effective multiplication factors were between the experimental values and the VSOP calculations. To investigate the discrepancy of the effective multiplication factors between the experiments and calculations, sensitivity analyses were performed. As the result, the sensitivity of impurity boron concentration was fairly large. The reason of the present slight overestimation was not made clear at present. However, the presently existing difference was thought to be related to the impurity boron concentration, not to the modelling of the reactor and the used nuclear data. From the present study, it was confirmed that MCNP-BALL would have an advantage to conventional transport codes by comparing with their numerical results and the experimental values. As for the criticality experiment of PROTEUS, we would conclude that the two cores of Core 4.2 and 4.3 could be regarded as an equivalent experiment of a reference critical core, which was packed in the packing fraction of RLP. (authors)

  9. Determination of power distribution in the VVER-440 core on the basis of data from in-core monitors by means of a metric analysis

    Science.gov (United States)

    Kryanev, A. V.; Udumyan, D. K.; Kurchenkov, A. Yu.; Gagarinskiy, A. A.

    2014-12-01

    Problems associated with determining the power distribution in the VVER-440 core on the basis of a neutron-physics calculation and data from in-core monitors are considered. A new mathematical scheme is proposed for this on the basis of a metric analysis. In relation to the existing mathematical schemes, the scheme in question improves the accuracy and reliability of the resulting power distribution.

  10. Analysis of pan-genome to identify the core genes and essential genes of Brucella spp.

    Science.gov (United States)

    Yang, Xiaowen; Li, Yajie; Zang, Juan; Li, Yexia; Bie, Pengfei; Lu, Yanli; Wu, Qingmin

    2016-04-01

    Brucella spp. are facultative intracellular pathogens, that cause a contagious zoonotic disease, that can result in such outcomes as abortion or sterility in susceptible animal hosts and grave, debilitating illness in humans. For deciphering the survival mechanism of Brucella spp. in vivo, 42 Brucella complete genomes from NCBI were analyzed for the pan-genome and core genome by identification of their composition and function of Brucella genomes. The results showed that the total 132,143 protein-coding genes in these genomes were divided into 5369 clusters. Among these, 1710 clusters were associated with the core genome, 1182 clusters with strain-specific genes and 2477 clusters with dispensable genomes. COG analysis indicated that 44 % of the core genes were devoted to metabolism, which were mainly responsible for energy production and conversion (COG category C), and amino acid transport and metabolism (COG category E). Meanwhile, approximately 35 % of the core genes were in positive selection. In addition, 1252 potential essential genes were predicted in the core genome by comparison with a prokaryote database of essential genes. The results suggested that the core genes in Brucella genomes are relatively conservation, and the energy and amino acid metabolism play a more important role in the process of growth and reproduction in Brucella spp. This study might help us to better understand the mechanisms of Brucella persistent infection and provide some clues for further exploring the gene modules of the intracellular survival in Brucella spp.

  11. Free vibration analysis of simply supported sandwich beams with lattice truss core

    Energy Technology Data Exchange (ETDEWEB)

    Lou, Jia, E-mail: jiajia_smile@163.com [Center for Composite Materials and Structures, Harbin Institute of Technology, P.O. Box 3011, Science Park of HIT, No. 2 Yi-Kuang Street, Harbin 150080 (China); Ma, Li, E-mail: mali@hit.edu.cn [Center for Composite Materials and Structures, Harbin Institute of Technology, P.O. Box 3011, Science Park of HIT, No. 2 Yi-Kuang Street, Harbin 150080 (China); Wu, Lin-Zhi, E-mail: wlz@hit.edu.cn [Center for Composite Materials and Structures, Harbin Institute of Technology, P.O. Box 3011, Science Park of HIT, No. 2 Yi-Kuang Street, Harbin 150080 (China)

    2012-11-20

    Free vibration of AISI 304 stainless steel sandwich beams with pyramidal truss core is investigated in the present paper. The lattice truss core is transformed to a continuous homogeneous material. Considering the deformation characteristics of the sandwich beam, the following assumptions are made: (1) the thickness of the sandwich beam remains constant during deformation; (2) for the thin face sheets, only bending deformation is considered, neglecting the effect of transverse shear deformation; (3) for the core, only shear deformation is considered as the core is too weak to provide a significant contribution to the bending stiffness of the sandwich beam. The shear stress is assumed to be constant along the thickness of the core. The governing equation of free vibration is derived from Hamilton's principle, and the natural frequencies are calculated under simply supported boundary conditions. Finally, numerical simulation is carried out to get the mode shapes and natural frequencies. Our results show that the theoretical solutions agree well with the numerical results. It indicates the present method would be useful for free vibration analysis of sandwich beams with lattice truss core.

  12. Analysis of heterogeneous boron dilution transients during outages with APROS 3D nodal core model

    Energy Technology Data Exchange (ETDEWEB)

    Kuopanportti, Jaakko [Fortum Power and Heat Ltd, Nuclear Production, Fortum (Finland)

    2015-09-15

    A diluted water plug can form inside the primary coolant circuit if the coolant flow has stopped at least temporarily. The source of the clean water can be external or the fresh water can build up internally during boiling/condensing heat transfer mode, which can occur if the primary coolant inventory has decreased enough during an accident. If the flow restarts in the stagnant primary loop, the diluted water plug can enter the reactor core. During outages after the fresh fuel has been loaded and the temperature of the coolant is low, the dilution potential is the highest because the critical boron concentration is at the maximum. This paper examines the behaviour of the core as clean or diluted water plugs of different sizes enter the core during outages. The analysis were performed with the APROS 3D nodal core model of Loviisa VVER-440, which contains an own flow channel and 10 axial nodes for each fuel assembly. The widerange cross section data was calculated with CASMO-4E. According to the results, the core can withstand even large pure water plugs without fuel failures on natural circulation. The analyses emphasize the importance of the simulation of the backflows inside the core when the reactor is on natural circulation.

  13. Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

    Energy Technology Data Exchange (ETDEWEB)

    Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-05-15

    Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

  14. Comparison over the nuclear analysis of the HCLL blanket for the European DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, Giacomo; Aubert, Julien [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [ENEA, UTFUS-TECN, Via E. Fermi 4, 00044 Frascati, Rome (Italy); Fischer, Ulrich [Karlsruhe Institute of Technology, 76344 Eggenstein-Leopoldshafen, Karlsruhe (Germany)

    2016-11-01

    Highlights: • A complete nuclear analysis of the DEMO HCLL has been carried out at CEA with the TRIPOLI-4{sup ®} Monte Carlo code. • The DEMO tokamak model was generated by the CAD import tool McCad. • The HCLL blankets were implemented using a previous MCNP model developed at ENEA. • A good agreement is observed between the results obtained at CEA with TRIPOLI-4 and JEFF-3.1.1 and whose obtained at ENEA with MCNP and FENDL-2.1. - Abstract: This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4{sup ®} Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4{sup ®} representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4{sup ®} Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO.

  15. Analysis of Stainless Steel Sandwich Panels with a Metal Foam Core for Lightweight Fan Blade Design

    Science.gov (United States)

    Min, James B.; Ghosn, Louis J.; Lerch, Bradley A.; Raj, Sai V.; Holland, Frederic A., Jr.; Hebsur, Mohan G.

    2004-01-01

    The quest for cheap, low density and high performance materials in the design of aircraft and rotorcraft engine fan and propeller blades poses immense challenges to the materials and structural design engineers. The present study investigates the use of a sandwich foam fan blade mae up of solid face sheets and a metal foam core. The face sheets and the metal foam core material were an aerospace grade precipitation hardened 17-4 PH stainless steel with high strength and high toughness. The resulting structures possesses a high stiffness while being lighter than a similar solid construction. The material properties of 17-4 PH metal foam are reviewed briefly to describe the characteristics of sandwich structure for a fan blade application. A vibration analysis for natural frequencies and a detailed stress analysis on the 17-4 PH sandwich foam blade design for different combinations of kin thickness and core volume are presented with a comparison to a solid titanium blade.

  16. A method for analysis of vanillic acid in polar ice cores

    Directory of Open Access Journals (Sweden)

    M. M. Grieman

    2014-07-01

    Full Text Available Biomass burning generates a wide range of organic compounds that are transported via aerosols to the polar ice sheets. Vanillic acid is a product of conifer lignin combustion, which has previously been observed in laboratory and ambient biomass burning aerosols. In this study a method was developed for analysis of vanillic acid in melted polar ice core samples. Vanillic acid was chromatographically separated using reversed phase LC and detected using electrospray triple quadrupole mass spectrometry (ESI-MS/MS. Using a 100 μL injection loop and analysis time of 4 min, we obtained a detection limit (S : N = 2 of 58 ppt (parts per trillion by mass and an analytical precision of ±10 %. Measurements of vanillic acid in Arctic ice core samples from the Siberian Akademii Nauk core are shown as an example application of the method.

  17. Development of Uncertainty Analysis Method for SMART Digital Core Protection and Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Bon Seung; In, Wang Kee; Hwang, Dae Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The Korea Atomic Energy Research Institute has developed a system-integrated modular advanced reactor (SMART) for a seawater desalination and electricity generation. Online digital core protection and monitoring systems, called SCOPS and SCOMS respectively were developed. SCOPS calculates minimum DNBR and maximum LPD based on the several online measured system parameters. SCOMS calculates the variables of limiting conditions for operation. KAERI developed overall uncertainty analysis methodology which is used statistically combining uncertainty components of SMART core protection and monitoring system. By applying overall uncertainty factors in on-line SCOPS/SCOMS calculation, calculated LPD and DNBR are conservative with a 95/95 probability/confidence level. In this paper, uncertainty analysis method is described for SMART core protection and monitoring system

  18. Seismic analysis of spent nuclear fuel storage racks

    Energy Technology Data Exchange (ETDEWEB)

    Shah, S.J.; Biddle, J.R.; Bennett, S.M.; Schechter, C.B. [Framatome Cogema Fuels, Lynchburg, VA (United States); Harstead, G.A. [Harstead Engineering Associates, Inc., Old Tappan, NJ (United States); Marquet, F. [ATEA/FRAMATOME, Carquefou (France)

    1996-06-01

    In many nuclear power plants, existing storage racks are being replaced with high-density racks to accommodate the increasing inventory of spent fuel. In the hypothetical design considered here, the high-density arrangement of fuel assemblies, or consolidated fuel canisters, is accomplished through the use of borated stainless steel (BSS) plates acting as neutron absorbers. No structural benefit from the BSS is assumed. This paper describes the methods used to perform seismic analysis of high density spent fuel storage racks. The sensitivity of important parameters such as the effect of variation of coefficients of friction between the rack legs and the pool floor and fuel loading conditions (consolidated and unconsolidated) are also discussed in the paper. Results of this study are presented. The high-density fuel racks are simply supported by the pool floor with no structural connections to adjacent racks or to the pool walls or floor. Therefore, the racks are free standing and may slide and tip. Several time history, nonlinear, seismic analyses are required to account for variations in the coefficient of friction, rack loading configuration, and the type of the seismic event. This paper presents several of the mathematical models usually used. Friction cannot be precisely predicted, so a range of friction coefficients is assumed. The range assumed for the analysis is 0.2 to 0.8. A detailed model representing a single rack is used to evaluate the 3-D loading effects. This model is a controlling case for the stress analysis. A 2-D multi-rack model representing a row of racks between the spent fuel pool walls is used to evaluate the change in gaps between racks. The racks are normally analyzed for the fuel loading conditions of consolidated, full, empty, and half-loaded with fuel assemblies.

  19. Compression After Impact on Honeycomb Core Sandwich Panels with Thin Facesheets, Part 2: Analysis

    Science.gov (United States)

    Mcquigg, Thomas D.; Kapania, Rakesh K.; Scotti, Stephen J.; Walker, Sandra P.

    2012-01-01

    A two part research study has been completed on the topic of compression after impact (CAI) of thin facesheet honeycomb core sandwich panels. The research has focused on both experiments and analysis in an effort to establish and validate a new understanding of the damage tolerance of these materials. Part 2, the subject of the current paper, is focused on the analysis, which corresponds to the CAI testings described in Part 1. Of interest, are sandwich panels, with aerospace applications, which consist of very thin, woven S2-fiberglass (with MTM45-1 epoxy) facesheets adhered to a Nomex honeycomb core. Two sets of materials, which were identical with the exception of the density of the honeycomb core, were tested in Part 1. The results highlighted the need for analysis methods which taken into account multiple failure modes. A finite element model (FEM) is developed here, in Part 2. A commercial implementation of the Multicontinuum Failure Theory (MCT) for progressive failure analysis (PFA) in composite laminates, Helius:MCT, is included in this model. The inclusion of PFA in the present model provided a new, unique ability to account for multiple failure modes. In addition, significant impact damage detail is included in the model. A sensitivity study, used to assess the effect of each damage parameter on overall analysis results, is included in an appendix. Analysis results are compared to the experimental results for each of the 32 CAI sandwich panel specimens tested to failure. The failure of each specimen is predicted using the high-fidelity, physicsbased analysis model developed here, and the results highlight key improvements in the understanding of honeycomb core sandwich panel CAI failure. Finally, a parametric study highlights the strength benefits compared to mass penalty for various core densities.

  20. Nuclear Propulsion and Power Non-Nuclear Test Facility (NP2NTF): Preliminary Analysis and Feasibility Assessment Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors, which power nuclear propulsion and power systems, and the nuclear radiation and residual radioactivity associated with these systems, impose...

  1. Analysis constants for database of neutron nuclear data

    Science.gov (United States)

    Bedenko, S. V.; Jeremiah, J. Joseph; Knyshev, V. V.; Shamanin, I. V.

    2016-07-01

    At present there is a variety of experimental and calculation nuclear data which are rather entirely presented in the following evaluated nuclear data libraries: ENDF (USA), JEFF (Europe), JENDL (Japan), TENDL (Russian Federation), ROSFOND (Russian Federation). Libraries of nuclear data, used for neutron-physics calculations in programs: Scale (Origen-Arp), MCNP, WIMS, MCU, and others. Nevertheless all existing nuclear data bases, including evaluated ones, contain practically no information about threshold neutron reactions on 232Th nuclei; available values of outputs and cross-sections significantly differ by orders. The work shows necessity of nuclear constants corrections which are used in the calculations of grids and thorium storage systems. The results of numerical experiments lattices and storage systems with thorium.

  2. Monte-Carlo Application for Nondestructive Nuclear Waste Analysis

    Science.gov (United States)

    Carasco, C.; Engels, R.; Frank, M.; Furletov, S.; Furletova, J.; Genreith, C.; Havenith, A.; Kemmerling, G.; Kettler, J.; Krings, T.; Ma, J.-L.; Mauerhofer, E.; Neike, D.; Payan, E.; Perot, B.; Rossbach, M.; Schitthelm, O.; Schumann, M.; Vasquez, R.

    2014-06-01

    Radioactive waste has to undergo a process of quality checking in order to check its conformance with national regulations prior to its transport, intermediate storage and final disposal. Within the quality checking of radioactive waste packages non-destructive assays are required to characterize their radio-toxic and chemo-toxic contents. The Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety of the Forschungszentrum Jülich develops in the framework of cooperation nondestructive analytical techniques for the routine characterization of radioactive waste packages at industrial-scale. During the phase of research and development Monte Carlo techniques are used to simulate the transport of particle, especially photons, electrons and neutrons, through matter and to obtain the response of detection systems. The radiological characterization of low and intermediate level radioactive waste drums is performed by segmented γ-scanning (SGS). To precisely and accurately reconstruct the isotope specific activity content in waste drums by SGS measurement, an innovative method called SGSreco was developed. The Geant4 code was used to simulate the response of the collimated detection system for waste drums with different activity and matrix configurations. These simulations allow a far more detailed optimization, validation and benchmark of SGSreco, since the construction of test drums covering a broad range of activity and matrix properties is time consuming and cost intensive. The MEDINA (Multi Element Detection based on Instrumental Neutron Activation) test facility was developed to identify and quantify non-radioactive elements and substances in radioactive waste drums. MEDINA is based on prompt and delayed gamma neutron activation analysis (P&DGNAA) using a 14 MeV neutron generator. MCNP simulations were carried out to study the response of the MEDINA facility in terms of gamma spectra, time dependence of the neutron energy spectrum

  3. Analysis of Early Severe Accident Initiated by LBLOCA for Qinshan Phase II Nuclear Power Project

    Directory of Open Access Journals (Sweden)

    Shi Xing-Wei

    2013-07-01

    Full Text Available The purpose of this study is to simulate an early Severe Accident (SA scenario more detail through transferring the thermal-hydraulic status of the plant predicted by RELAP5 computer code to SA Program (SAP. Based on the criterion of date extract time, the RELAP5 thermal-hydraulic calculation data is extracted to form a file for SAP input card at 1477K of cladding surface. Relying on the thermal-hydraulic boundary parameters calculated by RELAP5 code, analysis of early SA initiated by the Large Break Loss-of-Coolant Accident (LBLOCA without mitigation measures for Qinshan Phase II Nuclear Power Plant (QSP-II performed by SAP through finding the key events of accident sequence, estimating the amount of hydrogen generation and oxidation behavior of the cladding and evaluating the relocation order of the materials collapsed in the central region of the core. The results of this study are expected to improve the SA analysis methodology more detail through analyzing early SA scenario.

  4. Computational Design and Analysis of Core Material of Single-Phase Capacitor Run Induction Motor

    Directory of Open Access Journals (Sweden)

    Gurmeet Singh

    2014-07-01

    Full Text Available A Single-phase induction motor (SPIM has very crucial role in industrial, domestic and commercial sectors. So, the efficient SPIM is a foremost requirement of today's market. For efficient motors, many research methodologies and propositions have been given by researchers in past. Various parameters like as stator/rotor slot variation, size and shape of stator/rotor slots, stator/rotor winding configuration, choice of core material etc. have momentous impact on machine design. Core material influences the motor performance to a degree. Magnetic flux linkage and leakage preliminary depends upon the magnetic properties of core material and air gap. The analysis of effects of core material on the magnetic flux distribution and the performance of induction motor is of immense importance to meet out the desirable performance. An increase in the air gap length will result in the air gap performance characteristics deterioration and decrease in air gap length will lead to serious mechanical balancing concern. So possibility of much variation in air gap beyond the limits on both sides is not feasible. For the optimized performance of the induction motor the core material plays a significant role. Using higher magnetic flux density, reduction on a magnetizing reactance and leakage of flux can be achieved. In this thesis work the analysis of single phase induction motor has been carried out with different core materials. The four models have been simulated using Ansys Maxwell 15.0. Higher flux density selection for same machine dimensions result into huge amount of reduction in iron core losses and thereby improve the efficiency. In this paper 2% higher efficiency has been achieved with Steel_1010 as compared to the machine using conventional D23 material. Out of four models result reflected by the machine using steel_1010 and steel_1008 are found to be better.

  5. Analysis of the 1957-58 Soviet nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Trabalka, J.R.; Eyman, L.D.; Auerbach, S.I.

    1979-12-01

    The occurrence of a Soviet accident in the winter of 1957-58, involving the atmospheric release of reprocessed fission wastes (cooling time approximately 1-2 yrs.), appears to have been confirmed, primarily by an analysis of the USSR radioecology literature. Due to the high population density in the affected region (Cheliabinsk Province in the highly industrialized Urals Region) and the reported level of /sup 90/Sr contamination , the event probably resulted in the evacuation and/or resettlement of the human population from a significant area (100-1000 km/sup 2/). The resulting contamination zone is estimated to have contained approximately 10/sup 6/ Ci of /sup 90/Sr (reference radionuclide); a relatively small fraction of the total may have been dispersed as an aerosol. Although a plausible explanation for the incident exists (i.e., use of now-obsolete waste storage-/sup 137/Cs isotope separation techniques), it is not yet possible, based on the limited information presently available, to completely dismiss this phenomenon as a purely historical event. It seems imperative that we have a complete explanation of the causes and consequences of this incident. Soviet experience gained in application of corrective measures would be invaluable to the rest of the world nuclear community.

  6. Nuclear Safety Analysis for the Mars Exploration Rover 2003 Project

    Science.gov (United States)

    Firstenberg, Henry; Rutger, Lyle L.; Mukunda, Meera; Bartram, Bart W.

    2004-02-01

    The National Aeronautics and Space Administration's Mars Exploration Rover (MER) 2003 project is designed to place two mobile laboratories (Rovers) on Mars to remotely characterize a diversity of rocks and soils. Milestones accomplished so far include two successful launches of identical spacecraft (the MER-A and MER-B missions) from Cape Canaveral Air Force Station, Florida on June 10 and July 7, 2003. Each Rover uses eight Light Weight Radioisotope Heater Units (LWRHUs) fueled with plutonium-238 dioxide to provide local heating of Rover components. The LWRHUs are provided by the U.S. Department of Energy. In addition, small quantities of radioactive materials in sealed sources are used in scientific instrumentation on the Rover. Due to the radioactive nature of these materials and the potential for accidents, a formal Launch Approval Process requires the preparation of a Final Safety Analysis Report (FSAR) for submittal to and independent review by an Interagency Nuclear Safety Review Panel. This paper presents a summary of the FSAR in terms of potential accident scenarios, probabilities, source terms, radiological consequences, mission risks, and uncertainties in the reported results.

  7. Application of CFD Codes in Nuclear Reactor Safety Analysis

    Directory of Open Access Journals (Sweden)

    T. Höhne

    2010-01-01

    Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

  8. Cost benefit analysis for remediation of a nuclear industry landfill

    Energy Technology Data Exchange (ETDEWEB)

    Parker, Tom; Hardisty, Paul [WorleyParsons Komex, Bristol (United Kingdom); Dennis, Frank; Liddiard, Mark; McClelland, Paul [UKAEA, Dounreay (United Kingdom)

    2006-09-15

    An old landfill site, licensed to receive inert construction waste, is situated on the top of hard rock cliffs adjacent to the sea at the Dounreay nuclear facility in Scotland. During restoration and investigation work at the landfill, radioactively contaminated material and asbestos was identified. UKAEA subsequently investigated the feasibility of remediating the landfill with the aim of removing any remaining radioactive or otherwise-contaminated material. The cost of landfill remediation would be considerable, making Cost Benefit Analysis (CBA) an ideal tool for assessing remediation options. The overall conclusion of the CBA, from a remedial decision making point of view, is that the remediation objective for the landfill should be to reduce any impacts to the current receptors through a comprehensive pathway control scheme. This would be considerably less expensive than even a limited source removal approach. Aggressive source removal objectives are not likely to be economic, even under the most conservative assumptions. A natural monitored attenuation approach will not be economic. All remediation options are considered assuming compliance with the existing regulatory requirements to monitor and cap the landfill before and after closure.

  9. A study on the core analysis methodology for SMART CEA ejection accident-I

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Lee, Chung Chan; Kim, Kyo Yoon; Cho, Byung Oh

    1999-04-01

    A methodology to analyze the fuel enthalpy is developed based on MASTER that is a time dependent 3 dimensional core analysis code. Using the proposed methodology, SMART CEA ejection accident is analyzed. Moreover, radiation doses are estimated at the exclusion area boundary and low population zone to confirm the criteria for the accident. (Author). 31 refs., 13 tabs., 18 figs.

  10. Direct chemical analysis of frozen ice cores by UV-laser ablation ICPMS

    DEFF Research Database (Denmark)

    Müller, Wolfgang; Shelley, J. Michael G.; Rasmussen, Sune Olander

    2011-01-01

    Cryo-cell UV-LA-ICPMS is a new technique for direct chemical analysis of frozen ice cores at high spatial resolution (micrometer). It was tested in a pilot study on NGRIP ice and reveals sea ice/dust records and annual layer signatures at unprecedented spatial/time resolution. Uniquely...

  11. Survey Analysis on Nuclear Security Culture Recognition of Nuclear Facility in 2014

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Yunjeong; Lee, Jeongho; Kim, Jaekwang [Korea Institute of Nonproliferation and Control International Nuclear Security Academy, Daejeon (Korea, Republic of)

    2015-05-15

    All organizations involved in implementing physical protection should give due priority to the security culture, to its development and maintenance necessary to ensure its effective implementation in the entire organization. In this context, Korea Institute of Non-proliferation and Control(KINAC) confirms recognition about protection of people who work in nuclear field and developed questionnaire for utilizing fundamental data for nuclear security culture enhancement activity and conducted a survey. As a result, systematic education needs to employees. Choosing differentiated topic is required to consider employees because recognition level of age, position and division is different. And a variety of education technology as obligatory education such as filling the course time or the one-off thing has limitation. And taking complementary measures needs since there were many opinions that employees feel difficult to understand papers such as regulation and guidelines and so on related security. Finally, we hope to make fundament available to evaluate nuclear security culture recognition level based on the existing questionnaire would be changed to realistic and enhancement in recognition survey for future nuclear security culture.

  12. A Study for Appropriateness of National Nuclear Policy by using Economic Analysis Methodology after Fukushima accident

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Jong Myoung; Roh, Myung Sub [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    The aim of this paper is to clarify the appropriateness of national nuclear policy in BPE of Korea from an economic perspective. To do this, this paper only focus on the economic analysis methodology without any considering other conditions such as political, cultural, or historical things. In a number of countries, especially Korea, nuclear energy policy is keeping the status quo after Fukushima accident. However the nation's nuclear policy may vary depending on the choice of people. Thus, to make the right decisions, it is important to deliver accurate information and knowledge about nuclear energy to the people. As proven in this paper, the levelized cost of nuclear power is the most inexpensive among the base load units. As the reliance on nuclear power is getting stronger through the economic logic, the nuclear safety and environmental elements will be strengthened. Based on this, national nuclear policy should be promoted. In the aftermath of the Fukushima accident recognized as the world's worst nuclear disaster since the Chernobyl, there are some changes in the nuclear energy policy of various countries. Germany, for example, called a halt to operate Nuclear Power Plant (NPP) which accounts for about 7.5% of the national power generation capacity of 6.3GW. In developing countries such as China and India they conducted the safety check of the nuclear power plants again before preceding their nuclear business. Korea government announced 'The 6th Basic Plan for Long-term Electricity Supply and Demand (BPE)', considering the safety and general public acceptance of the nuclear power plants. According to BPE, they postponed a plan for additional NPP construction, except for constructions that had been already reflected in the 5th BPE. All told, the responses for nuclear energy policy of countries are different depending on their own circumstances.

  13. Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core

    Science.gov (United States)

    Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.

    2006-12-01

    Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62149Sm and its dependence on the shift of a resonance position Er (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73⩽ΔEr⩽62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant α. We obtain new, more accurate limits of -4×10-17⩽α·/α⩽3×10-17yr-1. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

  14. Numerical analysis on reduction of radioactive actinides by recycling of nuclear fuel; Analisis numerico sobre reduccion de actinidos radiactivos por reciclado de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Balboa L, H. E.

    2014-07-01

    Worldwide, human growth has reached unparalleled levels historically, this implies a need for more energy, and just in 2007 was consumed in the USA 4157 x 10{sup 9} kWh of electricity and there were 6 x 10{sup 9} metric tons of carbon dioxide, which causes a devastating effect on our environment. To this problem, a solution to the demand for non-fossil energy is nuclear energy, which is one of the least polluting and the cheapest among non-fossil energy; however, a problem remains unresolved the waste generation of nuclear fuels. In this work the option of a possible transmutation of actinides in a nuclear reactor of BWR was analyzed, an example of this are the nuclear reactors at the Laguna Verde nuclear power plant, which have generated spent fuel stored in pools awaiting a decision for final disposal or any other existing alternative. Assuming that the spent fuel was reprocessed to separate useful materials and actinides such as plutonium and uranium remaining, could take these actinides and to recycle them inside the same reactor that produced them, so il will be reduced the radiotoxicity of spent fuel. The main idea of this paper is to evaluate by means of numeric simulation (using the Core Management System (CMS)) the reduction of minor actinides in the case of being recycled in fresh fuel of the type BWR. The actinides were introduced hypothetically in the fuel pellets to 6% by weight, and then use a burned in the range of 0-65 G Wd/Tm, in order to have a better panorama of their behavior and thus know which it is the best choice for maximum reduction of actinides. Several cases were studied, that is to say were used as fuels; the UO{sub 2} and MOX. Six different cases were also studied to see the behavior of actinides in different situations. The CMS platform calculation was used for the analysis of the cases presented. Favorable results were obtained, having decreased from a range of 35% to 65% of minor actinides initially introduced in the fuel rods

  15. Lack of nuclear clusters in dwarf sferoidal galaxies: implications for massive black holes formation and the cusp/core problem

    CERN Document Server

    Arca-Sedda, Manuel

    2016-01-01

    One of the leading scenarios for the formation of nuclear star clusters in galaxies is related to the orbital decay of globular clusters (GCs) and their subsequent merging, though alternative theories are currently debated. The availability of high-quality data for GCs structural and orbital parameters allow to test different nuclear star cluster formation scenarios. The Fornax dwarf spheroidal (dSph) galaxy is the heaviest satellite of the Milky Way and it is the only known dwarf spheroidal hosting 5 GCs, whereas there are no clear signatures for the presence of a central massive black hole. For this reason, it represents a suited place to study the orbital decay process in dwarf galaxies. In this paper we model the future evolution of the Fornax GCs simulating them and the host galaxy by means of direct $N$-body simulations. Our simulations take in account also the gravitational field generated by the Milky Way. We found that if the Fornax galaxy is embedded in a standard Cold Dark Matter Halo, the nuclear ...

  16. Natural Nuclear Reactor Oklo and Variation of Fundamental Constants Part 1: Computation of Neutronic of Fresh Core

    CERN Document Server

    Petrov, Yu V; Onegin, M S; Petrov, V Yu; Sakhnovskii, E G; Petrov, Yu.V.

    2006-01-01

    Using a modern methods of reactor physics we have performed the full-scale calculations of the natural reactor Oklo. For reliability we have used the recent version of two Monte Carlo codes: the Russian code MCU REA and world wide known code MCNP (USA). Both codes produce close results. We constructed computer model of zone RZ2 of reactor Oklo which takes into account all details of design and composition. The calculations were performed for the three fresh cores with different uranium contents. Multiplication factors, reactivities and neutron fluxes were calculated. We estimated also the temperature and void effects for the fresh core. As would be expected, we have found for the fresh core a great difference between reactor spectra and Maxwell's one, which was used before for averaging cross sections in the Oklo reactor. The averaged cross section of Sm and its dependence on the shift of resonance position (due to variation of fundamental constants) are significantly different from previous results. Contrary...

  17. Oil prices, nuclear energy consumption, and economic growth: New evidence using a heterogeneous panel analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chien-Chiang, E-mail: cclee@cm.nsysu.edu.t [Department of Finance, National Sun Yat-sen University Kaohsiung, Taiwan (China); Chiu, Yi-Bin [Department of Finance, National Sun Yat-sen University Kaohsiung, Taiwan (China)

    2011-04-15

    This paper applies panel data analysis to examine the short-run dynamics and long-run equilibrium relationships among nuclear energy consumption, oil prices, oil consumption, and economic growth for developed countries covering the period 1971-2006. The panel cointegration results show that in the long run, oil prices have a positive impact on nuclear energy consumption, suggesting the existence of the substitution relationship between nuclear energy and oil. The long-run elasticity of nuclear energy with respect to real income is approximately 0.89, and real income has a greater impact on nuclear energy than do oil prices in the long run. Furthermore, the panel causality results find evidence of unidirectional causality running from oil prices and economic growth to nuclear energy consumption in the long run, while there is no causality between nuclear energy consumption and economic growth in the short run. - Research highlights: {yields} We examine the relationship among nuclear energy consumption, oil prices, oil consumption, and economic growth for developed countries. {yields} The existence of the substitution relationship between nuclear energy and oil. {yields} Real income has a greater impact on nuclear energy than do oil prices in the long run. {yields} An unidirectional causality running from oil prices and economic growth to nuclear energy consumption in the long run.

  18. Dynamic Systems Analysis Report for Nuclear Fuel Recycle

    Energy Technology Data Exchange (ETDEWEB)

    Brent Dixon; Sonny Kim; David Shropshire; Steven Piet; Gretchen Matthern; Bill Halsey

    2008-12-01

    This report examines the time-dependent dynamics of transitioning from the current United States (U.S.) nuclear fuel cycle where used nuclear fuel is disposed in a repository to a closed fuel cycle where the used fuel is recycled and only fission products and waste are disposed. The report is intended to help inform policy developers, decision makers, and program managers of system-level options and constraints as they guide the formulation and implementation of advanced fuel cycle development and demonstration efforts and move toward deployment of nuclear fuel recycling infrastructure.

  19. Prospective analysis. Nuclear deterrence in 2030; Essai de prospective. La dissuasion nucleaire en 2030

    Energy Technology Data Exchange (ETDEWEB)

    Tertrais, B

    2006-12-15

    This study is a prospective analysis of the long-term future of nuclear weapons, and particularly the future of French nuclear deterrence after 2015. The selected time period is 2025-2030. The principal objective is to reflect on what the nuclear world might look like during the first part of the 21 st century, beyond the modernization decisions already planned or envisaged, and to draw conclusions for the future of the French deterrent. (author)

  20. Optimization of High-Resolution Continuous Flow Analysis for Transient Climate Signals in Ice Cores

    DEFF Research Database (Denmark)

    Bigler, Matthias; Svensson, Anders; Kettner, Ernesto

    2011-01-01

    Over the past two decades, continuous flow analysis (CFA) systems have been refined and widely used to measure aerosol constituents in polar and alpine ice cores in very high-depth resolution. Here we present a newly designed system consisting of sodium, ammonium, dust particles, and electrolytic...... meltwater conductivity detection modules. The system is optimized for high- resolution determination of transient signals in thin layers of deep polar ice cores. Based on standard measurements and by comparing sections of early Holocene and glacial ice from Greenland, we find that the new system features...

  1. A High-Resolution Continuous Flow Analysis System for Polar Ice Cores

    DEFF Research Database (Denmark)

    Dallmayr, Remi; Goto-Azuma, Kumiko; Kjær, Helle Astrid

    2016-01-01

    of Polar Research (NIPR) in Tokyo. The system allows the continuous analysis of stable water isotopes and electrical conductivity, as well as the collection of discrete samples from both inner and outer parts of the core. This CFA system was designed to have sufficiently high temporal resolution to detect...... signals of abrupt climate change in deep polar ice cores. To test its performance, we used the system to analyze different climate intervals in ice drilled at the NEEM (North Greenland Eemian Ice Drilling) site, Greenland. The quality of our continuous measurement of stable water isotopes has been...

  2. Evaluation of CANDU6 PCR (power coefficient of reactivity) with a 3-D whole-core Monte Carlo Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Motalab, Mohammad Abdul; Kim, Woosong; Kim, Yonghee, E-mail: yongheekim@kaist.ac.kr

    2015-12-15

    Highlights: • The PCR of the CANDU6 reactor is slightly negative at low power, e.g. <80% P. • Doppler broadening of scattering resonances improves noticeably the FTC and make the PCR more negative or less positive in CANDU6. • The elevated inlet coolant condition can worsen significantly the PCR of CANDU6. • Improved design tools are needed for the safety evaluation of CANDU6 reactor. - Abstract: The power coefficient of reactivity (PCR) is a very important parameter for inherent safety and stability of nuclear reactors. The combined effect of a relatively less negative fuel temperature coefficient and a positive coolant temperature coefficient make the CANDU6 (CANada Deuterium Uranium) PCR very close to zero. In the original CANDU6 design, the PCR was calculated to be clearly negative. However, the latest physics design tools predict that the PCR is slightly positive for a wide operational range of reactor power. It is upon this contradictory observation that the CANDU6 PCR is re-evaluated in this work. In our previous study, the CANDU6 PCR was evaluated through a standard lattice analysis at mid-burnup and was found to be negative at low power. In this paper, the study was extended to a detailed 3-D CANDU6 whole-core model using the Monte Carlo code Serpent2. The Doppler broadening rejection correction (DBRC) method was implemented in the Serpent2 code in order to take into account thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. Time-average equilibrium core was considered for the evaluation of the representative PCR of CANDU6. Two thermal hydraulic models were considered in this work: one at design condition and the other at operating condition. Bundle-wise distributions of the coolant properties are modeled and the bundle-wise fuel temperature is also considered in this study. The evaluated nuclear data library ENDF/B-VII.0 was used throughout this Serpent2 evaluation. In these Monte Carlo calculations, a large number

  3. Nuclear reaction analysis (NRA) for trace element detection

    Energy Technology Data Exchange (ETDEWEB)

    Doebeli, M. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Noll, K. [Bern Univ. (Switzerland)

    1997-09-01

    Ion beam induced nuclear reactions can be used to analyse trace element concentrations in materials. The method is especially suited for the detection of light contaminants in heavy matrices. (author) 3 figs., 2 refs.

  4. Stability Analysis of the EBR-I Mark-II Core Meltdown Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae-Yong; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of this paper is to analyze the stability of the EBR-I core meltdown accident using the NuSTAB code. The result of NuSTAB analysis is compared with previous stability analysis by Sandmeier using the root locus method. The Experimental Breeder Reactor I (EBR-1) at Argonne National Laboratory was designed to demonstrate fast reactor breeding and to prove the use of liquid-metal coolant for power production and reached criticality in August 1951. The EBR-I reactor was undergoing a series of physics experiments and the Mark-II core was melted accidentally on Nov. 29, 1955. The experiment was going to increase core temperature to 500C to see if the reactor loses reactivity, and scram when the power reached 1500 kW or doubling of fission rate per second. However the operator scrammed with a slow moving control and missed the shutdown by two seconds and caused the core meltdown. The NuSTAB code has an advantage of analyzing space-dependent fast reactors and predicting regional oscillations compared to the point kinetics. Also, NuSTAB can be useful when the coupled neutronic-thermal-hydraulic codes cannot be used for stability analysis. Future work includes analyses of the PGSFR for various operating conditions as well as further validation of the NuSTAB calculations against SFR stability experiments when such experiments become available.

  5. Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2015-01-01

    Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

  6. High Level Analysis, Design and Validation of Distributed Mobile Systems with CoreASM

    Science.gov (United States)

    Farahbod, R.; Glässer, U.; Jackson, P. J.; Vajihollahi, M.

    System design is a creative activity calling for abstract models that facilitate reasoning about the key system attributes (desired requirements and resulting properties) so as to ensure these attributes are properly established prior to actually building a system. We explore here the practical side of using the abstract state machine (ASM) formalism in combination with the CoreASM open source tool environment for high-level design and experimental validation of complex distributed systems. Emphasizing the early phases of the design process, a guiding principle is to support freedom of experimentation by minimizing the need for encoding. CoreASM has been developed and tested building on a broad scope of applications, spanning computational criminology, maritime surveillance and situation analysis. We critically reexamine here the CoreASM project in light of three different application scenarios.

  7. An improved continuous flow analysis system for high-resolution field measurements on ice cores.

    Science.gov (United States)

    Kaufmann, Patrik R; Federer, Urs; Hutterli, Manuel A; Bigler, Matthias; Schüpbach, Simon; Ruth, Urs; Schmitt, Jochen; Stocker, Thomas F

    2008-11-01

    Continuous flow analysis (CFA) is a well-established method to obtain information about impurity contents in ice cores as indicators of past changes in the climate system. A section of an ice core is continuously melted on a melter head supplying a sample water flow which is analyzed online. This provides high depth and time resolution of the ice core records and very efficient sample decontamination as only the inner part of the ice sample is analyzed. Here we present an improved CFA system which has been totally redesigned in view of a significantly enhanced overall efficiency and flexibility, signal quality, compactness, and ease of use. These are critical requirements especially for operations of CFA during field campaigns, e.g., in Antarctica or Greenland. Furthermore, a novel deviceto measure the total air content in the ice was developed. Subsequently, the air bubbles are now extracted continuously from the sample water flow for subsequent gas measurements.

  8. Analysis and experiment of eddy current loss in Homopolar magnetic bearings with laminated rotor cores

    Science.gov (United States)

    Jinji, Sun; Dong, Chen

    2013-08-01

    This paper analyses the eddy current loss in Homopolar magnetic bearings with laminated rotor cores produced by the high speed rotation in order to reduce the power loss for the aerospace applications. The analytical model of rotational power loss is proposed in Homopolar magnetic bearings with laminated rotor cores considering the magnetic circuit difference between Homopolar and Heteropolar magnetic bearings. Therefore, the eddy current power loss can be calculated accurately using the analytical model by magnetic field solutions according to the distribution of magnetic fields around the pole surface and boundary conditions at the surface of the rotor cores. The measurement method of rotational power loss in Homopolar magnetic bearing is proposed, and the results of the theoretical analysis are verified by experiments in the prototype MSCMG. The experimental results show the correctness of calculation results.

  9. Optimization of high-resolution continuous flow analysis for transient climate signals in ice cores.

    Science.gov (United States)

    Bigler, Matthias; Svensson, Anders; Kettner, Ernesto; Vallelonga, Paul; Nielsen, Maibritt E; Steffensen, Jørgen Peder

    2011-05-15

    Over the past two decades, continuous flow analysis (CFA) systems have been refined and widely used to measure aerosol constituents in polar and alpine ice cores in very high-depth resolution. Here we present a newly designed system consisting of sodium, ammonium, dust particles, and electrolytic meltwater conductivity detection modules. The system is optimized for high-resolution determination of transient signals in thin layers of deep polar ice cores. Based on standard measurements and by comparing sections of early Holocene and glacial ice from Greenland, we find that the new system features a depth resolution in the ice of a few millimeters which is considerably better than other CFA systems. Thus, the new system can resolve ice strata down to 10 mm thickness and has the potential of identifying annual layers in both Greenland and Antarctic ice cores throughout the last glacial cycle.

  10. Microbial Analysis of Australian Dry Lake Cores; Analogs For Biogeochemical Processes

    Science.gov (United States)

    Nguyen, A. V.; Baldridge, A. M.; Thomson, B. J.

    2014-12-01

    Lake Gilmore in Western Australia is an acidic ephemeral lake that is analogous to Martian geochemical processes represented by interbedded phyllosilicates and sulfates. These areas demonstrate remnants of a global-scale change on Mars during the late Noachian era from a neutral to alkaline pH to relatively lower pH in the Hesperian era that continues to persist today. The geochemistry of these areas could possibly be caused by small-scale changes such as microbial metabolism. Two approaches were used to determine the presence of microbes in the Australian dry lake cores: DNA analysis and lipid analysis. Detecting DNA or lipids in the cores will provide evidence of living or deceased organisms since they provide distinct markers for life. Basic DNA analysis consists of extraction, amplification through PCR, plasmid cloning, and DNA sequencing. Once the sequence of unknown DNA is known, an online program, BLAST, will be used to identify the microbes for further analysis. The lipid analysis approach consists of phospholipid fatty acid analysis that is done by Microbial ID, which will provide direct identification any microbes from the presence of lipids. Identified microbes are then compared to mineralogy results from the x-ray diffraction of the core samples to determine if the types of metabolic reactions are consistent with the variation in composition in these analog deposits. If so, it provides intriguing implications for the presence of life in similar Martian deposits.

  11. Analysis of keff Uncertainty on Nuclear Data of CEFR

    Institute of Scientific and Technical Information of China (English)

    YANG; Jun; YU; Hong; XU; Li; HU; Yun

    2013-01-01

    The uncertainty to nuclear data is an important part of the total uncertainty of keff calculation value.In order to analyze the nuclear data uncertainty of keff for CEFR,a general expression of sensitivity to an integral quantity was derived based on the generalize perturbation theory(GPT).Using this general expression,the specific calculation formula of the sensitivity of keff was given,and the multi-group

  12. Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; David W. Nigg

    2009-11-01

    One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

  13. Risk analysis and solving the nuclear waste siting problem

    Energy Technology Data Exchange (ETDEWEB)

    Inhaber, H.

    1993-12-01

    In spite of millions of dollars and countless human resources being expended on finding nuclear wastes sites, the search has proved extremely difficult for the nuclear industry. This may be due to the approach followed, rather than inadequacies in research or funding. A new approach to the problem, the reverse Dutch auction, is suggested. It retains some of the useful elements of the present system, but it also adds new ones.

  14. Dynamical analysis of innovative core designs facing unprotected transients with the MAT5 DYN code

    Energy Technology Data Exchange (ETDEWEB)

    Darmet, G.; Massara, S. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01

    Since 2007, advanced Sodium-cooled Fast Reactors (SFR) are investigated by CEA, AREVA and EDF in the framework of a joint French collaboration. A prototype called ASTRID, sets out to demonstrate progress made in SFR technology, is due to operate in the years 2020's. The modeling of unprotected transients by computer codes is one of the key safety issues in the design approach to such SFR systems. For that purpose, the activity on CATHARE, which is the reference code for the transient analysis of ASTRID, has been strengthened during last years by CEA. In the meantime, EDF has developed a simplified and multi-channel code, named MAT5 DYN, to analyze and validate innovative core designs facing protected and unprotected transients. First, the paper consists in a description of MAT5 DYN: a code based on the existing code MAT4 DYN including major improvements on geometry description and physical modeling. Second, two core designs based on the CFV core design developed at CEA are presented. Then, the dynamic response of those heterogeneous cores is analyzed during unprotected loss of flow (ULOF) transient and unprotected transient of power (UTOP). The results highlight the importance of the low void core effect specific to the CFV design. Such an effect, when combined with a sufficient primary pump halving time and an optimized cooling group scheme, allows to delay (or, possibly, avoid) the sodium boiling onset during ULOF accidents. (authors)

  15. 3D Finite element analysis of functionally graded multilayered dental ceramic cores.

    Science.gov (United States)

    Al-Maqtari, Ali Abdullah; Razak, Abdul Aziz Abdul; Hamdi, Mohd

    2014-01-01

    This study aimed at investigating and establishing stress distributions in graded multilayered zirconia/alumina ceramic cores and at veneer-core-cement-dentin interfaces, using finite element analysis (FEA), to facilitate the structural design of ceramic cores through computer modeling. An intact maxillary premolar was digitized using CT scanning. An imaging software, Mimics, was used t