ZZ THERMOS, Multigroup P0 to P5 Thermal Scattering Kernels from ENDF/B Scattering Law Data
International Nuclear Information System (INIS)
McCrosson, F.J.; Finch, D.R.
1975-01-01
1 - Description of problem or function: Number of groups: 30-group THERMOS thermal scattering kernels. Nuclides: Molecular H 2 O, Molecular D 2 O, Graphite, Polyethylene, Benzene, Zr bound in ZrHx, H bound in ZrHx, Beryllium-9, Beryllium Oxide, Uranium Dioxide. Origin: ENDF/B library. Weighting Spectrum: yes. These data are 30-group THERMOS thermal scattering kernels for P0 to P5 Legendre orders for every temperature of every material from s(alpha,beta) data stored in the ENDF/B library. These scattering kernels were generated using the FLANGE2 computer code (NESC Abstract 368). To test the kernels, the integral properties of each set of kernels were determined by a precision integration of the diffusion length equation and compared to experimental measurements of these properties. In general, the agreement was very good. Details of the methods used and results obtained are contained in the reference. The scattering kernels are organized into a two volume magnetic tape library from which they may be retrieved easily for use in any 30-group THERMOS library. The contents of the tapes are as follows - (Material: ZA/Temperatures (degrees K)): Molecular H 2 O: 100.0/296, 350, 400, 450, 500, 600, Molecular D 2 O: 101.0/296, 350, 400, 450, 500, 600, Graphite: 6000.0/296, 400, 500, 600, 700, 800, Polyethylene: 205.0/296, 350 Benzene: 106.0/296, 350, 400, 450, 500, 600, Zr bound in ZrHx: 203.0/296, 400, 500, 600, 700, 800, H bound in ZrHx: 230.0/296, 400, 500, 600, 700, 800, Beryllium-9: 4009.0/296, 400, 500, 600, 700, 800, Beryllium Oxide: 200.0/296, 400, 500, 600, 700, 800, Uranium Dioxide: 207.0/296, 400, 500, 600, 700, 800 2 - Method of solution: Kernel generation is performed by direct integration of the thermal scattering law data to obtain the differential scattering cross sections for each Legendre order. The integral parameter calculation is done by precision integration of the diffusion length equation for several moderator absorption cross sections followed by a
CHARTB multigroup transport package
International Nuclear Information System (INIS)
Baker, L.
1979-03-01
The physics and numerical implementation of the radiation transport routine used in the CHARTB MHD code are discussed. It is a one-dimensional (Cartesian, cylindrical, and spherical symmetry), multigroup,, diffusion approximation. Tests and applications will be discussed as well
International Nuclear Information System (INIS)
Ganesan, S.; Muir, D.W.
1992-01-01
Selected neutron reaction nuclear data libraries and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into MATXSR format using the NJOY system on the VAX4000 computer of the IAEA. This document lists the resulting multigroup data libraries. All the multigroup data generated are available cost-free upon request from the IAEA Nuclear Data Section. (author). 9 refs
NDS multigroup cross section libraries
International Nuclear Information System (INIS)
DayDay, N.
1981-12-01
A summary description and documentation of the multigroup cross section libraries which exist at the IAEA Nuclear Data Section are given in this report. The libraries listed are available either on tape or in printed form. (author)
Procedure to Generate the MPACT Multigroup Library
Energy Technology Data Exchange (ETDEWEB)
Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2015-12-17
The CASL neutronics simulator MPACT is under development for the neutronics and T-H coupled simulation for the light water reactor. The objective of this document is focused on reviewing the current procedure to generate the MPACT multigroup library. Detailed methodologies and procedures are included in this document for further discussion to improve the MPACT multigroup library.
Angular correlations in γγ → p0p0 near threshold
International Nuclear Information System (INIS)
Althoff, M.; Braunschweig, W.; Gather, K.; Kirschfink, F.J.; Luebelsmeyer, K.; Martyn, H.U.; Peise, G.; Rimkus, J.; Sander, H.G.; Schmitz, D.
1982-09-01
We present an analysis of rho 0 rho 0 production by two photons in the rho 0 rho 0 invariant mass range from 1.2 to 2.0 GeV. From a study of the angular correlations in the process γγ → rho 0 rho 0 → π + π - π + π - we exclude a dominant contribution from Jsup(P) = 0 - or 2 - states. The data indicate sizeable contributions from Jsup(P) = 0 + for four pion masses Msub(4π) + for Msub(4π) > 1.7 GeV. The data are also well described by a model with isotropic production and uncorrelated isotropic decay of the rho 0 's. The cross section stays high below the nominal rho 0 rho 0 threshold, i.e. Msub(4π) 0 rho 0 production is found to decrease steeply with increasing Msub(4π). Upper limits for the couplings of the iota(1440) and the THETA(1640) to γγ and rho 0 rho 0 are given: GAMMA(iota → γγ) x B(iota → rho 0 rho 0 ) 0 rho 0 ) < 1.2 keV (95% C.L.). (orig.)
The isotope density inverse problem in multigroup neutron transport
International Nuclear Information System (INIS)
Zazula, J.M.
1981-01-01
The inverse problem for stationary multigroup anisotropic neutron transport is discussed in order to search for isotope densities in multielement medium. The spatial- and angular-integrated form of neutron transport equation, in terms of the flux in a group - density of an element spatial correlation, leads to a set of integral functionals for the densities weighted by the group fluxes. Some methods of approximation to make the problem uniquently solvable are proposed. Particularly P 0 angular flux information and the spherically-symetrical geometry of an infinite medium are considered. The numerical calculation using this method related to sooner evaluated direct problem data gives promising agreement with primary densities. This approach would be the basis for further application in an elemental analysis of a medium, using an isotopic neutron source and a moving, energy-dependent neutron detector. (author)
Range calculations using multigroup transport methods
International Nuclear Information System (INIS)
Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.
1979-01-01
Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems
Photodetachment cross sections for He-(4P0)
International Nuclear Information System (INIS)
Compton, R.N.; Alton, G.D.; Pegg, D.J.
1980-01-01
The first measurements are reported of photodetachment cross sections for He - (formed through charge exchange of He + with Ca vapor) over the photon energy range from 1.77 to 2.75 eV. The energies of autodetached electrons from the metastable He - beam have also been determined for the first time. The autodetached electron energy agrees within experimental error (+-0.25 eV) with the known 1s2s2p) 4 P 0 He - energy level. This taken with measurements for the lifetime of He - infers that charge exchange of He + with Ca vapor produces 4 P 0 He -
Generating and verification of ACE-multigroup library for MCNP
International Nuclear Information System (INIS)
Chen Chaobin; Hu Zehua; Chen Yixue; Wu Jun; Yang Shouhai
2012-01-01
The Monte Carlo code MCNP can handle multigroup calculations and a sample multigroup set based on ENDF/B-V, MGXSNP, is available for MCNP for coupled neutron-photon transport. However, this library is not suit- able for all problems, and there is a need for users to be able to generate multigroup libraries tailored to their specific applications. For these purposes CSPT (cross section processing tool) is created to generate multigroup library for MCNP from deterministic multigroup cross sections (GENDF or ANISN format at present). Several ACE-multigroup libraries based on ENDF/B-VII.0 converted and verified in this work, we drawn the conclusion that the CSPT code works correctly and the libraries produced are credible. (authors)
International Nuclear Information System (INIS)
Raskach, K. F.
2012-01-01
In multigroup calculations of reactivity and sensitivity coefficients, methodical errors can appear if the interdependence of multigroup constants is not taken into account. For this effect to be taken into account, so-called implicit components of the aforementioned values are introduced. A simple technique for computing these values is proposed. It is based on the use of subgroup parameters.
A multigroup treatment of radiation transport
International Nuclear Information System (INIS)
Tahir, N.A.; Laing, E.W.; Nicholas, D.J.
1980-12-01
A multi-group radiation package is outlined which will accurately handle radiation transfer problems in laser-produced plasmas. Bremsstrahlung, recombination and line radiation are included as well as fast electron Bremsstrahlung radiation. The entire radiation field is divided into a large number of groups (typically 20), which diffuse radiation energy in real space as well as in energy space, the latter occurring via electron-radiation interaction. Using this model a radiation transport code will be developed to be incorporated into MEDUSA. This modified version of MEDUSA will be used to study radiative preheat effects in laser-compression experiments at the Central Laser Facility, Rutherford Laboratory. The model is also relevant to heavy ion fusion studies. (author)
Multigroup cross section library; WIMS library
International Nuclear Information System (INIS)
Kannan, Umasankari
2000-01-01
The WIMS library has been extensively used in thermal reactor calculations. This multigroup constants library was originally developed from the UKNDL in the late 60's and has been updated in 1986. This library has been distributed with the WIMS-D code by NEA data bank. The references to WIMS library in literature are the 'old' which is the original as developed by the AEA Winfrith and the 'new' which is the current 1986 WIMS library. IAEA has organised a CRP where a new and fully updated WIMS library will soon be available. This paper gives an overview of the definitions of the group constants that go into any basic nuclear data library used for reactor calculations. This paper also outlines the contents of the WIMS library and some of its shortcomings
Multi-group neutron transport theory
International Nuclear Information System (INIS)
Zelazny, R.; Kuszell, A.
1962-01-01
Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author) [fr
Final report [on solving the multigroup diffusion equations
International Nuclear Information System (INIS)
Birkhoff, G.
1975-01-01
Progress achieved in the development of variational methods for solving the multigroup neutron diffusion equations is described. An appraisal is made of the extent to which improved variational methods could advantageously replace difference methods currently used
General multi-group macroscopic modeling for thermo-chemical non-equilibrium gas mixtures
Energy Technology Data Exchange (ETDEWEB)
Liu, Yen, E-mail: yen.liu@nasa.gov; Vinokur, Marcel [NASA Ames Research Center, Moffett Field, California 94035 (United States); Panesi, Marco; Sahai, Amal [University of Illinois, Urbana-Champaign, Illinois 61801 (United States)
2015-04-07
This paper opens a new door to macroscopic modeling for thermal and chemical non-equilibrium. In a game-changing approach, we discard conventional theories and practices stemming from the separation of internal energy modes and the Landau-Teller relaxation equation. Instead, we solve the fundamental microscopic equations in their moment forms but seek only optimum representations for the microscopic state distribution function that provides converged and time accurate solutions for certain macroscopic quantities at all times. The modeling makes no ad hoc assumptions or simplifications at the microscopic level and includes all possible collisional and radiative processes; it therefore retains all non-equilibrium fluid physics. We formulate the thermal and chemical non-equilibrium macroscopic equations and rate coefficients in a coupled and unified fashion for gases undergoing completely general transitions. All collisional partners can have internal structures and can change their internal energy states after transitions. The model is based on the reconstruction of the state distribution function. The internal energy space is subdivided into multiple groups in order to better describe non-equilibrium state distributions. The logarithm of the distribution function in each group is expressed as a power series in internal energy based on the maximum entropy principle. The method of weighted residuals is applied to the microscopic equations to obtain macroscopic moment equations and rate coefficients succinctly to any order. The model’s accuracy depends only on the assumed expression of the state distribution function and the number of groups used and can be self-checked for accuracy and convergence. We show that the macroscopic internal energy transfer, similar to mass and momentum transfers, occurs through nonlinear collisional processes and is not a simple relaxation process described by, e.g., the Landau-Teller equation. Unlike the classical vibrational energy relaxation model, which can only be applied to molecules, the new model is applicable to atoms, molecules, ions, and their mixtures. Numerical examples and model validations are carried out with two gas mixtures using the maximum entropy linear model: one mixture consists of nitrogen molecules undergoing internal excitation and dissociation and the other consists of nitrogen atoms undergoing internal excitation and ionization. Results show that the original hundreds to thousands of microscopic equations can be reduced to two macroscopic equations with almost perfect agreement for the total number density and total internal energy using only one or two groups. We also obtain good prediction of the microscopic state populations using 5-10 groups in the macroscopic equations.
General multi-group macroscopic modeling for thermo-chemical non-equilibrium gas mixtures
Liu, Yen; Panesi, Marco; Sahai, Amal; Vinokur, Marcel
2015-04-01
This paper opens a new door to macroscopic modeling for thermal and chemical non-equilibrium. In a game-changing approach, we discard conventional theories and practices stemming from the separation of internal energy modes and the Landau-Teller relaxation equation. Instead, we solve the fundamental microscopic equations in their moment forms but seek only optimum representations for the microscopic state distribution function that provides converged and time accurate solutions for certain macroscopic quantities at all times. The modeling makes no ad hoc assumptions or simplifications at the microscopic level and includes all possible collisional and radiative processes; it therefore retains all non-equilibrium fluid physics. We formulate the thermal and chemical non-equilibrium macroscopic equations and rate coefficients in a coupled and unified fashion for gases undergoing completely general transitions. All collisional partners can have internal structures and can change their internal energy states after transitions. The model is based on the reconstruction of the state distribution function. The internal energy space is subdivided into multiple groups in order to better describe non-equilibrium state distributions. The logarithm of the distribution function in each group is expressed as a power series in internal energy based on the maximum entropy principle. The method of weighted residuals is applied to the microscopic equations to obtain macroscopic moment equations and rate coefficients succinctly to any order. The model's accuracy depends only on the assumed expression of the state distribution function and the number of groups used and can be self-checked for accuracy and convergence. We show that the macroscopic internal energy transfer, similar to mass and momentum transfers, occurs through nonlinear collisional processes and is not a simple relaxation process described by, e.g., the Landau-Teller equation. Unlike the classical vibrational energy relaxation model, which can only be applied to molecules, the new model is applicable to atoms, molecules, ions, and their mixtures. Numerical examples and model validations are carried out with two gas mixtures using the maximum entropy linear model: one mixture consists of nitrogen molecules undergoing internal excitation and dissociation and the other consists of nitrogen atoms undergoing internal excitation and ionization. Results show that the original hundreds to thousands of microscopic equations can be reduced to two macroscopic equations with almost perfect agreement for the total number density and total internal energy using only one or two groups. We also obtain good prediction of the microscopic state populations using 5-10 groups in the macroscopic equations.
General multi-group macroscopic modeling for thermo-chemical non-equilibrium gas mixtures.
Liu, Yen; Panesi, Marco; Sahai, Amal; Vinokur, Marcel
2015-04-07
This paper opens a new door to macroscopic modeling for thermal and chemical non-equilibrium. In a game-changing approach, we discard conventional theories and practices stemming from the separation of internal energy modes and the Landau-Teller relaxation equation. Instead, we solve the fundamental microscopic equations in their moment forms but seek only optimum representations for the microscopic state distribution function that provides converged and time accurate solutions for certain macroscopic quantities at all times. The modeling makes no ad hoc assumptions or simplifications at the microscopic level and includes all possible collisional and radiative processes; it therefore retains all non-equilibrium fluid physics. We formulate the thermal and chemical non-equilibrium macroscopic equations and rate coefficients in a coupled and unified fashion for gases undergoing completely general transitions. All collisional partners can have internal structures and can change their internal energy states after transitions. The model is based on the reconstruction of the state distribution function. The internal energy space is subdivided into multiple groups in order to better describe non-equilibrium state distributions. The logarithm of the distribution function in each group is expressed as a power series in internal energy based on the maximum entropy principle. The method of weighted residuals is applied to the microscopic equations to obtain macroscopic moment equations and rate coefficients succinctly to any order. The model's accuracy depends only on the assumed expression of the state distribution function and the number of groups used and can be self-checked for accuracy and convergence. We show that the macroscopic internal energy transfer, similar to mass and momentum transfers, occurs through nonlinear collisional processes and is not a simple relaxation process described by, e.g., the Landau-Teller equation. Unlike the classical vibrational energy relaxation model, which can only be applied to molecules, the new model is applicable to atoms, molecules, ions, and their mixtures. Numerical examples and model validations are carried out with two gas mixtures using the maximum entropy linear model: one mixture consists of nitrogen molecules undergoing internal excitation and dissociation and the other consists of nitrogen atoms undergoing internal excitation and ionization. Results show that the original hundreds to thousands of microscopic equations can be reduced to two macroscopic equations with almost perfect agreement for the total number density and total internal energy using only one or two groups. We also obtain good prediction of the microscopic state populations using 5-10 groups in the macroscopic equations.
General multi-group macroscopic modeling for thermo-chemical non-equilibrium gas mixtures
International Nuclear Information System (INIS)
Liu, Yen; Vinokur, Marcel; Panesi, Marco; Sahai, Amal
2015-01-01
This paper opens a new door to macroscopic modeling for thermal and chemical non-equilibrium. In a game-changing approach, we discard conventional theories and practices stemming from the separation of internal energy modes and the Landau-Teller relaxation equation. Instead, we solve the fundamental microscopic equations in their moment forms but seek only optimum representations for the microscopic state distribution function that provides converged and time accurate solutions for certain macroscopic quantities at all times. The modeling makes no ad hoc assumptions or simplifications at the microscopic level and includes all possible collisional and radiative processes; it therefore retains all non-equilibrium fluid physics. We formulate the thermal and chemical non-equilibrium macroscopic equations and rate coefficients in a coupled and unified fashion for gases undergoing completely general transitions. All collisional partners can have internal structures and can change their internal energy states after transitions. The model is based on the reconstruction of the state distribution function. The internal energy space is subdivided into multiple groups in order to better describe non-equilibrium state distributions. The logarithm of the distribution function in each group is expressed as a power series in internal energy based on the maximum entropy principle. The method of weighted residuals is applied to the microscopic equations to obtain macroscopic moment equations and rate coefficients succinctly to any order. The model’s accuracy depends only on the assumed expression of the state distribution function and the number of groups used and can be self-checked for accuracy and convergence. We show that the macroscopic internal energy transfer, similar to mass and momentum transfers, occurs through nonlinear collisional processes and is not a simple relaxation process described by, e.g., the Landau-Teller equation. Unlike the classical vibrational energy relaxation model, which can only be applied to molecules, the new model is applicable to atoms, molecules, ions, and their mixtures. Numerical examples and model validations are carried out with two gas mixtures using the maximum entropy linear model: one mixture consists of nitrogen molecules undergoing internal excitation and dissociation and the other consists of nitrogen atoms undergoing internal excitation and ionization. Results show that the original hundreds to thousands of microscopic equations can be reduced to two macroscopic equations with almost perfect agreement for the total number density and total internal energy using only one or two groups. We also obtain good prediction of the microscopic state populations using 5-10 groups in the macroscopic equations
International Nuclear Information System (INIS)
Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.
1987-01-01
Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory
On the convergence of multigroup discrete-ordinates approximations
International Nuclear Information System (INIS)
Victory, H.D. Jr.; Allen, E.J.; Ganguly, K.
1987-01-01
Our analysis is divided into two distinct parts which we label for convenience as Part A and Part B. In Part A, we demonstrate that the multigroup discrete-ordinates approximations are well-defined and converge to the exact transport solution in any subcritical setting. For the most part, we focus on transport in two-dimensional Cartesian geometry. A Nystroem technique is used to extend the discrete ordinates multigroup approximates to all values of the angular and energy variables. Such an extension enables us to employ collectively compact operator theory to deduce stability and convergence of the approximates. In Part B, we perform a thorough convergence analysis for the multigroup discrete-ordinates method for an anisotropically-scattering subcritical medium in slab geometry. The diamond-difference and step-characteristic spatial approximation methods are each studied. The multigroup neutron fluxes are shown to converge in a Banach space setting under realistic smoothness conditions on the solution. This is the first thorough convergence analysis for the fully-discretized multigroup neutron transport equations
Multigroup Moderation Test in Generalized Structured Component Analysis
Directory of Open Access Journals (Sweden)
Angga Dwi Mulyanto
2016-05-01
Full Text Available Generalized Structured Component Analysis (GSCA is an alternative method in structural modeling using alternating least squares. GSCA can be used for the complex analysis including multigroup. GSCA can be run with a free software called GeSCA, but in GeSCA there is no multigroup moderation test to compare the effect between groups. In this research we propose to use the T test in PLS for testing moderation Multigroup on GSCA. T test only requires sample size, estimate path coefficient, and standard error of each group that are already available on the output of GeSCA and the formula is simple so the user does not need a long time for analysis.
The Suppression of Energy Discretization Errors in Multigroup Transport Calculations
International Nuclear Information System (INIS)
Larsen, Edward
2013-01-01
The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to 'coarsen' the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial an energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.
WIMSD5, Deterministic Multigroup Reactor Lattice Calculations
International Nuclear Information System (INIS)
2004-01-01
1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of
Review of multigroup nuclear cross-section processing
Energy Technology Data Exchange (ETDEWEB)
Trubey, D.K.; Hendrickson, H.R. (comps.)
1978-10-01
These proceedings consist of 18 papers given at a seminar--workshop on ''Multigroup Nuclear Cross-Section Processing'' held at Oak Ridge, Tennessee, March 14--16, 1978. The papers describe various computer code systems and computing algorithms for producing multigroup neutron and gamma-ray cross sections from evaluated data, and experience with several reference data libraries. Separate abstracts were prepared for 13 of the papers. The remaining five have already been cited in ERA, and may be located by referring to the entry CONF-780334-- in the Report Number Index. (RWR)
Microstructural evolution in Fe-0.13P-0.05C steel during compression at elevated temperatures
Mehta, Y.; K, Rajput S.; P, Chaudhari G.; V, Dabhade V.
2018-03-01
The microstructural evolution was studied in order to adjust the processing parameters for hot forming. Fe-0.13P-0.05C steel was subjected to hot compression tests using a thermo-mechanical simulator. The tests were performed at the temperatures ranging from 800°C-950°C. The strain rates chosen at all these temperatures were 0.01, 0.1 and 1 s‑1. The effects of the strain rates and hot compression temperatures on the microstructural aspects of the steel were examined using optical microscopy. The outcomes indicate that the mean grain dimension of the hot compressed Fe-0.13P-0.05C steel escalates with increases in the deformation temperature and also with decreases in strain rate. Dynamic recrystallization was observed to be the instrument of grain refinement. The minimum grain dimension of 5.6 μm was attained at 800°C and 0.1s‑1.
International Nuclear Information System (INIS)
Wang Xuan; Chen Xiaofei
2009-01-01
Objective: To evaluate the clinical efficacy of transcatheter hepatic arterial thermo-chemotherapy and thermo-lipiodol embolization in the treatment of hepatic metastases from colorectal carcinoma. Methods: Sixty-eight cases with hepatic metastases from colorectal carcinoma were equally and randomly divided into two groups. The patients in study group were treated with transcatheter hepatic arterial thermo-chemotherapy and thermo-lipiodol embolization, while the patients in control group were treated with conventional (normal temperature) transcatheter hepatic arterial chemotherapy lipiodol embolization. Results: The effective rate of study group and control group was 65%(22/34) and 32%(11/34) respectively, the difference between two groups was statistically significant (P<0.05). No significant difference in the postoperative changes of hepatic function tests was found between the two groups. The survival rate at 6,12,18 and 24 months after the treatment was 100%, 82%, 44% and 18% respectively in study group, while it was 91%, 47%, 15% and 6% respectively in control group. Conclusion: Transcatheter hepatic arterial thermo-chemotherapy and thermo-lipiodol embolization is an effective and safe treatment for the hepatic metastases from colorectal carcinoma and has no obvious damage to the hepatic function. (authors)
On the calculation of multi-group fission spectrum vectors
International Nuclear Information System (INIS)
Mueller, E.Z.
1984-05-01
In this report, the problem of calculating fission spectrum vectors in a consistent manner is formulated. The practical implications of using fission spectrum vectors in multi-group transport calculations are also addressed. The significance of the weighting spectra used for the calculation of fission spectrum vectors is illustrated for the case of a simple neutronic assembly
FINELM: a multigroup finite element diffusion code. Part II
International Nuclear Information System (INIS)
Davierwalla, D.M.
1981-05-01
The author presents the axisymmetric case in cylindrical coordinates for the finite element multigroup neutron diffusion code, FINELM. The numerical acceleration schemes incorporated viz. the Lebedev extrapolations and the coarse mesh rebalancing, space collapsing, are discussed. A few benchmark computations are presented as validation of the code. (Auth.)
RZ calculations for self shielded multigroup cross sections
Energy Technology Data Exchange (ETDEWEB)
Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)
2006-07-01
A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)
RZ calculations for self shielded multigroup cross sections
International Nuclear Information System (INIS)
Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.
2006-01-01
A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)
Calculation of multigroup reaction rates for the Ghana Research ...
African Journals Online (AJOL)
The discrete ordinate spatial model, which pro-vides solution to the differential form of the transport equation by the Carlson-SN (N=4) approach was adopted to solve the Ludwig-Boltzmann multigroup neutron transport equation for this analysis. The results show that for any fissile resonance absorber, the reaction rates ...
International Nuclear Information System (INIS)
Georges, J.-L.; Veyret, J.-F.
1973-01-01
Description is given of a thermo-pump for electrically conductive liquid fluids, e.g. for a liquid metal such as sodium. This pump is characterized in that the piping for the circulation of the conductive liquid is constituted by a plurality of conduits defined by two co-axial cylinders and two walls parallel to their axis. Each conduit limited outside by a magnet, inside by a mild-iron tube, and laterally by two materials forming a thermocouple. The electric current generated by that thermo-couple and the magnetic flux generated by the magnets both loop the loop through an outer cylindrical nickel shell. This can be applied to sodium circulation loops for testing nuclear fuel elements [fr
Surface enhanced thermo lithography
Coluccio, Maria Laura
2017-01-13
We used electroless deposition to fabricate clusters of silver nanoparticles (NPs) on a silicon substrate. These clusters are plasmonics devices that induce giant electromagnetic (EM) field increments. When those EM field are absorbed by the metal NPs clusters generate, in turn, severe temperature increases. Here, we used the laser radiation of a conventional Raman set-up to transfer geometrical patterns from a template of metal NPs clusters into a layer of thermo sensitive Polyphthalaldehyde (PPA) polymer. Temperature profile on the devices depends on specific arrangements of silver nanoparticles. In plane temperature variations may be controlled with (i) high nano-meter spatial precision and (ii) single Kelvin temperature resolution on varying the shape, size and spacing of metal nanostructures. This scheme can be used to generate strongly localized heat amplifications for applications in nanotechnology, surface enhanced thermo-lithography (SETL), biology and medicine (for space resolved cell ablation and treatment), nano-chemistry.
Surface enhanced thermo lithography
Coluccio, Maria Laura; Alabastri, Alessandro; Bonanni, Simon; Majewska, Roksana; Dattoli, Elisabetta; Barberio, Marianna; Candeloro, Patrizio; Perozziello, Gerardo; Mollace, Vincenzo; Di Fabrizio, Enzo M.; Gentile, Francesco
2017-01-01
We used electroless deposition to fabricate clusters of silver nanoparticles (NPs) on a silicon substrate. These clusters are plasmonics devices that induce giant electromagnetic (EM) field increments. When those EM field are absorbed by the metal NPs clusters generate, in turn, severe temperature increases. Here, we used the laser radiation of a conventional Raman set-up to transfer geometrical patterns from a template of metal NPs clusters into a layer of thermo sensitive Polyphthalaldehyde (PPA) polymer. Temperature profile on the devices depends on specific arrangements of silver nanoparticles. In plane temperature variations may be controlled with (i) high nano-meter spatial precision and (ii) single Kelvin temperature resolution on varying the shape, size and spacing of metal nanostructures. This scheme can be used to generate strongly localized heat amplifications for applications in nanotechnology, surface enhanced thermo-lithography (SETL), biology and medicine (for space resolved cell ablation and treatment), nano-chemistry.
Cyclotron radiation by a multi-group method
International Nuclear Information System (INIS)
Chu, T.C.
1980-01-01
A multi-energy group technique is developed to study conditions under which cyclotron radiation emission can shift a Maxwellian electron distribution into a non-Maxwellian; and if the electron distribution is non-Maxwellian, to study the rate of cyclotron radiation emission as compared to that emitted by a Maxwellian having the same mean electron density and energy. The assumptions in this study are: the electrons should be in an isotropic medium and the magnetic field should be uniform. The multi-group technique is coupled into a multi-group Fokker-Planck computer code to study electron behavior under the influence of cyclotron radiation emission in a self-consistent fashion. Several non-Maxwellian distributions were simulated to compare their cyclotron emissions with the corresponding energy and number density equivalent Maxwellian distribtions
The Multigroup Neutron Diffusion Equations/1 Space Dimension
Energy Technology Data Exchange (ETDEWEB)
Linde, Sven
1960-06-15
A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix.
Nuclear data processing and multigroup cross section generation
International Nuclear Information System (INIS)
Kim, Jeong Do; Kil, Chung Sub
1996-01-01
The multigroup constants for WIMS/CASMO were updated with ENDF/B-VI and were tested. The continuous energy MCNP library developed last year was validated against the LWR-simulated critical experiments. The MCNP library will be used to design and analyze nuclear and shielding facilities. The system for generation of MATXS multigroup library and TRANSX code, which is able to prepare the data for the discrete ordinates and diffusion codes from the MATXS library, was established. The MATXS libraries for analyses of thermal and fast critical experiments were generated and tested. The MATXS/TRANSX system for the discrete ordinates and diffusion codes will be useful for nuclear analyses. 10 tabs., 5 figs., 17 refs. (Author)
The Multigroup Neutron Diffusion Equations/1 Space Dimension
International Nuclear Information System (INIS)
Linde, Sven
1960-06-01
A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix
Scalable Multi-group Key Management for Advanced Metering Infrastructure
Benmalek , Mourad; Challal , Yacine; Bouabdallah , Abdelmadjid
2015-01-01
International audience; Advanced Metering Infrastructure (AMI) is composed of systems and networks to incorporate changes for modernizing the electricity grid, reduce peak loads, and meet energy efficiency targets. AMI is a privileged target for security attacks with potentially great damage against infrastructures and privacy. For this reason, Key Management has been identified as one of the most challenging topics in AMI development. In this paper, we propose a new Scalable multi-group key ...
Optimal calculational schemes for solving multigroup photon transport problem
International Nuclear Information System (INIS)
Dubinin, A.A.; Kurachenko, Yu.A.
1987-01-01
A scheme of complex algorithm for solving multigroup equation of radiation transport is suggested. The algorithm is based on using the method of successive collisions, the method of forward scattering and the spherical harmonics method, and is realized in the FORAP program (FORTRAN, BESM-6 computer). As an example the results of calculating reactor photon transport in water are presented. The considered algorithm being modified may be used for solving neutron transport problems
Multi-group diffusion perturbation calculation code. PERKY (2002)
Energy Technology Data Exchange (ETDEWEB)
Iijima, Susumu; Okajima, Shigeaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-12-01
Perturbation calculation code based on the diffusion theory ''PERKY'' is designed for nuclear characteristic analyses of fast reactor. The code calculates reactivity worth on the multi-group diffusion perturbation theory in two or three dimensional core model and kinetics parameters such as effective delayed neutron fraction, prompt neutron lifetime and absolute reactivity scale factor ({rho}{sub 0} {delta}k/k) for FCA experiments. (author)
Multigroup or multipoint thermal neutron data preparation. Programme SIGMA
International Nuclear Information System (INIS)
Matausek, M.V.; Kunc, M.
1974-01-01
When calculating the space energy distribution of thermal neutrons in reactor lattices, in either the multigroup or the multipoint approximation, it is convenient to divide the problem into two independent parts. Firstly, for all material regions of the given reactor lattice cell, the group or the point values of cross sections, scattering kernel and the outer source of thermal neutrons are calculated by a data preparation programme. These quantities are then used as input, by the programme which solves multigroup or multipoint transport equations, to generate the space energy neutron spectra in the cell considered and to determine the related integral quantities, namely the different reaction rates. The present report deals with the first part of the problem. An algorithm for constructing a set of thermal neutron input data, to be used with the multigroup or multipoint version of the code MULTI /1,2,3/, is presented and the new version of the programme SIGMA /4/, written in FORTRAN IV for the CDC-3600 computer, is described. For a given reactor cell material, composed of a number of different isotopes, this programme calculates the group or the point values of the scattering macroscopic absorption cross section, macroscopic scattering cross section, kernel and the outer source of thermal neutrons. Numerous options are foreseen in the programme, concerning the energy variation of cross sections and a scattering kernel, concerning the weighting spectrum in multigroup scheme or the procedure for constructing the scattering matrix in the multipoint scheme and, finally, concerning the organization of output. The details of the calculational algorithm are presented in Section 2 of the paper. Section 3 contains the description of the programme and the instructions for its use (author)
CASTRO: A NEW COMPRESSIBLE ASTROPHYSICAL SOLVER. III. MULTIGROUP RADIATION HYDRODYNAMICS
International Nuclear Information System (INIS)
Zhang, W.; Almgren, A.; Bell, J.; Howell, L.; Burrows, A.; Dolence, J.
2013-01-01
We present a formulation for multigroup radiation hydrodynamics that is correct to order O(v/c) using the comoving-frame approach and the flux-limited diffusion approximation. We describe a numerical algorithm for solving the system, implemented in the compressible astrophysics code, CASTRO. CASTRO uses a Eulerian grid with block-structured adaptive mesh refinement based on a nested hierarchy of logically rectangular variable-sized grids with simultaneous refinement in both space and time. In our multigroup radiation solver, the system is split into three parts: one part that couples the radiation and fluid in a hyperbolic subsystem, another part that advects the radiation in frequency space, and a parabolic part that evolves radiation diffusion and source-sink terms. The hyperbolic subsystem and the frequency space advection are solved explicitly with high-order Godunov schemes, whereas the parabolic part is solved implicitly with a first-order backward Euler method. Our multigroup radiation solver works for both neutrino and photon radiation.
Nuclear data and multigroup methods in fast reactor calculations
International Nuclear Information System (INIS)
Gur, Y.
1975-03-01
The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)
The Polerovirus F box protein P0 targets ARGONAUTE1 to suppress RNA silencing.
Bortolamiol, Diane; Pazhouhandeh, Maghsoud; Marrocco, Katia; Genschik, Pascal; Ziegler-Graff, Véronique
2007-09-18
Plants employ post-transcriptional gene silencing (PTGS) as an antiviral defense response. In this mechanism, viral-derived small RNAs are incorporated into the RNA-induced silencing complex (RISC) to guide degradation of the corresponding viral RNAs. ARGONAUTE1 (AGO1) is a key component of RISC: it carries the RNA slicer activity. As a counter-defense, viruses have evolved various proteins that suppress PTGS. Recently, we showed that the Polerovirus P0 protein carries an F box motif required to form an SCF-like complex, which is also essential for P0's silencing suppressor function. Here, we investigate the molecular mechanism by which P0 impairs PTGS. First we show that P0's expression does not affect the biogenesis of primary siRNAs in an inverted repeat-PTGS assay, but it does affect their activity. Moreover, P0's expression in transformed Arabidopsis plants leads to various developmental abnormalities reminiscent of mutants affected in miRNA pathways, which is accompanied by enhanced levels of several miRNA-target transcripts, suggesting that P0 acts at the level of RISC. Interestingly, ectopic expression of P0 triggered AGO1 protein decay in planta. Finally, we provide evidence that P0 physically interacts with AGO1. Based on these results, we propose that P0 hijacks the host SCF machinery to modulate gene silencing by destabilizing AGO1.
International Nuclear Information System (INIS)
Barros, R.C. de; Larsen, E.W.
1991-01-01
A generalization of the one-group Spectral Green's Function (SGF) method is developed for multigroup, slab-geometry discrete ordinates (S N ) problems. The multigroup SGF method is free from spatial truncation errors; it generated numerical values for the cell-edge and cell-average angular fluxes that agree with the analytic solution of the multigroup S N equations. Numerical results are given to illustrate the method's accuracy
Multi-Group Covariance Data Generation from Continuous-Energy Monte Carlo Transport Calculations
International Nuclear Information System (INIS)
Lee, Dong Hyuk; Shim, Hyung Jin
2015-01-01
The sensitivity and uncertainty (S/U) methodology in deterministic tools has been utilized for quantifying uncertainties of nuclear design parameters induced by those of nuclear data. The S/U analyses which are based on multi-group cross sections can be conducted by an simple error propagation formula with the sensitivities of nuclear design parameters to multi-group cross sections and the covariance of multi-group cross section. The multi-group covariance data required for S/U analysis have been produced by nuclear data processing codes such as ERRORJ or PUFF from the covariance data in evaluated nuclear data files. However in the existing nuclear data processing codes, an asymptotic neutron flux energy spectrum, not the exact one, has been applied to the multi-group covariance generation since the flux spectrum is unknown before the neutron transport calculation. It can cause an inconsistency between the sensitivity profiles and the covariance data of multi-group cross section especially in resolved resonance energy region, because the sensitivities we usually use are resonance self-shielded while the multi-group cross sections produced from an asymptotic flux spectrum are infinitely-diluted. In order to calculate the multi-group covariance estimation in the ongoing MC simulation, mathematical derivations for converting the double integration equation into a single one by utilizing sampling method have been introduced along with the procedure of multi-group covariance tally
Energy Technology Data Exchange (ETDEWEB)
Reither, M; Schorn, B; Schneider, E
1981-01-01
The development of paediatric radiology which began in the late 195O's has been characterised by the need to limit the dose of ionising radiation to which the child is subjected. The aim has been to keep radiation exposure as low as possible by the introduction of suitable techniques and by the development of new methods. It is therefore surprising that studies in dosimetry in the paediaytric age range have only been carried out in recent years. One reason for this may have been the fact that a suitable technique of measurement was not available at the time. The introduction of solid state dosimetry based on thermo-luminescence, first into radiotherapy (1968) and subsequently into radiodiagnosis, has made it possible to abandon the previously widely used ionisation chamber. The purpose of the present paper is to indicate the suitability of this form of dose measurement for paediatric radiological purposes and to stimulate its application in this field.
The developmental outcomes of P0-mediated ARGONAUTE destabilization in tomato.
Hendelman, Anat; Kravchik, Michael; Stav, Ran; Zik, Moriyah; Lugassi, Nitsan; Arazi, Tzahi
2013-01-01
The plant protein ARGONAUTE1 (AGO1) functions in multiple RNA-silencing pathways, including those of microRNAs, key regulators of growth and development. Genetic analysis of ago1 mutants with informative defects has provided valuable insights into AGO1's biological functions. Tomato encodes two AGO1 homologs (SlAGO1s), but mutants have not been described to date. To analyze SlAGO1s' involvement in development, we confirmed that both undergo decay in the presence of the Polerovirus silencing suppressor P0 and produce a transgenic responder line (OP:P0HA) that, upon transactivation, expresses P0 C-terminally fused to a hemagglutinin (HA) tag (P0HA) and destabilizes SlAGO1s at the site of expression. By crossing OP:P0HA with a battery of driver lines, constitutive as well as organ- and stage-specific SlAGO1 downregulation was induced in the F1 progeny. Activated plants exhibited various developmental phenotypes that partially overlapped with those of Arabidopsis ago1 mutants. Plants that constitutively expressed P0HA had reduced SlAGO1 levels and increased accumulation of miRNA targets, indicating compromised SlAGO1-mediated silencing. Consistent with this, they exhibited pleiotropic morphological defects and their growth was arrested post-germination. Transactivation of P0HA in young leaf and floral organ primordia dramatically modified corresponding organ morphology, including the radialization of leaflets, petals and anthers, suggesting that SlAGO1s' activities are required for normal lateral organ development and polarity. Overall, our results suggest that the OP:P0HA responder line can serve as a valuable tool to suppress SlAGO1 silencing pathways in tomato. The suppression of additional SlAGOs by P0HA and its contribution to the observed phenotypes awaits investigation.
From Fourier Transforms to Singular Eigenfunctions for Multigroup Transport
International Nuclear Information System (INIS)
Ganapol, B.D.
2001-01-01
A new Fourier transform approach to the solution of the multigroup transport equation with anisotropic scattering and isotropic source is presented. Through routine analytical continuation, the inversion contour is shifted from the real line to produce contributions from the poles and cuts in the complex plane. The integrand along the branch cut is then recast in terms of matrix continuum singular eigenfunctions, demonstrating equivalence of Fourier transform inversion and the singular eigenfunction expansion. The significance of this paper is that it represents the initial step in revealing the intimate connection between the Fourier transform and singular eigenfunction approaches as well as serves as a basis for a numerical algorithm
Status of multigroup cross-section data for shielding applications
International Nuclear Information System (INIS)
Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.
1983-01-01
Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V
Adjustement of multigroup cross sections using fast reactor integral data
International Nuclear Information System (INIS)
Renke, C.A.C.
1982-01-01
A methodology for the adjustment of multigroup cross section is presented, structured with aiming to compatibility the limitated number of measured values of integral parameters known and disponible, and the great number of cross sections to be adjusted the group of cross section used is that obtained from the Carnaval II calculation system, understanding as formular the sets of calculation methods and data bases. The adjustment is realized, using the INCOAJ computer code, developed in function of one statistical formulation, structural from the bayer considerations, taking in account the measurement processes of cross section and integral parameters defined on statistical bases. (E.G.) [pt
SERKON program for compiling a multigroup library to be used in BETTY calculation
International Nuclear Information System (INIS)
Nguyen Phuoc Lan.
1982-11-01
A SERKON-type program was written to compile data sets generated by FEDGROUP-3 into a multigroup library for BETTY calculation. A multigroup library was generated from the ENDF/B-IV data file and tested against the TRX-1 and TRX-2 lattices with good results. (author)
MUXS: a code to generate multigroup cross sections for sputtering calculations
International Nuclear Information System (INIS)
Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.
1982-10-01
This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc
Semi-continuous and multigroup models in extended kinetic theory
International Nuclear Information System (INIS)
Koller, W.
2000-01-01
The aim of this thesis is to study energy discretization of the Boltzmann equation in the framework of extended kinetic theory. In case that external fields can be neglected, the semi- continuous Boltzmann equation yields a sound basis for various generalizations. Semi-continuous kinetic equations describing a three component gas mixture interacting with monochromatic photons as well as a four component gas mixture undergoing chemical reactions are established and investigated. These equations reflect all major aspects (conservation laws, equilibria, H-theorem) of the full continuous kinetic description. For the treatment of the spatial dependence, an expansion of the distribution function in terms of Legendre polynomials is carried out. An implicit finite differencing scheme is combined with the operator splitting method. The obtained numerical schemes are applied to the space homogeneous study of binary chemical reactions and to spatially one-dimensional laser-induced acoustic waves. In the presence of external fields, the developed overlapping multigroup approach (with the spline-interpolation as its extension) is well suited for numerical studies. Furthermore, two formulations of consistent multigroup approaches to the non-linear Boltzmann equation are presented. (author)
International Nuclear Information System (INIS)
Calloo, A.A.
2012-01-01
, heterogeneous clusters and 1D core-reflector calculations. The main results are given below: - a P3 expansion is sufficient to model the scattering law with respect to the biases due to the other approximations used for calculations (self-shielding, spatial resolution method). This order of expansion is converged for anisotropy representation in the modelling of light water reactors. - the transport correction, P0c is not suited for calculations, especially for B 4 C absorbent. (author) [fr
Induced Pluripotent Stem Cells Generated from P0-Cre;Z/EG Transgenic Mice.
Ogawa, Yasuhiro; Eto, Akira; Miyake, Chisato; Tsuchida, Nana; Miyake, Haruka; Takaku, Yasuhiro; Hagiwara, Hiroaki; Oishi, Kazuhiko
2015-01-01
Neural crest (NC) cells are a migratory, multipotent cell population that arises at the neural plate border, and migrate from the dorsal neural tube to their target tissues, where they differentiate into various cell types. Abnormal development of NC cells can result in severe congenital birth defects. Because only a limited number of cells can be obtained from an embryo, mechanistic studies are difficult to perform with directly isolated NC cells. Protein zero (P0) is expressed by migrating NC cells during the early embryonic period. In the P0-Cre;Z/EG transgenic mouse, transient activation of the P0 promoter induces Cre-mediated recombination, indelibly tagging NC-derived cells with enhanced green fluorescent protein (EGFP). Induced pluripotent stem cell (iPSC) technology offers new opportunities for both mechanistic studies and development of stem cell-based therapies. Here, we report the generation of iPSCs from the P0-Cre;Z/EG mouse. P0-Cre;Z/EG mouse-derived iPSCs (P/G-iPSCs) exhibited pluripotent stem cell properties. In lineage-directed differentiation studies, P/G-iPSCs were efficiently differentiated along the neural lineage while expressing EGFP. These results suggest that P/G-iPSCs are useful to study NC development and NC-associated diseases.
Induced Pluripotent Stem Cells Generated from P0-Cre;Z/EG Transgenic Mice.
Directory of Open Access Journals (Sweden)
Yasuhiro Ogawa
Full Text Available Neural crest (NC cells are a migratory, multipotent cell population that arises at the neural plate border, and migrate from the dorsal neural tube to their target tissues, where they differentiate into various cell types. Abnormal development of NC cells can result in severe congenital birth defects. Because only a limited number of cells can be obtained from an embryo, mechanistic studies are difficult to perform with directly isolated NC cells. Protein zero (P0 is expressed by migrating NC cells during the early embryonic period. In the P0-Cre;Z/EG transgenic mouse, transient activation of the P0 promoter induces Cre-mediated recombination, indelibly tagging NC-derived cells with enhanced green fluorescent protein (EGFP. Induced pluripotent stem cell (iPSC technology offers new opportunities for both mechanistic studies and development of stem cell-based therapies. Here, we report the generation of iPSCs from the P0-Cre;Z/EG mouse. P0-Cre;Z/EG mouse-derived iPSCs (P/G-iPSCs exhibited pluripotent stem cell properties. In lineage-directed differentiation studies, P/G-iPSCs were efficiently differentiated along the neural lineage while expressing EGFP. These results suggest that P/G-iPSCs are useful to study NC development and NC-associated diseases.
Na3Co2(As0.52P0.48)O4(As0.95P0.05)2O7.
Ben Smida, Youssef; Guesmi, Abderrahmen; Zid, Mohamed Faouzi; Driss, Ahmed
2013-11-30
The title compound, trisodium dicobalt(II) (arsenate/phosphate) (diarsenate/diphosphate), was prepared by a solid-state reaction. It is isostructural with Na3Co2AsO4As2O7. The framework shows the presence of CoX22O12 (X2 is statistically disordered with As0.95P0.05) units formed by sharing corners between Co1O6 octa-hedra and X22O7 groups. These units form layers perpendicular to [010]. Co2O6 octa-hedra and X1O4 (X1 = As0.54P0.46) tetra-hedra form Co2X1O8 chains parallel to [001]. Cohesion between layers and chains is ensured by the X22O7 groups, giving rise to a three-dimensional framework with broad tunnels, running along the a- and c-axis directions, in which the Na(+) ions reside. The two Co(2+) cations, the X1 site and three of the seven O atoms lie on special positions, with site symmetries 2 and m for the Co, m for the X1, and 2 and m (× 2) for the O sites. One of two Na atoms is disordered over three special positions [occupancy ratios 0.877 (10):0.110 (13):0.066 (9)] and the other is in a general position with full occupancy. A comparison between structures such as K2CdP2O7, α-NaTiP2O7 and K2MoO2P2O7 is made. The proposed structural model is supported by charge-distribution (CHARDI) analysis and bond-valence-sum (BVS) calculations. The distortion of the coordination polyhedra is analyzed by means of the effective coordination number.
ERRORJ, Multigroup covariance matrices generation from ENDF-6 format
International Nuclear Information System (INIS)
Chiba, Go
2007-01-01
1 - Description of program or function: ERRORJ produces multigroup covariance matrices from ENDF-6 format following mainly the methods of the ERRORR module in NJOY94.105. New version differs from previous version in the following features: Additional features in ERRORJ with respect to the NJOY94.105/ERRORR module: - expands processing for the covariance matrices of resolved and unresolved resonance parameters; - processes average cosine of scattering angle and fission spectrum; - treats cross-correlation between different materials and reactions; - accepts input of multigroup constants with various forms (user input, GENDF, etc.); - outputs files with various formats through utility NJOYCOVX (COVERX format, correlation matrix, relative error and standard deviation); - uses a 1% sensitivity method for processing of resonance parameters; - ERRORJ can process the JENDL-3.2 and 3.3 covariance matrices. Additional features of the version 2 with respect to the previous version of ERRORJ: - Since the release of version 2, ERRORJ has been modified to increase its reliability and stability, - calculation of the correlation coefficients in the resonance region, - Option for high-speed calculation is implemented, - Perturbation amount is optimised in a sensitivity calculation, - Effect of the resonance self-shielding can be considered, - a compact covariance format (LCOMP=2) proposed by N. M. Larson can be read. Additional features of the version 2.2.1 with respect to the previous version of ERRORJ: - Several routines were modified to reduce calculation time. The new one needs shorter calculation time (50-70%) than the old version without changing results. - In the U-233 and Pu-241 files of JENDL-3.3 an inconsistency between resonance parameters in MF=32 and those in MF=2 was corrected. NEA-1676/06: This version differs from the previous one (NEA-1676/05) in the following: ERRORJ2.2.1 was modified to treat the self-shielding effect accurately. NEA-1676/07: This version
Hyperfine quenching of the 23P0 state in heliumlike ions
International Nuclear Information System (INIS)
Mohr, P.J.
1975-01-01
An estimate is presented of the lifetime of the 2 3 P 0 state for odd-Z heliumlike ions in the range Z = 9 to 29. An approximation scheme is employed which utilizes the fact that both Z -1 and (Zα) 2 are small parameters for the range of Z under consideration. 1 fig, 2 tables, 14 refs
General method of calculation of any hadronic decay in the 3P0 model
International Nuclear Information System (INIS)
Roberts, W.
1992-01-01
The 3 P 0 pair creation model of hadron decays is generalized to be applicable to the decay of any hadron. The wave function of the decaying hadron is expanded in terms of two clusters. The transition amplitudes is derived for any combination of angular momenta, and for general wave functions in momentum space, expanded in terms of Gaussians times polynomials. (authors)
Properties of the malarial proteins Pf2, Pf9 and PfP0, which support ...
Indian Academy of Sciences (India)
Properties of the malarial proteins Pf2, Pf9 and PfP0, which support their roles as immune targets. Antibodies raised to each of these proteins (or purified from immune adults) inhibit the growth of Plasmodium falciparum at the red cell invasion step. The proteins are localized on the parasite cell surface. Each protein is ...
The Polerovirus silencing suppressor P0 targets ARGONAUTE proteins for degradation.
Baumberger, Nicolas; Tsai, Ching-Hsui; Lie, Miranda; Havecker, Ericka; Baulcombe, David C
2007-09-18
Plant and animal viruses encode suppressor proteins of an adaptive immunity mechanism in which viral double-stranded RNA is processed into 21-25 nt short interfering (si)RNAs. The siRNAs guide ARGONAUTE (AGO) proteins so that they target viral RNA. Most viral suppressors bind long dsRNA or siRNAs and thereby prevent production of siRNA or binding of siRNA to AGO. The one exception is the 2b suppressor of Cucumoviruses that binds to and inhibits AGO1. Here we describe a novel suppressor mechanism in which a Polerovirus-encoded F box protein (P0) targets the PAZ motif and its adjacent upstream sequence in AGO1 and mediates its degradation. F box proteins are components of E3 ubiquitin ligase complexes that add polyubiquitin tracts on selected lysine residues and thereby mark a protein for proteasome-mediated degradation. With P0, however, the targeted degradation of AGO is insensitive to inhibition of the proteasome, indicating that the proteasome is not involved. We also show that P0 does not block a mobile signal of silencing, indicating that the signal molecule does not have AGO protein components. The ability of P0 to block silencing without affecting signal movement may contribute to the phloem restriction of viruses in the Polerovirus group.
Multigroup P8 - elastic scattering matrices of main reactor elements
International Nuclear Information System (INIS)
Garg, S.B.; Shukla, V.K.
1979-01-01
To study the effect of anisotropic scattering phenomenon on shielding and neutronics of nuclear reactors multigroup P8-elastic scattering matrices have been generated for H, D, He, 6 Li, 7 Li, 10 B, C, N, O, Na, Cr, Fe, Ni, 233 U, 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu using their angular distribution, Legendre coefficient and elastic scattering cross-section data from the basic ENDF/B library. Two computer codes HSCAT and TRANS have been developed to complete this task for BESM-6 and CDC-3600 computers. These scattering matrices can be directly used as input to the transport theory codes ANISN and DOT. (auth.)
MORET: Version 4.B. A multigroup Monte Carlo criticality code
International Nuclear Information System (INIS)
Jacquet, Olivier; Miss, Joachim; Courtois, Gerard
2003-01-01
MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)
Multi-group dynamic quantum secret sharing with single photons
Energy Technology Data Exchange (ETDEWEB)
Liu, Hongwei [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Ma, Haiqiang, E-mail: hqma@bupt.edu.cn [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Wei, Kejin [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Yang, Xiuqing [School of Science, Beijing Jiaotong University, Beijing 100044 (China); Qu, Wenxiu; Dou, Tianqi; Chen, Yitian; Li, Ruixue; Zhu, Wu [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China)
2016-07-15
In this letter, we propose a novel scheme for the realization of single-photon dynamic quantum secret sharing between a boss and three dynamic agent groups. In our system, the boss can not only choose one of these three groups to share the secret with, but also can share two sets of independent keys with two groups without redistribution. Furthermore, the security of communication is enhanced by using a control mode. Compared with previous schemes, our scheme is more flexible and will contribute to a practical application. - Highlights: • A multi-group dynamic quantum secret sharing with single photons scheme is proposed. • Any one of the groups can be chosen to share secret through controlling the polarization of photons. • Two sets of keys can be shared simultaneously without redistribution.
A Laplace transform method for energy multigroup hybrid discrete ordinates
International Nuclear Information System (INIS)
Segatto, C.F.; Vilhena, M.T.; Barros, R.C.
2010-01-01
In typical lattice cells where a highly absorbing, small fuel element is embedded in the moderator, a large weakly absorbing medium, high-order transport methods become unnecessary. In this work we describe a hybrid discrete ordinates (S N) method for energy multigroup slab lattice calculations. This hybrid S N method combines the convenience of a low-order S N method in the moderator with a high-order S N method in the fuel. The idea is based on the fact that in weakly absorbing media whose physical size is several neutron mean free paths in extent, even the S 2 method (P 1 approximation), leads to an accurate result. We use special fuel-moderator interface conditions and the Laplace transform (LTS N ) analytical numerical method to calculate the two-energy group neutron flux distributions and the thermal disadvantage factor. We present numerical results for a range of typical model problems.
Parallel computation of multigroup reactivity coefficient using iterative method
Susmikanti, Mike; Dewayatna, Winter
2013-09-01
One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.
The Thermos process heat reactor
International Nuclear Information System (INIS)
Lerouge, Bernard
1979-01-01
The THERMOS process heat reactor was born from the following idea: the hot water energy vector is widely used for heating purposes in cities, so why not save on traditional fossil fuels by simply substituting a nuclear boiler of comparable power for the classical boiler installed in the same place. The French Atomic Energy Commission has techniques for heating in the big French cities which provide better guarantees for national independence and for the environment. This THERMOS technique would result in a saving of 40,000 to 80,000 tons of oil per year [fr
Discussion of the 3P0 model applied to the decay of mesons into two mesons
International Nuclear Information System (INIS)
Bonnaz, R.; Silvestre-Brac, B.
1999-01-01
The 3 P 0 model for the decay of a meson into two mesons is revisited. In particular, the formalism is extended in order to deal with an arbitrary form for the creation vertex and with the exact meson wave functions. A careful analysis of both effects is performed and discussed. The model is then applied to a large class of transitions known experimentally. Two types of quark-antiquark potentials have been tested and compared. (author)
RGENDF - An interface program between the NJOY code and codes using multigroup cross-sections
International Nuclear Information System (INIS)
Chalhoub, E.S.; Anaf, J.
1988-02-01
An interface program for reformatting multigroup cross-section libraries generated by NJOY into ENDF/B-V format and the EXPANDA, PFCOND and COMPAR input formats is presented. (author). 7 refs, 1 fig., 1 tab
Kalpakkam multigroup cross section set for fast reactor applications - status and performance
International Nuclear Information System (INIS)
Ramanadhan, M.M.; Gopalakrishnan, M.M.
1986-01-01
This report documents the status of the presently created set of multigroup constants at Kalpakkam. The list of nuclides processed and the details of multigroup structure are given. Also included are the particulars of dilutions and temperatures for each nuclide in the multigroup cross section set for which self shielding factors have been calculated. Using this new multigroup cross section set, measured integral quantities such as K-eff, central reaction rate ratios, central reactivity worths etc. were calculated for a few fast critical benchmark assemblies and the calculated values of neutronic parameters obtained were compared with those obtained using the available Cadarache cross section library and those published in literature for ENDF/B-IV based set and Japanese evaluated nuclear data library (JENDL). The details of analyses are documented along with the conclusions. (author). 17 refs., 12 tabs
Energy Technology Data Exchange (ETDEWEB)
Coste-Delclaux, M
2006-03-15
This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)
Energy Technology Data Exchange (ETDEWEB)
Coste-Delclaux, M
2006-03-15
This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)
Cantisano, Gabriela Topa; Domínguez, J Francisco Morales; García, J Luis Caeiro
2007-05-01
This study focuses on the mediator role of social comparison in the relationship between perceived breach of psychological contract and burnout. A previous model showing the hypothesized effects of perceived breach on burnout, both direct and mediated, is proposed. The final model reached an optimal fit to the data and was confirmed through multigroup analysis using a sample of Spanish teachers (N = 401) belonging to preprimary, primary, and secondary schools. Multigroup analyses showed that the model fit all groups adequately.
Complex of two-dimensional multigroup programs for neutron-physical computations of nuclear reactor
International Nuclear Information System (INIS)
Karpov, V.A.; Protsenko, A.N.
1975-01-01
Briefly stated mathematical aspects of the two-dimensional multigroup method of neutron-physical computation of nuclear reactor. Problems of algorithmization and BESM-6 computer realisation of multigroup diffuse approximations in hexagonal and rectangular calculated lattices are analysed. The results of computation of fast critical assembly having complicated composition of the core are given. The estimation of computation accuracy of criticality, neutron fields distribution and efficiency of absorbing rods by means of computer programs developed is done. (author)
Application of direct discrete method (DDM) to multigroup neutron transport problems
International Nuclear Information System (INIS)
Vosoughi, Naser; Salehi, Ali Akbar; Shahriari, Majid
2003-01-01
The Direct Discrete Method (DDM), which produced excellent results for one-group neutron transport problems, has been developed for multigroup energy. A multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without associated coolant regions with two boundary conditions. The calculations are illustrated for two-group energy by graphs showing the fast and thermal fluxes. The validity of the results are tested against the results obtained by the ANISN code. (author)
International Nuclear Information System (INIS)
Halilou, A.; Lounici, A.
1981-01-01
The subject is divided in two parts: In the first part a nodal method has been worked out to solve the steady state multigroup diffusion equation. This method belongs to the same set of nodal methods currently used to calculate the exact fission powers and neutron fluxes in a very short computing time. It has been tested on a two dimensional idealized reactors. The effective multiplication factor and the fission powers for each fuel element have been calculated. The second part consists in studying and mastering the multigroup diffusion code DAHRA - a reduced version of DIANE - a two dimensional code using finite difference method
Macroscopic multigroup constants for accelerator driven system core calculation
International Nuclear Information System (INIS)
Heimlich, Adino; Santos, Rubens Souza dos
2011-01-01
The high-level wastes stored in facilities above ground or shallow repositories, in close connection with its nuclear power plant, can take almost 106 years before the radiotoxicity became of the order of the background. While the disposal issue is not urgent from a technical viewpoint, it is recognized that extended storage in the facilities is not acceptable since these ones cannot provide sufficient isolation in the long term and neither is it ethical to leave the waste problem to future generations. A technique to diminish this time is to transmute these long-lived elements into short-lived elements. The approach is to use an Accelerator Driven System (ADS), a sub-critical arrangement which uses a Spallation Neutron Source (SNS), after separation the minor actinides and the long-lived fission products (LLFP), to convert them to short-lived isotopes. As an advanced reactor fuel, still today, there is a few data around these type of core systems. In this paper we generate macroscopic multigroup constants for use in calculations of a typical ADS fuel, take into consideration, the ENDF/BVI data file. Four energy groups are chosen to collapse the data from ENDF/B-VI data file by PREPRO code. A typical MOX fuel cell is used to validate the methodology. The results are used to calculate one typical subcritical ADS core. (author)
FINELM: a multigroup finite element diffusion code. Part I
International Nuclear Information System (INIS)
Davierwalla, D.M.
1980-12-01
The author presents a two dimensional code for multigroup diffusion using the finite element method. It was realized that the extensive connectivity which contributes significantly to the accuracy, results in a matrix which, although symmetric and positive definite, is wide band and possesses an irregular profile. Hence, it was decided to introduce sparsity techniques into the code. The introduction of the R-Z geometry lead to a great deal of changes in the code since the rotational invariance of the removal matrices in X-Y geometry did not carry over in R-Z geometry. Rectangular elements were introduced to remedy the inability of the triangles to model essentially one dimensional problems such as slab geometry. The matter is discussed briefly in the text in the section on benchmark problems. This report is restricted to the general theory of the triangular elements and to the sparsity techniques viz. incomplete disections. The latter makes the size of the problem that can be handled independent of core memory and dependent only on disc storage capacity which is virtually unlimited. (Auth.)
Travelling Wave Solutions in Multigroup Age-Structured Epidemic Models
Ducrot, Arnaut; Magal, Pierre; Ruan, Shigui
2010-01-01
Age-structured epidemic models have been used to describe either the age of individuals or the age of infection of certain diseases and to determine how these characteristics affect the outcomes and consequences of epidemiological processes. Most results on age-structured epidemic models focus on the existence, uniqueness, and convergence to disease equilibria of solutions. In this paper we investigate the existence of travelling wave solutions in a deterministic age-structured model describing the circulation of a disease within a population of multigroups. Individuals of each group are able to move with a random walk which is modelled by the classical Fickian diffusion and are classified into two subclasses, susceptible and infective. A susceptible individual in a given group can be crisscross infected by direct contact with infective individuals of possibly any group. This process of transmission can depend upon the age of the disease of infected individuals. The goal of this paper is to provide sufficient conditions that ensure the existence of travelling wave solutions for the age-structured epidemic model. The case of two population groups is numerically investigated which applies to the crisscross transmission of feline immunodeficiency virus (FIV) and some sexual transmission diseases.
Multigroup perturbation model for kinetic analysis of nuclear reactors
International Nuclear Information System (INIS)
Souza, G.M.
1989-01-01
The scope of this work is the development of a multigroup perturbation theory for the purpose of Kinetic and dynamic analysis of nuclear reactors. The equations that describe the reactor behavior were presented in all generality and written in the shorthand notation of matrices and vectors. In the derivation of those equations indetermined operators and discretizing factors were introduced and then determined by comparision with conventional equations. Fick's Law was developed in higher orders for neutron and importance current density. The solution of the direct and adjoint fields were represented by combination of the eigenfunctions of the B and B* operators and the eigenvalue modulus equality was established mathematically. In the derivation of the reactivity expression the B operator perturbation was split in two non coupled to the flux form and level. The prompt neutrons effective mean life was derived from reactor equations and importance conservation. The establishment of the Nordheim's equation, although modified, was based on Gandini. Finally, a mathematical interpretation of the flux-trap region was avented. (author)
Wang, Ken-Der; Empleo, Roman; Nguyen, Tan Tri V; Moffett, Peter; Sacco, Melanie Ann
2015-06-01
Plant disease resistance (R) proteins that confer resistance to viruses recognize viral gene products with diverse functions, including viral suppressors of RNA silencing (VSRs). The P0 protein from poleroviruses is a VSR that targets the ARGONAUTE1 (AGO1) protein for degradation, thereby disrupting RNA silencing and antiviral defences. Here, we report resistance against poleroviruses in Nicotiana glutinosa directed against Turnip yellows virus (TuYV) and Potato leafroll virus (PLRV). The P0 proteins from TuYV (P0(T) (u) ), PLRV (P0(PL) ) and Cucurbit aphid-borne yellows virus (P0(CA) ) were found to elicit a hypersensitive response (HR) in N. glutinosa accession TW59, whereas other accessions recognized P0(PL) only. Genetic analysis showed that recognition of P0(T) (u) by a resistance gene designated RPO1 (Resistance to POleroviruses 1) is inherited as a dominant allele. Expression of P0 from a Potato virus X (PVX) expression vector transferred recognition to the recombinant virus on plants expressing RPO1, supporting P0 as the unique Polerovirus factor eliciting resistance. The induction of HR required a functional P0 protein, as P0(T) (u) mutants with substitutions in the F-box motif that abolished VSR activity were unable to elicit HR. We surmised that the broad P0 recognition seen in TW59 and the requirement for the F-box protein motif could indicate detection of P0-induced AGO1 degradation and disruption of RNA silencing; however, other viral silencing suppressors, including the PVX P25 that also causes AGO1 degradation, failed to elicit HR in N. glutinosa. Investigation of P0 elicitation of RPO1 could provide insight into P0 activities within the cell that trigger resistance. © 2014 BSPP AND JOHN WILEY & SONS LTD.
Data-processing program from the operating modes of the nuclear reactor (P0DER)
International Nuclear Information System (INIS)
Totev, T.L.; Boyadzhiev, A.I.
1981-01-01
A program PODER for processing data from the operating modes of the reactors taking into account the effects of corrosion, hydration, and deformation of the nuclear reactor fuel element sheathing, the formation of the corrosion product deposits, the change in the geometric dimensions of the nuclear reactor fuel element due to the temperature deformation, as well as the various gas fillers, are elaborated. The ''hot channel'' method determining the reliability of the system is realized. The basic equations describing the thermohydraulic processes in nuclear reactors are solved by the finite difference method. Approximations are carried out with the approach of least squares. The temperature distribution versus the zirconium sheathing height is computed for the case of WWER-440 type reactors. The advantages of the proposed program P0DER are discussed
Observation of electrons from the 1P0 resonance of D-
International Nuclear Information System (INIS)
Duncan, M.M.; Menendez, M.G.
1989-01-01
We have measured the electron energy spectra near 0 0 produced in collisions of D - with Ar. Using a 400-keV D - beam and with good experimental energy and angular resolution we have found structure in the ejected electron energy spectra which is due to the decay of the 1 P 0 shape resonance. The doubly differential cross sections (DDCS's) have been measured as a function of angle and it was found that this structure disappeared for laboratory angles greater than 1 0 as expected. A resonance contribution to the DDCS's was extracted at θ/sub L/ = 0 0 , transformed to the projectile frame, and fit with a Breit-Wigner shape. Our resonant energy is in reasonable agreement with other experiments. We also find a small asymmetry in the two resonant structures in the laboratory measurements at θ/sub L/ = 0 0
Energy Technology Data Exchange (ETDEWEB)
Park, Ho Jin; Cho, Jin Young [KAERI, Daejeon (Korea, Republic of); Kim, Kang Seog [Oak Ridge National Laboratory, Oak Ridge (United States); Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)
2016-05-15
In this study, multi-group cross section libraries for the DeCART code were generated using a new procedure. The new procedure includes generating the RI tables based on the MC calculations, correcting the effective fission product yield calculations, and considering most of the fission products as resonant nuclides. KAERI (Korea Atomic Energy Research Institute) has developed the transport lattice code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and DeCART (Deterministic Core Analysis based on Ray Tracing) for a multi-group neutron transport analysis of light water reactors (LWRs). These codes adopt the method of characteristics (MOC) to solve the multi-group transport equation and resonance fixed source problem, the subgroup and the direct iteration method with resonance integral tables for resonance treatment. With the development of the DeCART and KARMA code, KAERI has established its own library generation system for a multi-group transport calculation. In the KAERI library generation system, the multi-group average cross section and resonance integral (RI) table are generated and edited using PENDF (point-wise ENDF) and GENDF (group-wise ENDF) produced by the NJOY code. The new method does not need additional processing because the MC method can handle any geometry information and material composition. In this study, the new method is applied to the dominant resonance nuclide such as U{sup 235} and U{sup 238} and the conventional method is applied to the minor resonance nuclides. To examine the newly generated multi-group cross section libraries, various benchmark calculations such as pin-cell, FA, and core depletion problem are performed and the results are compared with the reference solutions. Overall, the results by the new method agree well with the reference solution. The new procedure based on the MC method were verified and provided the multi-group library that can be used in the SMR nuclear design analysis.
International Nuclear Information System (INIS)
Fusaro, Adriana F.; Correa, Regis L.; Nakasugi, Kenlee; Jackson, Craig; Kawchuk, Lawrence; Vaslin, Maite F.S.; Waterhouse, Peter M.
2012-01-01
The P0 protein of poleroviruses and P1 protein of sobemoviruses suppress the plant's RNA silencing machinery. Here we identified a silencing suppressor protein (SSP), P0 PE , in the Enamovirus Pea enation mosaic virus-1 (PEMV-1) and showed that it and the P0s of poleroviruses Potato leaf roll virus and Cereal yellow dwarf virus have strong local and systemic SSP activity, while the P1 of Sobemovirus Southern bean mosaic virus supresses systemic silencing. The nuclear localized P0 PE has no discernable sequence conservation with known SSPs, but proved to be a strong suppressor of local silencing and a moderate suppressor of systemic silencing. Like the P0s from poleroviruses, P0 PE destabilizes AGO1 and this action is mediated by an F-box-like domain. Therefore, despite the lack of any sequence similarity, the poleroviral and enamoviral SSPs have a conserved mode of action upon the RNA silencing machinery.
Almasi, Reza; Miller, W Allen; Ziegler-Graff, Véronique
2015-10-02
Viral pathogenicity has often been correlated to the expression of the viral encoded-RNA silencing suppressor protein (SSP). The silencing suppressor activity of the P0 protein encoded by cereal yellow dwarf virus-RPV (CYDV-RPV) and -RPS (CYDV-RPS), two poleroviruses differing in their symptomatology was investigated. CYDV-RPV displays milder symptoms in oat and wheat whereas CYDV-RPS is responsible for more severe disease. We showed that both P0 proteins (P0(CY-RPV) and P0(CY-RPS)) were able to suppress local RNA silencing induced by either sense or inverted repeat transgenes in an Agrobacterium tumefaciens-mediated expression assay in Nicotiana benthamiana. P0(CY-RPS) displayed slightly higher activity. Systemic spread of the silencing signal was not impaired. Analysis of short-interfering RNA (siRNA) abundance revealed that accumulation of primary siRNA was not affected, but secondary siRNA levels were reduced by both CYDV P0 proteins, suggesting that they act downstream of siRNA production. Correlated with this finding we showed that both P0 proteins partially destabilized ARGONAUTE1. Finally both P0(CY-RPV) and P0(CY-RPS) interacted in yeast cells with ASK2, a component of an E3-ubiquitin ligase, with distinct affinities. Copyright © 2015 Elsevier B.V. All rights reserved.
Multigroup calculations of low-energy neutral transport in tokamak plasmas
International Nuclear Information System (INIS)
Gilligan, J.G.; Gralnick, S.L.; Price, W.G. Jr.; Kammash, T.
1978-01-01
Multigroup discrete ordinates methods avoid many of the approximations that have been used in previous neutral transport analyses. Of particular interest are the neutral profiles generated as an integral part of larger plasma system simulation codes. To determine the appropriateness of utilizing a particular multigroup code, ANISN, for this purpose, results are compared with the neutral transport module of the Duechs code. For a typical TFTR plasma, predicted neutral densities differ by a maximum factor of three on axis and outfluxes at the plasma boundary by approximately 40%. This is found to be significant for a neutral transport module. Possible sources of the observed discrepancies are indicated from an analysis of the approximations used in the Duechs model. Recommendations are made concerning the future application of the multigroup method. (author)
Proposal to extend CSEWG neutron and photon multigroup structures for wider applications
International Nuclear Information System (INIS)
LaBauve, R.J.; Wilson, W.B.
1976-02-01
The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented
Proposal to extend CSEWG neutron and photon multigroup structures for wider applications. [Tables
Energy Technology Data Exchange (ETDEWEB)
LaBauve, R.J.; Wilson, W.B.
1976-02-01
The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented.
International Nuclear Information System (INIS)
Ganapol, B.D.
2011-01-01
Highlights: → Coupled neutron and gamma transport is considered in the multigroup diffusion approximation. → The model accommodates fission, up- and down-scattering and common neutron-gamma interactions. → The exact solution to the diffusion equation in a heterogeneous media of any number of regions is found. → The solution is shown to parallel the one-group case in a homogeneous medium. → The discussion concludes with a heterogeneous, 2 fuel-plate 93.2% enriched reactor fuel benchmark demonstration. - Abstract: The angular flux for the 'rod model' describing coupled neutron/gamma (n, γ) diffusion has a particularly straightforward analytical representation when viewed from the perspective of a one-group homogeneous medium. Cast in the form of matrix functions of a diagonalizable matrix, the solution to the multigroup equations in heterogeneous media is greatly simplified. We shall show exactly how the one-group homogeneous medium solution leads to the multigroup solution.
Thermo Wigner operator in thermo field dynamics: its introduction and application
International Nuclear Information System (INIS)
Fan Hongyi; Jiang Nianquan
2008-01-01
Because in thermo-field dynamics (TFD) the thermo-operator has a neat expression in the thermo-entangled state representation, we need to introduce the thermo-Wigner operator (THWO) in the same representation. We derive the THWO in a direct way, which brings much conveniece to calculating the Wigner functions of thermo states in TFD. We also discuss the condition for existence of a wavefunction corresponding to a given Wigner function in the context of TFD by using the explicit form of the THWO.
Chen, Guiqian; Ishan, Mohamed; Yang, Jingwen; Kishigami, Satoshi; Fukuda, Tomokazu; Scott, Greg; Ray, Manas K; Sun, Chenming; Chen, Shi-You; Komatsu, Yoshihiro; Mishina, Yuji; Liu, Hong-Xiang
2017-06-01
P0-Cre and Wnt1-Cre mouse lines have been widely used in combination with loxP-flanked mice to label and genetically modify neural crest (NC) cells and their derivatives. Wnt1-Cre has been regarded as the gold standard and there have been concerns about the specificity of P0-Cre because it is not clear about the timing and spatial distribution of the P0-Cre transgene in labeling NC cells at early embryonic stages. We re-visited P0-Cre and Wnt1-Cre models in the labeling of NC cells in early mouse embryos with a focus on cranial NC. We found that R26-lacZ Cre reporter responded to Cre activity more reliably than CAAG-lacZ Cre reporter during early embryogenesis. Cre immunosignals in P0-Cre and reporter (lacZ and RFP) activity in P0-Cre/R26-lacZ and P0-Cre/R26-RFP embryos was detected in the cranial NC and notochord regions in E8.0-9.5 (4-19 somites) embryos. P0-Cre transgene expression was observed in migrating NC cells and was more extensive in the forebrain and hindbrain but not apparent in the midbrain. Differences in the Cre distribution patterns of P0-Cre and Wnt1-Cre were profound in the midbrain and hindbrain regions, that is, extensive in the midbrain of Wnt1-Cre and in the hindbrain of P0-Cre embryos. The difference between P0-Cre and Wnt1-Cre in labeling cranial NC may provide a better explanation of the differential distributions of their NC derivatives and of the phenotypes caused by Cre-driven genetic modifications. © 2017 Wiley Periodicals, Inc.
Objective and subjective measures of exercise intensity during thermo-neutral and hot yoga.
Boyd, Corinne N; Lannan, Stephanie M; Zuhl, Micah N; Mora-Rodriguez, Ricardo; Nelson, Rachael K
2018-04-01
While hot yoga has gained enormous popularity in recent years, owing in part to increased environmental challenge associated with exercise in the heat, it is not clear whether hot yoga is more vigorous than thermo-neutral yoga. Therefore, the aim of this study was to determine objective and subjective measures of exercise intensity during constant intensity yoga in a hot and thermo-neutral environment. Using a randomized, crossover design, 14 participants completed 2 identical ∼20-min yoga sessions in a hot (35.3 ± 0.8 °C; humidity: 20.5% ± 1.4%) and thermo-neutral (22.1 ± 0.2 °C; humidity: 27.8% ± 1.6%) environment. Oxygen consumption and heart rate (HR) were recorded as objective measures (percentage of maximal oxygen consumption and percentage of maximal HR (%HRmax)) and rating of perceived exertion (RPE) was recorded as a subjective measure of exercise intensity. There was no difference in exercise intensity based on percentage of maximal oxygen consumption during hot versus thermo-neutral yoga (30.9% ± 2.3% vs. 30.5% ± 1.8%, p = 0.68). However, exercise intensity was significantly higher during hot versus thermo-neutral yoga based on %HRmax (67.0% ± 2.3% vs. 60.8% ± 1.9%, p = 0.01) and RPE (12 ± 1 vs. 11 ± 1, p = 0.04). According to established exercise intensities, hot yoga was classified as light-intensity exercise based on percentage of maximal oxygen consumption but moderate-intensity exercise based on %HRmax and RPE while thermo-neutral yoga was classified as light-intensity exercise based on percentage of maximal oxygen uptake, %HRmax, and RPE. Despite the added hemodynamic stress and perception that yoga is more strenuous in a hot environment, we observed similar oxygen consumption during hot versus thermo-neutral yoga, classifying both exercise modalities as light-intensity exercise.
Thermo-hydrodynamic lubrication in hydrodynamic bearings
Bonneau, Dominique; Souchet, Dominique
2014-01-01
This Series provides the necessary elements to the development and validation of numerical prediction models for hydrodynamic bearings. This book describes the thermo-hydrodynamic and the thermo-elasto-hydrodynamic lubrication. The algorithms are methodically detailed and each section is thoroughly illustrated.
Thermo-elastic optical coherence tomography
Wang, Tianshi; Pfeiffer, Tom; Wu, Min; Wieser, Wolfgang; Amenta, Gaetano; Draxinger, Wolfgang; van der Steen, A.F.W.; Huber, Robert; Van Soest, Gijs
2017-01-01
The absorption of nanosecond laser pulses induces rapid thermo-elastic deformation in tissue. A sub-micrometer scale displacement occurs within a few microseconds after the pulse arrival. In this Letter, we investigate the laser-induced thermo-elastic deformation using a 1.5 MHz phase-sensitive
Energy Technology Data Exchange (ETDEWEB)
Silva, Davi J.M.; Nunes, Carlos E.A.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: ceanunes@yahoo.com.br, E-mail: rcbarros@pq.cnpq.br [Secretaria Municipal de Educacao de Itaborai, RJ (Brazil); Universidade Estacio de Sa (UNESA), Rio de Janeiro, RJ (Brazil); Universidade do Estado do Rio de Janeiro (UERJ), Novra Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional
2017-11-01
Discussed here is the accuracy of approximate albedo boundary conditions for energy multigroup discrete ordinates (S{sub N}) eigenvalue problems in two-dimensional rectangular geometry for criticality calculations in neutron fission reacting systems, such as nuclear reactors. The multigroup (S{sub N}) albedo matrix substitutes approximately the non-multiplying media around the core, e.g., baffle and reflector, as we neglect the transverse leakage terms within these non-multiplying regions. Numerical results to a typical model problem are given to illustrate the accuracy versus the computer running time. (author)
International Nuclear Information System (INIS)
Huang, Mi; Yi, Ce; Manalo, Kevin L.; Sjoden, Glenn E.
2011-01-01
Multigroup optimization is performed on a neutron detector assembly to examine the validity of transport response in forward and adjoint modes. For SN transport simulations, we discuss the multigroup collapse of an 80 group library to 40, 30, and 16 groups, constructed from using the 3-D parallel PENTRAN and macroscopic cross section collapsing with YGROUP contribution weighting. The difference in using P_1 and P_3 Legendre order in scattering cross sections is investigated; also, associated forward and adjoint transport responses are calculated. We conclude that for the block analyzed, a 30 group cross section optimizes both computation time and accuracy relative to the 80 group transport calculations. (author)
Multi-level methods for solving multigroup transport eigenvalue problems in 1D slab geometry
International Nuclear Information System (INIS)
Anistratov, D. Y.; Gol'din, V. Y.
2009-01-01
A methodology for solving eigenvalue problems for the multigroup neutron transport equation in 1D slab geometry is presented. In this paper we formulate and compare different variants of nonlinear multi-level iteration methods. They are defined by means of multigroup and effective one-group low-order quasi diffusion (LOQD) equations. We analyze the effects of utilization of the effective one-group LOQD problem for estimating the eigenvalue. We present numerical results to demonstrate the performance of the iteration algorithms in different types of reactor-physics problems. (authors)
Multi-level nonlinear diffusion acceleration method for multigroup transport k-Eigenvalue problems
International Nuclear Information System (INIS)
Anistratov, Dmitriy Y.
2011-01-01
The nonlinear diffusion acceleration (NDA) method is an efficient and flexible transport iterative scheme for solving reactor-physics problems. This paper presents a fast iterative algorithm for solving multigroup neutron transport eigenvalue problems in 1D slab geometry. The proposed method is defined by a multi-level system of equations that includes multigroup and effective one-group low-order NDA equations. The Eigenvalue is evaluated in the exact projected solution space of smallest dimensionality, namely, by solving the effective one- group eigenvalue transport problem. Numerical results that illustrate performance of the new algorithm are demonstrated. (author)
Energy Technology Data Exchange (ETDEWEB)
Ghrayeb, S. Z. [Dept. of Mechanical and Nuclear Engineering, Pennsylvania State Univ., 230 Reber Building, Univ. Park, PA 16802 (United States); Ouisloumen, M. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Ougouag, A. M. [Idaho National Laboratory, MS-3860, PO Box 1625, Idaho Falls, ID 83415 (United States); Ivanov, K. N.
2012-07-01
A multi-group formulation for the exact neutron elastic scattering kernel is developed. This formulation is intended for implementation into a lattice physics code. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. A computer program has been written to test the formulation for various nuclides. Results of the multi-group code have been verified against the correct analytic scattering kernel. In both cases neutrons were started at various energies and temperatures and the corresponding scattering kernels were tallied. (authors)
Multigroup neutron transport equation in the diffusion and P{sub 1} approximation
Energy Technology Data Exchange (ETDEWEB)
Obradovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)
1970-07-01
Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P{sub 1} approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P{sub 1} approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)
Thermo-plasmonics of Irradiated Metallic Nanostructures
DEFF Research Database (Denmark)
Ma, Haiyan
Thermo-plasmonics is an emerging field in photonics which aims at harnessing the kinetic energy of light to generate nanoscopic sources of heat. Localized surface plasmons (LSP) supported by metallic nanostructures greatly enhance the interactions of light with the structure. By engineering...... delivery, nano-surgeries and thermo-transportations. Apart from generating well-controlled temperature increase in functional thermo-plasmonic devices, thermo-plasmonics can also be used in understanding complex phenomena in thermodynamics by creating drastic temperature gradients which are not accessible...... using conventional techniques. In this thesis, we present novel experimental and numerical tools to characterize thermo-plasmonic devices in a biologically relevant environment, and explore the thermodiffusion properties and measure thermophoretic forces for particles in temperature gradients ranging...
Han, Yan-Hong; Xiang, Hai-Ying; Wang, Qian; Li, Yuan-Yuan; Wu, Wen-Qi; Han, Cheng-Gui; Li, Da-Wei; Yu, Jia-Lin
2010-10-10
Melon aphid-borne yellows virus (MABYV) is a newly identified polerovirus occurring in China. Here, we demonstrate that the MABYV encoded P0 (P0(MA)) protein is a strong suppressor of post-transcriptional gene silencing (PTGS) with activity comparable to tobacco etch virus (TEV) HC-Pro. In addition we have shown that the LP F-box motif present at the N-terminus of P0(MA) is required for suppressor activity. Detailed mutational analyses on P0(MA) revealed that changing the conserved Trp 212 with non-ring structured amino acids altered silencing suppressor functions. Ala substitutions at positions 12 and 211 for Phe had no effect on P0 suppression-activity, whereas Arg and Glu substitutions had greatly decreased suppressor activity. Furthermore, substitutions targeting Phe at position 30 also resulted in reduced P0 suppression-activity. Altogether, these results suggest that ring structured Trp/Phe residues in P0 have important roles in suppressor activity. Copyright © 2010 Elsevier Inc. All rights reserved.
Energy Technology Data Exchange (ETDEWEB)
Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G., E-mail: ansar.calloo@cea.fr, E-mail: jean-francois.vidal@cea.fr, E-mail: romain.le-tellier@cea.fr, E-mail: gerald.rimpault@cea.fr [CEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance (France)
2011-07-01
This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S{sub n} method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)
Finally! A valid test of configural invariance using permutation in multigroup CFA
Jorgensen, T.D.; Kite, B.A.; Chen, P.-Y.; Short, S.D.; van der Ark, L.A.; Wiberg, M.; Culpepper, S.A.; Douglas, J.A.; Wang, W.-C.
2017-01-01
In multigroup factor analysis, configural measurement invariance is accepted as tenable when researchers either (a) fail to reject the null hypothesis of exact fit using a χ2 test or (b) conclude that a model fits approximately well enough, according to one or more alternative fit indices (AFIs).
AMZ, multigroup constant library for EXPANDA code, generated by NJOY code from ENDF/B-IV
International Nuclear Information System (INIS)
Chalhoub, E.S.; Moraes, Marisa de
1985-01-01
It is described a library of multigroup constants with 70 energy groups and 37 isotopes to fast reactor calculation. The cross sections, scattering matrices and self-shielding factors were generated by NJOY code and RGENDF interface program, from ENDF/B-IV'S evaluated data. The library is edited in adequated format to be used by EXPANDA code. (M.C.K.) [pt
MCFT: a program for calculating fast and thermal neutron multigroup constants
International Nuclear Information System (INIS)
Yang Shunhai; Sang Xinzeng
1993-01-01
MCFT is a program for calculating the fast and thermal neutron multigroup constants, which is redesigned from some codes for generation of thermal neutron multigroup constants and for fast neutron multigroup constants adapted on CYBER 825 computer. It uses indifferently as basic input with the evaluated nuclear data contained in the ENDF/B (US), KEDAK (Germany) and UK (United Kingdom) libraries. The code includes a section devoted to the generation of resonant Doppler broadened cross section in the framework of single-or multi-level Breit-Wigner formalism. The program can compute the thermal neutron scattering law S (α, β, T) as the input data in tabular, free gas or diffusion motion form. It can treat up to 200 energy groups and Legendre moments up to P 5 . The output consists of various reaction multigroup constants in all neutron energy range desired in the nuclear reactor design and calculation. Three options in input file can be used by the user. The output format is arbitrary and defined by user with a minimum of program modification. The program includes about 15,000 cards and 184 subroutines. FORTRAN 5 computer language is used. The operation system is under NOS 2 on computer CYBER 825
MC^{2}-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis
Energy Technology Data Exchange (ETDEWEB)
Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)
2013-11-08
The MC^{2}-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC^{2}-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.
International Nuclear Information System (INIS)
Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G.
2011-01-01
This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S_n method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)
The problem of resonance self-shielding effect in neutron multigroup calculations
International Nuclear Information System (INIS)
Wang Qingming; Huang Jinghua
1991-01-01
It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations
Energy Technology Data Exchange (ETDEWEB)
Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)
2014-05-01
The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.
International Nuclear Information System (INIS)
Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh
2014-01-01
The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects
International Nuclear Information System (INIS)
Roussin, R.W.; Drischler, J.D.; Marable, J.H.
1980-01-01
In recent years multigroup sensitivity profiles and covariance matrices have been added to the Radiation Shielding Information Center's Data Library Collection (DLC). Sensitivity profiles are available in a single package. DLC-45/SENPRO, and covariance matrices are found in two packages, DLC-44/COVERX and DLC-77/COVERV. The contents of these packages are described and their availability is discussed
International Nuclear Information System (INIS)
Ozgener, B.
1998-01-01
A boundary integral equation (BIE) is developed for the application of the boundary element method to the multigroup neutron diffusion equations. The developed BIE contains no explicit scattering term; the scattering effects are taken into account by redefining the unknowns. Boundary elements of the linear and constant variety are utilised for validation of the developed boundary integral formulation
Thermo-sensitive intelligent track membrane
International Nuclear Information System (INIS)
Pang Deling; Ren Lihua; Qian Zhilin; Huang Gang; Zhang Jinhua
1999-01-01
Using N-isopropylacryl-amide (NIP AAm) thermo-sensitive function material as monomer and nuclear track microporous membrane (NTMM) as baseline material, a thermo-sensitive intelligent track membrane (TsITM) has been prepared by the over-oxidization and pre-irradiation grafting techniques. The TsITM can be used to make a micro-switch controlled by temperature and to adjust particle screening and osmosis. To obtain sub-micron responsive grafted track pores only a very thin thermo-sensitive layer is needed. The TsITM pores are capable of swelling and shrinking rapidly and respond more sensitively to temperature
International Nuclear Information System (INIS)
Nguyen Quoc Thang
2004-08-01
We show the validity of te Corestriction Principle for non-abelian cohomology of connected reductive groups over local ad global fields of characteristic p > 0 , by extending some results by Kneser and Douai. (author)
DEFF Research Database (Denmark)
Rosberg, Mette Romer; Alvarez, Susana; Krarup, Christian
2013-01-01
Mice with a heterozygous knock-out of the myelin protein P0 gene (P0+/-) develop a neuropathy similar to human Charcot-Marie-Tooth disease. They are indistinguishable from wild-types (WT) at birth and develop a slowly progressing demyelinating neuropathy. The aim of this study was to investigate...... whether the regeneration capacity of early symptomatic P0+/- is impaired as compared to age matched WT. Right sciatic nerves were lesioned at the thigh in 7-8 months old mice. Tibial motor axons at ankle were investigated by conventional motor conduction studies and axon excitability studies using...... threshold tracking. To evaluate regeneration we monitored the recovery of motor function after crush, and then compared the fiber distribution by histology. The overall motor performance was investigated using Rotor-Rod. P0+/- had reduced compound motor action potential amplitudes and thinner myelinated...
International Nuclear Information System (INIS)
Kelsey IV, Charles T.; Prinja, Anil K.
2011-01-01
We evaluate the Monte Carlo calculation efficiency for multigroup transport relative to continuous energy transport using the MCNPX code system to evaluate secondary neutron doses from a proton beam. We consider both fully forward simulation and application of a midway forward adjoint coupling method to the problem. Previously we developed tools for building coupled multigroup proton/neutron cross section libraries and showed consistent results for continuous energy and multigroup proton/neutron transport calculations. We observed that forward multigroup transport could be more efficient than continuous energy. Here we quantify solution efficiency differences for a secondary radiation dose problem characteristic of proton beam therapy problems. We begin by comparing figures of merit for forward multigroup and continuous energy MCNPX transport and find that multigroup is 30 times more efficient. Next we evaluate efficiency gains for coupling out-of-beam adjoint solutions with forward in-beam solutions. We use a variation of a midway forward-adjoint coupling method developed by others for neutral particle transport. Our implementation makes use of the surface source feature in MCNPX and we use spherical harmonic expansions for coupling in angle rather than solid angle binning. The adjoint out-of-beam transport for organs of concern in a phantom or patient can be coupled with numerous forward, continuous energy or multigroup, in-beam perturbations of a therapy beam line configuration. Out-of-beam dose solutions are provided without repeating out-of-beam transport. (author)
AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B
Energy Technology Data Exchange (ETDEWEB)
Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.
1976-03-01
AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)
International Nuclear Information System (INIS)
Mi Aijun; Li Junjie
2010-01-01
In this paper the multi-group libraries were constructed by processing ENDF/B-VII neutron incident files into multi-group structure, and the application of the multi-group libraries in the pressurized-water reactor(PWR) design was studied. The construction of the multi-group library is realized by using the NJOY nuclear data processing system. The code can process the neutron cross section files form ENDF format to MATXS format which was required in SN code. Two dimension transport theory code of discrete ordinates DORT was used to verify the multi-group libraries and the method of the construction by comparing calculations for some representative benchmarks. We made the PWR shielding calculation by using the multi-group libraries and studied the influence of the parameters involved during the construction of the libraries such as group structure, temperatures and weight functions on the shielding design of the PWR. This work is the preparation for the construction of the multi-group library which will be used in PWR shielding design in engineering. (authors)
Energy Technology Data Exchange (ETDEWEB)
Fusaro, Adriana F. [University of Sydney, NSW 2006 (Australia); CSIRO Plant Industry, Canberra, P.O. Box 1600, ACT 2601 (Australia); Correa, Regis L. [CSIRO Plant Industry, Canberra, P.O. Box 1600, ACT 2601 (Australia); Depto. de Virologia, IMPPG, UFRJ, 21941-902 (Brazil); Nakasugi, Kenlee; Jackson, Craig [University of Sydney, NSW 2006 (Australia); Kawchuk, Lawrence [Research Centre, Agriculture and Agri-Food Canada, Lethbridge, AB T1J4B1 (Canada); Vaslin, Maite F.S. [Depto. de Virologia, IMPPG, UFRJ, 21941-902 (Brazil); Waterhouse, Peter M., E-mail: peter.waterhouse@sydney.edu.au [University of Sydney, NSW 2006 (Australia); CSIRO Plant Industry, Canberra, P.O. Box 1600, ACT 2601 (Australia)
2012-05-10
The P0 protein of poleroviruses and P1 protein of sobemoviruses suppress the plant's RNA silencing machinery. Here we identified a silencing suppressor protein (SSP), P0{sup PE}, in the Enamovirus Pea enation mosaic virus-1 (PEMV-1) and showed that it and the P0s of poleroviruses Potato leaf roll virus and Cereal yellow dwarf virus have strong local and systemic SSP activity, while the P1 of Sobemovirus Southern bean mosaic virus supresses systemic silencing. The nuclear localized P0{sup PE} has no discernable sequence conservation with known SSPs, but proved to be a strong suppressor of local silencing and a moderate suppressor of systemic silencing. Like the P0s from poleroviruses, P0{sup PE} destabilizes AGO1 and this action is mediated by an F-box-like domain. Therefore, despite the lack of any sequence similarity, the poleroviral and enamoviral SSPs have a conserved mode of action upon the RNA silencing machinery.
Delfosse, Verónica C; Agrofoglio, Yamila C; Casse, María F; Kresic, Iván Bonacic; Hopp, H Esteban; Ziegler-Graff, Véronique; Distéfano, Ana J
2014-02-13
Plants employ RNA silencing as a natural defense mechanism against viruses. As a counter-defense, viruses encode silencing suppressor proteins (SSPs) that suppress RNA silencing. Most, but not all, the P0 proteins encoded by poleroviruses have been identified as SSP. In this study, we demonstrated that cotton leafroll dwarf virus (CLRDV, genus Polerovirus) P0 protein suppressed local silencing that was induced by sense or inverted repeat transgenes in Agrobacterium co-infiltration assay in Nicotiana benthamiana plants. A CLRDV full-length infectious cDNA clone that is able to infect N. benthamiana through Agrobacterium-mediated inoculation also inhibited local silencing in co-infiltration assays, suggesting that the P0 protein exhibits similar RNA silencing suppression activity when expressed from the full-length viral genome. On the other hand, the P0 protein did not efficiently inhibit the spread of systemic silencing signals. Moreover, Northern blotting indicated that the P0 protein inhibits the generation of secondary but not primary small interfering RNAs. The study of CLRDV P0 suppression activity may contribute to understanding the molecular mechanisms involved in the induction of cotton blue disease by CLRDV infection. Copyright © 2013 Elsevier B.V. All rights reserved.
Thermo-elastic optical coherence tomography.
Wang, Tianshi; Pfeiffer, Tom; Wu, Min; Wieser, Wolfgang; Amenta, Gaetano; Draxinger, Wolfgang; van der Steen, Antonius F W; Huber, Robert; Soest, Gijs van
2017-09-01
The absorption of nanosecond laser pulses induces rapid thermo-elastic deformation in tissue. A sub-micrometer scale displacement occurs within a few microseconds after the pulse arrival. In this Letter, we investigate the laser-induced thermo-elastic deformation using a 1.5 MHz phase-sensitive optical coherence tomography (OCT) system. A displacement image can be reconstructed, which enables a new modality of phase-sensitive OCT, called thermo-elastic OCT. An analysis of the results shows that the optical absorption is a dominating factor for the displacement. Thermo-elastic OCT is capable of visualizing inclusions that do not appear on the structural OCT image, providing additional tissue type information.
International Nuclear Information System (INIS)
Anton, V.
1979-05-01
A new formulation of multigroup cross section collapsing based on the conservation of point or zone value of hamiltonian is presented. This attempt is proper to optimization problems solved by means of maximum principle of Pontryagin. (author)
International Nuclear Information System (INIS)
Anaf, J.; Chalhoub, E.S.
1987-11-01
A system, composed by the computer programs COMPAR and its interfaces, developed for comparing multigroup cross sections calculated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS, is presented. (author)
International Nuclear Information System (INIS)
Anaf, J.; Chalhoub, E.S.
1988-02-01
A system consisting of the COMPAR computer program and its interfaces which was developed for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS is presented. (author). 13 refs
International Nuclear Information System (INIS)
Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang
2010-01-01
A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.
Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis
International Nuclear Information System (INIS)
Arien, B.; Daniels, J.
1986-12-01
CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)
COMPAR, NJOY, GROUPIE, FLANGE-2, ETOG-3, XLACS Multigroup Cross-Sections General Comparison
International Nuclear Information System (INIS)
Anaf, Jaime; Chalhoub, E.S.
1990-01-01
1 - Description of program or function: A system for comparing multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS. This system comprises the COMPAR program and interface (auxiliary) programs developed for each of the programs under consideration. These are REDCOMP for GROUPIE, FLACOMP for FLANGE-II, ETOCOMP for ETOG-3 and XLACOMP for XLACS. For the NJOY program there is RGENDF, a program developed apart from this system. It is a modular system in which the inclusion of new multigroup cross section generating program requires no more than the development of a new interface module. 2 - Method of solution: Refer to comments in main routine. 3 - Restrictions on the complexity of the problem: Refer to comments in main routine
International Nuclear Information System (INIS)
Greene, N.M.; Arwood, J.W.; Wright, R.Q.; Parks, C.V.
1994-08-01
The 238-group LAW Library is a new multigroup neutron cross-section library based on ENDF/B-V data, with five sets of data taken from ENDF/B-VI ( 14 N 7 , 15 N 7 , 16 O 8 , 154Eu 63 , and 155 Eu 63 ). These five nuclides are included because the new evaluations are thought to be superior to those in Version 5. The LAW Library contains data for over 300 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, although it has equal utility in any study requiring multigroup neutron cross sections
Second order time evolution of the multigroup diffusion and P1 equations for radiation transport
International Nuclear Information System (INIS)
Olson, Gordon L.
2011-01-01
Highlights: → An existing multigroup transport algorithm is extended to be second-order in time. → A new algorithm is presented that does not require a grey acceleration solution. → The two algorithms are tested with 2D, multi-material problems. → The two algorithms have comparable computational requirements. - Abstract: An existing solution method for solving the multigroup radiation equations, linear multifrequency-grey acceleration, is here extended to be second order in time. This method works for simple diffusion and for flux-limited diffusion, with or without material conduction. A new method is developed that does not require the solution of an averaged grey transport equation. It is effective solving both the diffusion and P 1 forms of the transport equation. Two dimensional, multi-material test problems are used to compare the solution methods.
Correction of multigroup cross sections for resolved resonance interference in mixed absorbers
International Nuclear Information System (INIS)
Williams, M.L.
1982-07-01
The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed
MC2-2: a code to calculate fast neutron spectra and multigroup cross sections
International Nuclear Information System (INIS)
Henryson, H. II; Toppel, B.J.; Stenberg, C.G.
1976-06-01
MC 2 -2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC 2 -2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC 2 -2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC 2 -2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC 2 -2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers
On efficiently computing multigroup multi-layer neutron reflection and transmission conditions
International Nuclear Information System (INIS)
Abreu, Marcos P. de
2007-01-01
In this article, we present an algorithm for efficient computation of multigroup discrete ordinates neutron reflection and transmission conditions, which replace a multi-layered boundary region in neutron multiplication eigenvalue computations with no spatial truncation error. In contrast to the independent layer-by-layer algorithm considered thus far in our computations, the algorithm here is based on an inductive approach developed by the present author for deriving neutron reflection and transmission conditions for a nonactive boundary region with an arbitrary number of arbitrarily thick layers. With this new algorithm, we were able to increase significantly the computational efficiency of our spectral diamond-spectral Green's function method for solving multigroup neutron multiplication eigenvalue problems with multi-layered boundary regions. We provide comparative results for a two-group reactor core model to illustrate the increased efficiency of our spectral method, and we conclude this article with a number of general remarks. (author)
Verification of KARMA GEOM/TRPT Module with Given Multi-group Cross Sections
International Nuclear Information System (INIS)
Koo, Bon Seung; Hong, Ser Gi; Song, Jae Seung
2009-01-01
KAERI has developed a two-dimensional multigroup transport theory code KARMA (Kernel Analyzer by Ray-tracing Method for Fuel Assembly). KARMA uses CMFD (Coarse Mesh Finite Difference) accelerated MOC (Method of Characteristics) method for burnup calculation on a single fuel pin, a fuel assembly and a core consisting of rectangular array of fuel pins. KARMA code intends to be employed as a nuclear design tool for the Korean commercial pressurizer water reactor. Prior to the application to actual assembly designs, the code has to be approved by regularity agency. Therefore, it is essential that the reliability of KARMA code should be sufficiently evaluated against well-defined benchmark problems. In this paper, verification of GEOM/TRPT modules of KARMA was performed to confirm a reliability of the KARMA transport solution via comparisons with Monte Carlo calculations by using a consistent set of multi-group macroscopic cross-sections
CLUB - a multigroup integral transport theory code for lattice calculations of PHWR cells
International Nuclear Information System (INIS)
Krishnani, P.D.
1992-01-01
The computer code CLUB has been developed to calculate lattice parameters as a function of burnup for a pressurised heavy water reactor (PHWR) lattice cell containing fuel in the form of cluster. It solves the multigroup integral transport equation by the method based on combination of small scale collision probability (CP) method and large scale interface current technique. The calculations are performed by using WIMS 69 group cross section library or its condensed versions of 27 or 28 group libraries. It can also compute Keff from the given geometrical buckling in the input using multigroup diffusion theory in fundamental mode. The first order differential burnup equations can be solved by either Trapezoidal rule or Runge-Kutta method. (author). 17 refs., 2 figs
International Nuclear Information System (INIS)
Honeck, H.C.
1984-01-01
1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups
TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticality studies
International Nuclear Information System (INIS)
Ermumcu, G.; Gonnord, J.; Nimal, J.C.
1980-01-01
TRIMARAN is developed for safety analysis of nuclear components containing fissionable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method, in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies
A code system to generate multigroup cross-sections using basic data
International Nuclear Information System (INIS)
Garg, S.B.; Kumar, Ashok
1978-01-01
For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)
Young Adults’ Attitude Towards Advertising: a multi-group analysis by ethnicity
Hiram Ting; Ernest Cyril de Run; Ramayah Thurasamy
2015-01-01
Objective – This study aims to investigate the attitude of Malaysian young adults towards advertising. How this segment responds to advertising, and how ethnic/cultural differences moderate are assessed.Design/methodology/approach – A quantitative questionnaire is used to collect data at two universities. Purposive sampling technique is adopted to ensure the sample represents the actual population. Structural equation modelling (SEM) and multi-group analysis (MGA) are utilized in analysis.Fin...
International Nuclear Information System (INIS)
Takeshi, Y.; Keisuke, K.
1983-01-01
The multigroup neutron diffusion equation for two-dimensional triangular geometry is solved by the finite Fourier transformation method. Using the zero-th-order equation of the integral equation derived by this method, simple algebraic expressions for the flux are derived and solved by the alternating direction implicit method. In sample calculations for a benchmark problem of a fast breeder reactor, it is shown that the present method gives good results with fewer mesh points than the usual finite difference method
Survey of computer codes which produce multigroup data from ENDF/B-IV
International Nuclear Information System (INIS)
Greene, N.M.
1975-01-01
The features of three code systems that produce multigroup neutron data are contrasted. This includes the ETOE-2/MC 2 -2/SDX, MINX/SPHINX and AMPX code packages. These systems all contain a fairly extensive set of processing capabilities with the current evaluated nuclear data files--ENDF/B. They were designed with different goals and applications in mind. This paper discusses some of their differences and the implications for particular situations
International Nuclear Information System (INIS)
Rubin, I.E.; Dneprovskaya, N.M.
2005-01-01
A technique for calculation of reactor lattices by means of the transmission probabilities with taking into account the scattering anisotropy is generalized for the multigroup case. The errors of the calculated multiplication coefficients and energy release distributions do noe exceed practically the errors, of these values, obtained by the Monte Carlo method. The proposed method is most effective when determining the small difference effects [ru
TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies
International Nuclear Information System (INIS)
Ermuncu, G.; Gonnord, J.; Nimal, J.C.
1980-04-01
TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies
International Nuclear Information System (INIS)
LaBauve, R.J.; Muir, D.W.
1978-01-01
A library of 30-group multigroup covariance data was prepared from preliminary ENDF/B-V data with the NJOY code. Data for Fe, Cr, Ni, 10 B, C, Cu, H, and Pb are included in this library. Reactions include total cross sections, elastic and inelastic scattering cross sections, and the most important absorption cross sections. Typical data from the file are shown. 3 tables
International Nuclear Information System (INIS)
Cullen, D.E.; Perkins, S.T.
1977-01-01
Multi-group averaged reaction rates and transfer matrices were calculated for charged particle induced elastic nuclear (plus interference) scattering. Results are presented using a ten group structure for all twenty-five permutations of projectile and target for the following charged particles: p, d, t, 3 He and alpha. Transfer matrices are presented in a simplified form for both incident projectile and the knock-ons; these matrices explicitly conserve energy
F-box-like domain in the polerovirus protein P0 is required for silencing suppressor function
Pazhouhandeh, Maghsoud; Dieterle, Monika; Marrocco, Katia; Lechner, Esther; Berry, Bassam; Brault, Véronique; Hemmer, Odile; Kretsch, Thomas; Richards, Kenneth E.; Genschik, Pascal; Ziegler-Graff, Véronique
2006-01-01
Plants employ small RNA-mediated posttranscriptional gene silencing as a virus defense mechanism. In response, plant viruses encode proteins that can suppress RNA silencing, but the mode of action of most such proteins is poorly understood. Here, we show that the silencing suppressor protein P0 of two Arabidopsis-infecting poleroviruses interacts by means of a conserved minimal F-box motif with Arabidopsis thaliana orthologs of S-phase kinase-related protein 1 (SKP1), a component of the SCF family of ubiquitin E3 ligases. Point mutations in the F-box-like motif abolished the P0–SKP1 ortholog interaction, diminished virus pathogenicity, and inhibited the silencing suppressor activity of P0. Knockdown of expression of a SKP1 ortholog in Nicotiana benthamiana rendered the plants resistant to polerovirus infection. Together, the results support a model in which P0 acts as an F-box protein that targets an essential component of the host posttranscriptional gene silencing machinery. PMID:16446454
Energy Technology Data Exchange (ETDEWEB)
Bayard, J P; Guillou, A; Lago, B; Bureau du Colombier, M J; Guillou, G; Vasseur, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1965-02-01
This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, R{theta}. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [French] Ce rapport decrit les specifications du programme ALCI. Ce programme resout le systeme d'equations aux differences approchant le probleme homogene de la diffusion neutronique multigroupe, a deux dimensions d'espace, dans les trois geometries XY, RZ, R{theta}. Il permet des calculs de criticalite geometrique et de composition et calcule sur demande le probleme adjoint. Le nombre maximum de points traites est de 6000. Le nombre maximum de groupes permis est de 12. Les iterations interieure sont traitees par la methode des directions alternees. Les iterations exterieures sont accelerees par la methode d'extrapolation de Tchebychev. (auteurs)
A multilevel in space and energy solver for multigroup diffusion eigenvalue problems
Directory of Open Access Journals (Sweden)
Ben C. Yee
2017-09-01
Full Text Available In this paper, we present a new multilevel in space and energy diffusion (MSED method for solving multigroup diffusion eigenvalue problems. The MSED method can be described as a PI scheme with three additional features: (1 a grey (one-group diffusion equation used to efficiently converge the fission source and eigenvalue, (2 a space-dependent Wielandt shift technique used to reduce the number of PIs required, and (3 a multigrid-in-space linear solver for the linear solves required by each PI step. In MSED, the convergence of the solution of the multigroup diffusion eigenvalue problem is accelerated by performing work on lower-order equations with only one group and/or coarser spatial grids. Results from several Fourier analyses and a one-dimensional test code are provided to verify the efficiency of the MSED method and to justify the incorporation of the grey diffusion equation and the multigrid linear solver. These results highlight the potential efficiency of the MSED method as a solver for multidimensional multigroup diffusion eigenvalue problems, and they serve as a proof of principle for future work. Our ultimate goal is to implement the MSED method as an efficient solver for the two-dimensional/three-dimensional coarse mesh finite difference diffusion system in the Michigan parallel characteristics transport code. The work in this paper represents a necessary step towards that goal.
Depletion Calculations for MTR Core Using MCNPX and Multi-Group Nodal Diffusion Methods
International Nuclear Information System (INIS)
Jaradata, Mustafa K.; Park, Chang Je; Lee, Byungchul
2013-01-01
In order to maintain a self-sustaining steady-state chain reaction, more fuel than is necessary in order to maintain a steady state chain reaction must be loaded. The introduction of this excess fuel increases the net multiplication capability of the system. In this paper MCNPX and multi-group nodal diffusion theory will be used for depletion calculations for MTR core. The eigenvalue and power distribution in the core will be compared for different burnup. Multi-group nodal diffusion theory with combination of NEWT-TRITON system was used to perform depletion calculations for 3Χ3 MTR core. 2G and 6G approximations were used and compared with MCNPX results for 2G approximation the maximum difference from MCNPX was 40 mk and for 6G approximation was 6 mk which is comparable to the MCNPX results. The calculated power using nodal code was almost the same MCNPX results. Finally the results of the multi-group nodal theory were acceptable and comparable to the calculated using MCNPX
International Nuclear Information System (INIS)
Abreu, M.P.; Filho, H.A.; Barros, R.C.
1993-01-01
The authors describe a new nodal method for multigroup slab-geometry discrete ordinates S N eigenvalue problems that is completely free from all spatial truncation errors. The unknowns in the method are the node-edge angular fluxes, the node-average angular fluxes, and the effective multiplication factor k eff . The numerical values obtained for these quantities are exactly those of the dominant analytic solution of the S N eigenvalue problem apart from finite arithmetic considerations. This method is based on the use of the standard balance equation and two nonstandard auxiliary equations. In the nonmultiplying regions, e.g., the reflector, we use the multigroup spectral Green's function (SGF) auxiliary equations. In the fuel regions, we use the multigroup spectral diamond (SD) auxiliary equations. The SD auxiliary equation is an extension of the conventional auxiliary equation used in the diamond difference (DD) method. This hybrid characteristic of the SD-SGF method improves both the numerical stability and the convergence rate
An energy recondensation method using the discrete generalized multigroup energy expansion theory
International Nuclear Information System (INIS)
Zhu Lei; Forget, Benoit
2011-01-01
Highlights: → Discrete-generalized multigroup method was implemented as a recondensation scheme. → Coarse group cross-sections were recondensed from core-level solution. → Neighboring effect of reflector and MOX bundle was improved. → Methodology was shown to be fully consistent when a flat angular flux approximation is used. - Abstract: In this paper, the discrete generalized multigroup (DGM) method was used to recondense the coarse group cross-sections using the core level solution, thus providing a correction for neighboring effect found at the core level. This approach was tested using a discrete ordinates implementation in both 1-D and 2-D. Results indicate that 2 or 3 iterations can substantially improve the flux and fission density errors associated with strong interfacial spectral changes as found in the presence of strong absorbers, reflector of mixed-oxide fuel. The methodology is also proven to be fully consistent with the multigroup methodology as long as a flat-flux approximation is used spatially.
International Nuclear Information System (INIS)
Gastaldi, B.
1986-07-01
This study intends to improve then to check on integral experiments, the calculation of the main neutronic parameters in light water moderated lattices: Uranium 238 capture and consequently Plutonium 239 build-up, multiplication factor, temperature coefficient. The first part of this work concerns the resonant reaction rate calculation method implemented in the APOLLO code, the so-called LIVOLANT and JEANPIERRE formalism. The errors introduced by the corresponding assumptions are quantified and we propose substitution methods which avoid large biases and supply satisfactory results. The second part is dedicated to the cross-section evaluation of uranium major isotopes and to the achievement of APOLLO multigroup cross-sections. This cross-section set takes into considerations on the one hand the recent differential information and the other hand the various integral information obtained in the French Atomic Energy Commission facilities. The nuclear data file (JEF abd ENDF/B5) processing, for multigroup and self-shielded cross-sections achieving enable us to check the new THEMIS computer code. In the last part, the experimental validation of the proposed procedure (accurate formalism mutuel shielding and new multigroup library) is presented. This qualification is based on the reinterpretation of critical experiments performed in the EOLE reactor at Cadarache and spent fuel analysis. The corresponding results demonstrate that our propositions provide improvements on the computation of the PWR neutronic parameters; calculation-experiment discrepancies are now consistent with experimental uncertainty margins. 46 refs; 31 figs; 23 tabl [fr
International Nuclear Information System (INIS)
Mosca, P.
2009-12-01
The deterministic transport codes solve the stationary Boltzmann equation in a discretized energy formalism called multigroup. The transformation of continuous data in a multigroup form is obtained by averaging the highly variable cross sections of the resonant isotopes with the solution of the self-shielding models and the remaining ones with the coarse energy spectrum of the reactor type. So far the error of such an approach could only be evaluated retrospectively. To remedy this, we studied in this thesis a set of methods to control a priori the accuracy and the cost of the multigroup transport computation. The energy mesh optimisation is achieved using a two step process: the creation of a reference mesh and its optimized condensation. In the first stage, by refining locally and globally the energy mesh, we seek, on a fine energy mesh with subgroup self-shielding, a solution equivalent to a reference solver (Monte Carlo or pointwise deterministic solver). In the second step, once fixed the number of groups, depending on the acceptable computational cost, and chosen the most appropriate self-shielding models to the reactor type, we look for the best bounds of the reference mesh minimizing reaction rate errors by the particle swarm optimization algorithm. This new approach allows us to define new meshes for fast reactors as accurate as the currently used ones, but with fewer groups. (author)
Physics of thermo-acoustic sound generation
Daschewski, M.; Boehm, R.; Prager, J.; Kreutzbruck, M.; Harrer, A.
2013-09-01
We present a generalized analytical model of thermo-acoustic sound generation based on the analysis of thermally induced energy density fluctuations and their propagation into the adjacent matter. The model provides exact analytical prediction of the sound pressure generated in fluids and solids; consequently, it can be applied to arbitrary thermal power sources such as thermophones, plasma firings, laser beams, and chemical reactions. Unlike existing approaches, our description also includes acoustic near-field effects and sound-field attenuation. Analytical results are compared with measurements of sound pressures generated by thermo-acoustic transducers in air for frequencies up to 1 MHz. The tested transducers consist of titanium and indium tin oxide coatings on quartz glass and polycarbonate substrates. The model reveals that thermo-acoustic efficiency increases linearly with the supplied thermal power and quadratically with thermal excitation frequency. Comparison of the efficiency of our thermo-acoustic transducers with those of piezoelectric-based airborne ultrasound transducers using impulse excitation showed comparable sound pressure values. The present results show that thermo-acoustic transducers can be applied as broadband, non-resonant, high-performance ultrasound sources.
On nonlinear thermo-electro-elasticity.
Mehnert, Markus; Hossain, Mokarram; Steinmann, Paul
2016-06-01
Electro-active polymers (EAPs) for large actuations are nowadays well-known and promising candidates for producing sensors, actuators and generators. In general, polymeric materials are sensitive to differential temperature histories. During experimental characterizations of EAPs under electro-mechanically coupled loads, it is difficult to maintain constant temperature not only because of an external differential temperature history but also because of the changes in internal temperature caused by the application of high electric loads. In this contribution, a thermo-electro-mechanically coupled constitutive framework is proposed based on the total energy approach. Departing from relevant laws of thermodynamics, thermodynamically consistent constitutive equations are formulated. To demonstrate the performance of the proposed thermo-electro-mechanically coupled framework, a frequently used non-homogeneous boundary-value problem, i.e. the extension and inflation of a cylindrical tube, is solved analytically. The results illustrate the influence of various thermo-electro-mechanical couplings.
THERMOS, district central heating nuclear reactors
International Nuclear Information System (INIS)
Patarin, L.
1981-02-01
In order to expand the penetration of uranium in the national energy balance sheet, the C.E.A. has been studying nuclear reactors for several years now, that are capable of providing heat at favourable economic conditions. In this paper the THERMOS model is introduced. After showing the attraction of direct town heating by nuclear energy, the author describes the THERMOS project, defines the potential market, notably in France, and applies the lay-out study to the Grenoble Nuclear Study Centre site with district communal heating in mind. The economic aspects of the scheme are briefly mentioned [fr
Photon Echoes in the 3P0 ← 3H4 Transition of Pr3+/LaF3
Morsink, Jos B.W.; Wiersma, Douwe A.
1979-01-01
Photon-echo quantum beats observed in the two-pulse and three-pulse photon echo of the 3P0 ← 3H4 transition in Pr3+/LaF3 were used to determine the excited-state spin-hamiltonian. In addition we report on the anomalous stimulated photon echo observed in the same transition which in a magnetic field
International Nuclear Information System (INIS)
Polivanskij, V.P.
1989-01-01
The method to solve two-dimensional equations of neutron transport using 4P 0 -approximation is presented. Previously such approach was efficiently used for the solution of one-dimensional problems. New an attempt is made to apply the approach to solution of two-dimensional problems. Algorithm of the solution is given, as well as results of test neutron-physical calculations. A considerable as compared with diffusion approximation is shown. 11 refs
Epitaxial growth of matched metallic ErP0.6As0.4 layers on GaAs
International Nuclear Information System (INIS)
Guivarc'h, A.; Le Corre, A.; Gaulet, J.; Guenais, B.; Minier, M.; Ropars, G.; Badoz, P.A.; Duboz, J.Y.
1990-01-01
Successful growth of (001)ErP 0.6 As 0.4 single crystal film on (001) GaAs has been demonstrated. The epitaxial metallic layers reproducibly showed lattice mismatch below 5 10 -4 . This is, to the authors' knowledge, the first report of a stable, epitaxial and lattice-matched metal/compound semiconductor heterostructure. The ErP 0.6 As 0.4 /n-GaAs diodes yielded excellent I-V characteristics with an ideality factor of 1.1 and barrier height of 0.88 eV. For a 240 Angstrom- thick film, metallic behavior was observed with resistivities of 25 and 86 μΩcm at 1.5 K and room temperature, respectively. As the other Er compounds ErP, ErAs, ErSb and ErSi 2 , ErP 0.6 As 0.4 presents an abrupt drop in resistivity in the vicinity of the liquid helium temperature, due to a paramagnetic to antiferromagnetic phase transition
Saher, Gesine; Quintes, Susanne; Möbius, Wiebke; Wehr, Michael C; Krämer-Albers, Eva-Maria; Brügger, Britta; Nave, Klaus-Armin
2009-05-13
Rapid impulse conduction requires electrical insulation of axons by myelin, a cholesterol-rich extension of the glial cell membrane with a characteristic composition of proteins and lipids. Mutations in several myelin protein genes cause endoplasmic reticulum (ER) retention and disease, presumably attributable to failure of misfolded proteins to pass the ER quality control. Because many myelin proteins partition into cholesterol-rich membrane rafts, their interaction with cholesterol could potentially be part of the ER quality control system. Here, we provide in vitro and in vivo evidence that the major peripheral myelin protein P0 requires cholesterol for exiting the ER and reaching the myelin compartment. Cholesterol dependency of P0 trafficking in heterologous cells is mediated by a cholesterol recognition/interaction amino acid consensus (CRAC) motif. Mutant mice lacking cholesterol biosynthesis in Schwann cells suffer from severe hypomyelination with numerous uncompacted myelin stretches. This demonstrates that high-level cholesterol coordinates P0 export with myelin membrane synthesis, which is required for the correct stoichiometry of myelin components and for myelin compaction.
Magnetic phase transition in MnFeP0.5As0.4Si0.1
International Nuclear Information System (INIS)
Wang, J L; Campbell, S J; Tegus, O; Brueck, E; Dou, S X
2010-01-01
We have carried out a detailed investigation of the magnetic phase transition in MnFeP 0.5 As 0.4 Si 0.1 . Temperature hysteresis has been observed in the variable temperature magnetization curves (B appl = 0.01 T) with T C W ∼ 302 K on warming and T C C ∼ 292 K on cooling. The first order nature of this transition in MnFeP 0.5 As 0.4 Si 0.1 is confirmed by the negative slope obtained from isotherms of M 2 versus B/M around the critical temperature. Linear thermal expansion measurements reveal a large volume change, ΔV/V∼8.7x10 -3 at the magnetic phase transition and that this magnetovolume effect is suppressed to ΔV/V ∼ 5.5x10 -3 in an applied field of B appl = 1.0 T. Analyses of 57 Fe Moessbauer spectra (4.5 - 300 K) using a random distribution model and taking nearest-neighbour environments into account, indicate that the paramagnetic and ferromagnetic phases coexist over a temperature range of ∼ 45 K around the Curie temperature. The Debye temperature for MnFeP 0.5 As 0.4 Si 0.1 has been evaluated as θ D = 350 ± 20 K from the temperature dependence of the average isomer shift.
Thermo-elektrische materialen : Peltier energy harvesting
Beurden, K.M.M. (Karin); Goselink, E.A. (Erik)
2013-01-01
Thermo-elektrische materialen zijn al sinds de 19e eeuw bekend. In 1834 ontdekte de Franse natuurkundige Jean Peltier dat er warmte wordt getransporteerd van de overgang tussen twee metalen wanneer er een elektrische stroom vloeit door het grensvlak. Het grote voordeel van Peltier elementen is dat
Biomass thermo-conversion. Research trends
International Nuclear Information System (INIS)
Rodriguez Machin, Lizet; Perez Bermudez, Raul; Quintana Perez, Candido Enrique; Ocanna Guevara, Victor Samuel; Duffus Scott, Alejandro
2011-01-01
In this paper is studied the state of the art in order to identify the main trends of the processes of thermo conversion of biomass into fuels and other chemicals. In Cuba, from total supply of biomass, wood is the 19% and sugar cane bagasse and straw the 80%, is why research in the country, should be directed primarily toward these. The methods for energy production from biomass can be group into two classes: thermo-chemical and biological conversion routes. The technology of thermo-chemical conversion includes three subclasses: pyrolysis, gasification, and direct liquefaction. Although pyrolysis is still under development, in the current energy scenario, has received special attention, because can convert directly biomass into solid, liquid and gaseous by thermal decomposition in absence of oxygen. The gasification of biomass is a thermal treatment, where great quantities of gaseous products and small quantities of char and ash are produced. In Cuba, studies of biomass thermo-conversion studies are limited to slow pyrolysis and gasification; but gas fuels, by biomass, are mainly obtained by digestion (biogas). (author)
Verification and validation of multi-group library MUSE1.0 created from ENDF/B-VII.0
International Nuclear Information System (INIS)
Chen Yixue; Wu Jun; Yang Shouhai; Zhang Bin; Lu Daogang; Chen Chaobin
2010-01-01
A multi-group library set named MUSE1.0 with 172-neutron group and 42-photon group is produced based on ENDF/B-VII.0 using NJOY code. Weight function of the multi-group library set is taken from the Vitanim-e library and the max legendre order of scattering matrix is six. All the nuclides have thermal scattering data created using free-gas scattering law and 10 Bondarenko background cross sections se lected to generate the self-shielded multi-group cross sections. The final libraries have GENDF-format, MATXS-format and ACE-multi-group sub-libraries and each sub-library generated under 4 temperatures(293 K,600 K,800 K and 900 K). This paper provides a summary of the procedure to produce the library set and a detail description of the validation of the multi-group library set by several critical benchmark devices and shielding benchmark devices using MCNP code. The ability to handle the thermal neutron transport and resonance self-shielding problems are investigated specially. In the end, we draw the conclusion that the multi-group libraries produced is credible and can be used in the R and D process of Supercritical Water Reactor Design. (authors)
Initial clinical results with the ThermoCool® SmartTouch® Surround Flow catheter.
Gonna, Hanney; Domenichini, Giulia; Zuberi, Zia; Norman, Mark; Kaba, Riyaz; Grimster, Alexander; Gallagher, Mark M
2017-08-01
The Biosense Webster ThermoCool® SmartTouch® Surround Flow (STSF) catheter is a recently developed ablation catheter incorporating Surround Flow (SF) technology to ensure efficient cooling and force sensing to quantify tissue contact. In our unit, it superseded the ThermoCool® SF catheter from the time of its introduction in May 2015. Procedure-related data were collected prospectively for the first 100 ablation procedures performed in our department using the STSF catheter. From a database of 654 procedures performed in our unit using the SF catheter, we selected one to match each STSF procedure, matching for procedure type, operator experience, patient age, and gender. The groups were well matched for patient age, gender, and procedure type. Procedure duration was similar in both groups (mean 225.5 vs. 221.4 min, IQR 106.5 vs. 91.5, P = 0.55), but fluoroscopy duration was shorter in the STSF group (mean 25.8 vs. 30.0, IQR 19.6 vs. 18.5, P = 0.03). No complication occurred in the STSF group. Complications occurred in two cases in the SF group (one pericardial effusion requiring drainage and one need for permanent pacing). Complete procedural success was achieved in 98 cases in the STSF group and 94 cases in the SF group (P = 0.15). The composite endpoint of procedure failure or acute complication was less common in the STSF group (2 vs. 8, P = 0.05). The STSF catheter is safe and effective in treating a range of arrhythmias. Compared with the SF catheter, it shows a trend towards improved safety-efficacy balance. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2016. For permissions please email: journals.permissions@oup.com.
Directory of Open Access Journals (Sweden)
Kawakami Minoru
2011-11-01
Full Text Available Abstract Background Neural crest cells (NCCs are embryonic, multipotent stem cells. Their long-range and precision-guided migration is one of their most striking characteristics. We previously reported that P0-Cre/CAG-CAT-lacZ double-transgenic mice showed significant lacZ expression in tissues derived from NCCs. Results In this study, by embedding a P0-Cre/CAG-CAT-EGFP embryo at E9.5 in collagen gel inside a culture glass slide, we were able to keep the embryo developing ex vivo for more than 24 hours; this development was with enough NCC fluorescent signal intensity to enable single-cell resolution analysis, with the accompanying NCC migration potential intact and with the appropriate NCC response to the extracellular signal maintained. By implantation of beads with absorbed platelet-derived growth factor-AA (PDGF-AA, we demonstrated that PDGF-AA acts as an NCC-attractant in embryos. We also performed assays with NCCs isolated from P0-Cre/CAG-CAT-EGFP embryos on culture plates. The neuromediator 5-hydroxytryptamine (5-HT has been known to regulate NCC migration. We newly demonstrated that dopamine, in addition to 5-HT, stimulated NCC migration in vitro. Two NCC populations, with different axial levels of origins, showed unique distribution patterns regarding migration velocity and different dose-response patterns to both 5-HT and dopamine. Conclusions Although avian species predominated over the other species in the NCC study, our novel system should enable us to use mice to assay many different aspects of NCCs in embryos or on culture plates, such as migration, division, differentiation, and apoptosis.
X-ray diffraction on MnFeP0.46As0.54 in a magnetic field
International Nuclear Information System (INIS)
Tegus, O.; Koyama, K.; Her, J.L.; Watanabe, K.; Brueck, E.; Buschow, K.H.J.; Boer, F.R. de
2007-01-01
We have performed powder X-ray-diffraction measurements on MnFeP 0.46 As 0.54 in fields up to 5T in the temperature range 8-310K. The compound which has the hexagonal Fe 2 P type of structure shows a field-induced isostructural phase transition. We found that the cell volume decreases slightly and continuously with increasing magnetic field, although the lattice parameter ratio c/a drastically changes. A tentative analysis of the dependence of the lattice parameters on the magnetization has been carried out using the extended Bean-Rodbell model
Group-decoupled multi-group pin power reconstruction utilizing nodal solution 1D flux profiles
International Nuclear Information System (INIS)
Yu, Lulin; Lu, Dong; Zhang, Shaohong; Wang, Dezhong
2014-01-01
Highlights: • A direct fitting multi-group pin power reconstruction method is developed. • The 1D nodal solution flux profiles are used as the condition. • The least square fit problem is analytically solved. • A slowing down source improvement method is applied. • The method shows good accuracy for even challenging problems. - Abstract: A group-decoupled direct fitting method is developed for multi-group pin power reconstruction, which avoids both the complication of obtaining 2D analytic multi-group flux solution and any group-coupled iteration. A unique feature of the method is that in addition to nodal volume and surface average fluxes and corner fluxes, transversely-integrated 1D nodal solution flux profiles are also used as the condition to determine the 2D intra-nodal flux distribution. For each energy group, a two-dimensional expansion with a nine-term polynomial and eight hyperbolic functions is used to perform a constrained least square fit to the 1D intra-nodal flux solution profiles. The constraints are on the conservation of nodal volume and surface average fluxes and corner fluxes. Instead of solving the constrained least square fit problem numerically, we solve it analytically by fully utilizing the symmetry property of the expansion functions. Each of the 17 unknown expansion coefficients is expressed in terms of nodal volume and surface average fluxes, corner fluxes and transversely-integrated flux values. To determine the unknown corner fluxes, a set of linear algebraic equations involving corner fluxes is established via using the current conservation condition on all corners. Moreover, an optional slowing down source improvement method is also developed to further enhance the accuracy of the reconstructed flux distribution if needed. Two test examples are shown with very good results. One is a four-group BWR mini-core problem with all control blades inserted and the other is the seven-group OECD NEA MOX benchmark, C5G7
Multi-group transport methods for high-resolution neutron activation analysis
International Nuclear Information System (INIS)
Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.
2009-01-01
The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)
Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh
International Nuclear Information System (INIS)
Aggery, A.
1999-12-01
The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)
Hydrogen transport in a toroidal plasma using multigroup discrete-ordinates methodology
International Nuclear Information System (INIS)
Wienke, B.R.; Miller, W.F. Jr.; Seed, T.J.
1979-01-01
Neutral hydrogen transport in a fully ionized two-dimensional tokamak plasma was examined using discrete ordinates and contrasted with earlier analyses. In particular, curvature effects induced by toroidal geometries and ray effects caused by possible source localization were investigated. From an overview of the multigroup discrete-ordinates approximation, methodology in two-dimensional cylindrical geometry is detailed, mesh and plasma zoning procedures are sketched, and the piecewise polynomial solution algorithm on a triangular domain is obtained. Toroidal effects and comparisons as related to reaction rates and perticle spectra are examined for various model and source configurations
Using the probability method for multigroup calculations of reactor cells in a thermal energy range
International Nuclear Information System (INIS)
Rubin, I.E.; Pustoshilova, V.S.
1984-01-01
The possibility of using the transmission probability method with performance inerpolation for determining spatial-energy neutron flux distribution in cells of thermal heterogeneous reactors is considered. The results of multigroup calculations of several uranium-water plane and cylindrical cells with different fuel enrichment in a thermal energy range are given. A high accuracy of results is obtained with low computer time consumption. The use of the transmission probability method is particularly reasonable in algorithms of the programmes compiled computer with significant reserve of internal memory
Simulate-HEX - The multi-group diffusion equation in hexagonal-z geometry
International Nuclear Information System (INIS)
Lindahl, S. O.
2013-01-01
The multigroup diffusion equation is solved for the hexagonal-z geometry by dividing each hexagon into 6 triangles. In each triangle, the Fourier solution of the wave equation is approximated by 8 plane waves to describe the intra-nodal flux accurately. In the end an efficient Finite Difference like equation is obtained. The coefficients of this equation depend on the flux solution itself and they are updated once per power/void iteration. A numerical example demonstrates the high accuracy of the method. (authors)
NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK
Directory of Open Access Journals (Sweden)
Filip Osuský
2016-12-01
Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.
REX1-87, Multigroup Neutron Cross-Sections from ENDF/B
International Nuclear Information System (INIS)
Gopalakrishnan, V.; Ganesan, S.
1988-01-01
1 - Description of program or function: The program calculates self- shielding factors for reactor applications from a pre-processed (linearized) evaluated nuclear data file in the ENDF/B format. 2 - Method of solution: Bondarenko definition of multigroup self- shielding factors invoking narrow resonance treatment is used. 3 - Restrictions on the complexity of the problem: a) Maximum no. of energy group is 620. b) Only the built-in forms of the weighting functions can be chosen. c) The program is strictly limited to resolved resonance region from physical considerations
Preparation of multigroup lumped fission product cross-sections from ENDF/B-VI for FBRs
International Nuclear Information System (INIS)
Devan, K.; Gopalakrishnan, V.; Mohanakrishnan, P.; Sridharan, M.S.
1997-01-01
Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed. (author)
Mining the multigroup-discrete ordinates algorithm for high quality solutions
International Nuclear Information System (INIS)
Ganapol, B.D.; Kornreich, D.E.
2005-01-01
A novel approach to the numerical solution of the neutron transport equation via the discrete ordinates (SN) method is presented. The new technique is referred to as 'mining' low order (SN) numerical solutions to obtain high order accuracy. The new numerical method, called the Multigroup Converged SN (MGCSN) algorithm, is a combination of several sequence accelerators: Romberg and Wynn-epsilon. The extreme accuracy obtained by the method is demonstrated through self consistency and comparison to the independent semi-analytical benchmark BLUE. (authors)
International Nuclear Information System (INIS)
Kim, Kyung-O; Jeong, Hae Sun; Jo, Daeseong
2017-01-01
Highlights: • Employing the Radial Point Interpolation Method (RPIM) in numerical analysis of multi-group neutron-diffusion equation. • Establishing mathematical formation of modified multi-group neutron-diffusion equation by RPIM. • Performing the numerical analysis for 2D critical problem. - Abstract: A mesh-free method is introduced to overcome the drawbacks (e.g., mesh generation and connectivity definition between the meshes) of mesh-based (nodal) methods such as the finite-element method and finite-difference method. In particular, the Point Interpolation Method (PIM) using a radial basis function is employed in the numerical analysis for the multi-group neutron-diffusion equation. The benchmark calculations are performed for the 2D homogeneous and heterogeneous problems, and the Multiquadrics (MQ) and Gaussian (EXP) functions are employed to analyze the effect of the radial basis function on the numerical solution. Additionally, the effect of the dimensionless shape parameter in those functions on the calculation accuracy is evaluated. According to the results, the radial PIM (RPIM) can provide a highly accurate solution for the multiplication eigenvalue and the neutron flux distribution, and the numerical solution with the MQ radial basis function exhibits the stable accuracy with respect to the reference solutions compared with the other solution. The dimensionless shape parameter directly affects the calculation accuracy and computing time. Values between 1.87 and 3.0 for the benchmark problems considered in this study lead to the most accurate solution. The difference between the analytical and numerical results for the neutron flux is significantly increased in the edge of the problem geometry, even though the maximum difference is lower than 4%. This phenomenon seems to arise from the derivative boundary condition at (x,0) and (0,y) positions, and it may be necessary to introduce additional strategy (e.g., the method using fictitious points and
A self-consistent nodal method in response matrix formalism for the multigroup diffusion equations
International Nuclear Information System (INIS)
Malambu, E.M.; Mund, E.H.
1996-01-01
We develop a nodal method for the multigroup diffusion equations, based on the transverse integration procedure (TIP). The efficiency of the method rests upon the convergence properties of a high-order multidimensional nodal expansion and upon numerical implementation aspects. The discrete 1D equations are cast in response matrix formalism. The derivation of the transverse leakage moments is self-consistent i.e. does not require additional assumptions. An outstanding feature of the method lies in the linear spatial shape of the local transverse leakage for the first-order scheme. The method is described in the two-dimensional case. The method is validated on some classical benchmark problems. (author)
International Nuclear Information System (INIS)
Ozgener, B.; Ozgener, H.A.
2005-01-01
A multiregion, multigroup collision probability method with white boundary condition is developed for thermalization calculations of light water moderated reactors. Hydrogen scatterings are treated by Nelkin's kernel while scatterings from other nuclei are assumed to obey the free-gas scattering kernel. The isotropic return (white) boundary condition is applied directly by using the appropriate collision probabilities. Comparisons with alternate numerical methods show the validity of the present formulation. Comparisons with some experimental results indicate that the present formulation is capable of calculating disadvantage factors which are closer to the experimental results than alternative methods
Specifications for a two-dimensional multi-group scattering code: ALCI
International Nuclear Information System (INIS)
Bayard, J.P.; Guillou, A.; Lago, B.; Bureau du Colombier, M.J.; Guillou, G.; Vasseur, Ch.
1965-02-01
This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, RΘ. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [fr
Development of a polynomial nodal model to the multigroup transport equation in one dimension
International Nuclear Information System (INIS)
Feiz, M.
1986-01-01
A polynomial nodal model that uses Legendre polynomial expansions was developed for the multigroup transport equation in one dimension. The development depends upon the least-squares minimization of the residuals using the approximate functions over the node. Analytical expressions were developed for the polynomial coefficients. The odd moments of the angular neutron flux over the half ranges were used at the internal interfaces, and the Marshak boundary condition was used at the external boundaries. Sample problems with fine-mesh finite-difference solutions of the diffusion and transport equations were used for comparison with the model
Thermo-stimulated current and dielectric loss in composite materials
International Nuclear Information System (INIS)
Nishijima, S.; Hagihara, T.; Okada, T.
1986-01-01
Thermo-stimulated current and dielectric loss measurements have been performed on five kinds of commercially available composite materials in order to study the electric properties of composite materials at low temperatures. Thermo-stimulated current measurements have been made on the composite materials in which the matrix quality was changed intentionally. The changes in the matrices were introduced by gamma irradiation or different curing conditions. Thermo-stimulated current and dielectric loss measurements revealed the number and the molecular weight of dipolar molecules. The different features of thermo-stimulated current and dielectric losses were determined for different composite materials. The gamma irradiation and the curing conditions especially affect the thermo-stimulated current features. The changes in macroscopic mechanical properties reflect those of thermo-stimulated current. It was found that the change in quality and/or degradation of the composite materials could be detected by means of thermo-stimulated current and/or dielectric loss measurements
Search for the Pentaquark via the Decay $P^0_{\\bar{c}s} \\to \\phi \\pi \\rho$
Energy Technology Data Exchange (ETDEWEB)
Beck, S.Maytal [Tel Aviv U.
1998-01-01
This work reports results of the first search for the pentaquark, which is predicted to be a doublet of states: $P^0_{\\bar{c}s} = \\mid \\bar{c}suud$ > and $P^-_{\\bar{c}s}$ = $\\mid \\bar{c}sddu>$. The color hyperfine interaction between their constituent quarks results in a maximal binding potential of 150 MeV. Calculations done using other models predict that the pentaquark is either bound or is a near-threshold resonance. A bound pentaquark would have a mass below 2.907 GeV /$c^2$ and its lifetime would be like that of other charm particles, of the order of $10^{-13}$ s. Crude estimates of the pentaquark production cross section predict values of the order of 1 % of that of the $D_s$. Observation of the pentaquark is interesting for its unusual structure and would contribute to the understanding of QCD and the concept of confinement....
Response of the 1P0 resonance near n = 3 in the H- continuum to external electric fields
International Nuclear Information System (INIS)
Cohen, S.
1986-05-01
The response to external electric fields of the 1 P 0 resonance in the H - photodetachment continuum below the n = 3 hydrogenic excitation threshold is investigated. Using the relativistic (β = 0.806) 650 MeV H - beam at the Clinton P. Anderson Meson Physics Facility (LAMPF) in Los Alamos, the fourth harmonic (2.66 nm) of a Nd:YAG laser is Doppler shifted to provide a continuously tunable photon beam in the rest frame of the ions. The magnetic field from pulsed Helmholtz coils, surrounding the photon-H - interaction point provides a Lorentz-transformed barycentric electric field. Relative total photodetachment cross sections were measured as a function of photon energy and electric field. The resulting spectra were fit to a Fano line shape. 70 refs., 28 figs., 7 tabs
Energy Technology Data Exchange (ETDEWEB)
Lillie, R.A.; Robinson, J.C.
1976-05-01
The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. A finite element formulation, utilizing a canonical form of the transport equation, is here developed to obtain both integral and pointwise solutions to neutron transport problems. The formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included by employing discrete ordinates-like approximations. In addition, multigroup source outer iteration techniques are employed to perform group-dependent calculations. The ability of the formulation to reduce substantially ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal-type lattice. A small, high-leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation. 8 tables, 29 figures.
International Nuclear Information System (INIS)
Lillie, R.A.; Robinson, J.C.
1976-05-01
The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. A finite element formulation, utilizing a canonical form of the transport equation, is here developed to obtain both integral and pointwise solutions to neutron transport problems. The formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included by employing discrete ordinates-like approximations. In addition, multigroup source outer iteration techniques are employed to perform group-dependent calculations. The ability of the formulation to reduce substantially ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal-type lattice. A small, high-leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation. 8 tables, 29 figures
Variational P1 approximations of general-geometry multigroup transport problems
International Nuclear Information System (INIS)
Rulko, R.P.; Tomasevic, D.; Larsen, E.W.
1995-01-01
A variational approximation is developed for general-geometry multigroup transport problems with arbitrary anisotropic scattering. The variational principle is based on a functional that approximates a reaction rate in a subdomain of the system. In principle, approximations that result from this functional ''optimally'' determine such reaction rates. The functional contains an arbitrary parameter α and requires the approximate solutions of a forward and an adjoint transport problem. If the basis functions for the forward and adjoint solutions are chosen to be linear functions of the angular variable Ω, the functional yields the familiar multigroup P 1 equations for all values of α. However, the boundary conditions that result from the functional depend on α. In particular, for problems with vacuum boundaries, one obtains the conventional mixed boundary condition, but with an extrapolation distance that depends continuously on α. The choice α = 0 yields a generalization of boundary conditions derived earlier by Federighi and Pomraning for a more limited class of problems. The choice α = 1 yields a generalization of boundary conditions derived previously by Davis for monoenergetic problems. Other boundary conditions are obtained by choosing different values of α. The authors discuss this indeterminancy of α in conjunction with numerical experiments
Interface discontinuity factors in the modal Eigenspace of the multigroup diffusion matrix
International Nuclear Information System (INIS)
Garcia-Herranz, N.; Herrero, J.J.; Cuervo, D.; Ahnert, C.
2011-01-01
Interface discontinuity factors based on the Generalized Equivalence Theory are commonly used in nodal homogenized diffusion calculations so that diffusion average values approximate heterogeneous higher order solutions. In this paper, an additional form of interface correction factors is presented in the frame of the Analytic Coarse Mesh Finite Difference Method (ACMFD), based on a correction of the modal fluxes instead of the physical fluxes. In the ACMFD formulation, implemented in COBAYA3 code, the coupled multigroup diffusion equations inside a homogenized region are reduced to a set of uncoupled modal equations through diagonalization of the multigroup diffusion matrix. Then, physical fluxes are transformed into modal fluxes in the Eigenspace of the diffusion matrix. It is possible to introduce interface flux discontinuity jumps as the difference of heterogeneous and homogeneous modal fluxes instead of introducing interface discontinuity factors as the ratio of heterogeneous and homogeneous physical fluxes. The formulation in the modal space has been implemented in COBAYA3 code and assessed by comparison with solutions using classical interface discontinuity factors in the physical space. (author)
Continuous energy Monte Carlo method based homogenization multi-group constants calculation
International Nuclear Information System (INIS)
Li Mancang; Wang Kan; Yao Dong
2012-01-01
The efficiency of the standard two-step reactor physics calculation relies on the accuracy of multi-group constants from the assembly-level homogenization process. In contrast to the traditional deterministic methods, generating the homogenization cross sections via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum, thus provides more accuracy parameters. Besides, the same code and data bank can be used for a wide range of applications, resulting in the versatility using Monte Carlo codes for homogenization. As the first stage to realize Monte Carlo based lattice homogenization, the track length scheme is used as the foundation of cross section generation, which is straight forward. The scattering matrix and Legendre components, however, require special techniques. The Scattering Event method was proposed to solve the problem. There are no continuous energy counterparts in the Monte Carlo calculation for neutron diffusion coefficients. P 1 cross sections were used to calculate the diffusion coefficients for diffusion reactor simulator codes. B N theory is applied to take the leakage effect into account when the infinite lattice of identical symmetric motives is assumed. The MCMC code was developed and the code was applied in four assembly configurations to assess the accuracy and the applicability. At core-level, A PWR prototype core is examined. The results show that the Monte Carlo based multi-group constants behave well in average. The method could be applied to complicated configuration nuclear reactor core to gain higher accuracy. (authors)
Global dynamics of multi-group SEI animal disease models with indirect transmission
International Nuclear Information System (INIS)
Wang, Yi; Cao, Jinde
2014-01-01
A challenge to multi-group epidemic models in mathematical epidemiology is the exploration of global dynamics. Here we formulate multi-group SEI animal disease models with indirect transmission via contaminated water. Under biologically motivated assumptions, the basic reproduction number R 0 is derived and established as a sharp threshold that completely determines the global dynamics of the system. In particular, we prove that if R 0 <1, the disease-free equilibrium is globally asymptotically stable, and the disease dies out; whereas if R 0 >1, then the endemic equilibrium is globally asymptotically stable and thus unique, and the disease persists in all groups. Since the weight matrix for weighted digraphs may be reducible, the afore-mentioned approach is not directly applicable to our model. For the proofs we utilize the classical method of Lyapunov, graph-theoretic results developed recently and a new combinatorial identity. Since the multiple transmission pathways may correspond to the real world, the obtained results are of biological significance and possible generalizations of the model are also discussed
Schnettler, Berta; Miranda, Horacio; Miranda-Zapata, Edgardo; Salinas-Oñate, Natalia; Grunert, Klaus G; Lobos, Germán; Sepúlveda, José; Orellana, Ligia; Hueche, Clementina; Bonilla, Héctor
2017-06-01
This study examined longitudinal measurement invariance in the Satisfaction with Food-related Life (SWFL) scale using follow-up data from university students. We examined this measure of the SWFL in different groups of students, separated by various characteristics. Through non-probabilistic longitudinal sampling, 114 university students (65.8% female, mean age: 22.5) completed the SWFL questionnaire three times, over intervals of approximately one year. Confirmatory factor analysis was used to examine longitudinal measurement invariance. Two types of analysis were conducted: first, a longitudinal invariance by time, and second, a multigroup longitudinal invariance by sex, age, socio-economic status and place of residence during the study period. Results showed that the 3-item version of the SWFL exhibited strong longitudinal invariance (equal factor loadings and equal indicator intercepts). Longitudinal multigroup invariance analysis also showed that the 3-item version of the SWFL displays strong invariance by socio-economic status and place of residence during the study period over time. Nevertheless, it was only possible to demonstrate equivalence of the longitudinal factor structure among students of both sexes, and among those older and younger than 22 years. Generally, these findings suggest that the SWFL scale has satisfactory psychometric properties for longitudinal measurement invariance in university students with similar characteristics as the students that participated in this research. It is also possible to suggest that satisfaction with food-related life is associated with sex and age. Copyright © 2017 Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
Olson, Gordon L.
2016-01-01
One-dimensional models for the transport of radiation through binary stochastic media do not work in multi-dimensions. Authors have attempted to modify or extend the 1D models to work in multidimensions without success. Analytic one-dimensional models are successful in 1D only when assuming greatly simplified physics. State of the art theories for stochastic media radiation transport do not address multi-dimensions and temperature-dependent physics coefficients. Here, the concept of effective opacities and effective heat capacities is found to well represent the ensemble averaged transport solutions in cases with gray or multigroup temperature-dependent opacities and constant or temperature-dependent heat capacities. In every case analyzed here, effective physics coefficients fit the transport solutions over a useful range of parameter space. The transport equation is solved with the spherical harmonics method with angle orders of n=1 and 5. Although the details depend on what order of solution is used, the general results are similar, independent of angular order. - Highlights: • Gray and multigroup radiation transport is done through 2D stochastic media. • Approximate models for the mean radiation field are found for all test problems. • Effective opacities are adjusted to fit the means of stochastic media transport. • Test problems include temperature dependent opacities and heat capacities • Transport solutions are done with angle orders n=1 and 5.
Nonequilibrium statistical averages and thermo field dynamics
International Nuclear Information System (INIS)
Marinaro, A.; Scarpetta, Q.
1984-01-01
An extension of thermo field dynamics is proposed, which permits the computation of nonequilibrium statistical averages. The Brownian motion of a quantum oscillator is treated as an example. In conclusion it is pointed out that the procedure proposed to computation of time-dependent statistical average gives the correct two-point Green function for the damped oscillator. A simple extension can be used to compute two-point Green functions of free particles
Meng, Fanli; Li, Yang; Zang, Zhenyuan; Li, Na; Ran, Ruixue; Cao, Yingxue; Li, Tianyu; Zhou, Quan; Li, Wenbin
2017-12-01
The soybean pod borer [SPB; Leguminivora glycinivorella (Matsumura) (Lepidoptera: Tortricidae)] is the most important soybean pest in northeastern Asia. Silencing genes using plant-mediated RNA-interference is a promising strategy for controlling SPB infestations. The ribosomal protein P0 is important for protein translation and DNA repair in the SPB. Thus, transferring P0 double-stranded RNA (dsRNA) into plants may help prevent SPB-induced damage. We investigated the effects of SpbP0 dsRNA injections and SpbP0 dsRNA-expressing transgenic soybean plants on the SPB. Larval mortality rates were greater for SpbP0 dsRNA-injected larvae (96%) than for the control larvae (31%) at 14 days after injections. Transgenic T 2 soybean plants expressing SpbP0 dsRNA sustained less damage from SPB larvae than control plants. In addition, the expression level of the SpbP0 gene decreased and the mortality rate increased when SPB larvae were fed on T 3 transgenic soybean pods. Moreover, the surviving larvae were deformed and exhibited inhibited growth. Silencing SpbP0 expression is lethal to the SPB. Transgenic soybean plants expressing SpbP0 dsRNA are more resistant to the SPB than wild-type plants. Thus, SpbP0 dsRNA-expressing transgenic plants may be useful for controlling insect pests. © 2017 Society of Chemical Industry. © 2017 Society of Chemical Industry.
The TRPM2 channel: A thermo-sensitive metabolic sensor.
Kashio, Makiko; Tominaga, Makoto
2017-09-03
Living organisms continually experience changes in ambient temperature. To detect such temperature changes for adaptive behavioral responses, we evolved the ability to sense temperature. Thermosensitive transient receptor potential (TRP) channels, so-called thermo-TRPs, are involved in many physiologic functions in diverse organisms and constitute important temperature sensors. One of the important roles of thermo-TRPs is detecting ambient temperature in sensory neurons. Importantly, the functional expression of thermo-TRPs is observed not only in sensory neurons but also in tissues and cells that are not exposed to drastic temperature changes, indicating that thermo-TRPs are involved in many physiologic functions within the body's normal temperature range. Among such thermo-TRPs, this review focuses on one thermo-sensitive metabolic sensor in particular, TRPM2, and summarizes recent progress to clarify the regulatory mechanisms and physiologic functions of TRPM2 at body temperature under various metabolic states.
International Nuclear Information System (INIS)
Stankovski, Z.; Zmijarevic, I.
1987-06-01
This paper presents two approximations used in multigroup two-dimensional transport calculations in large, very homogeneous media: isotropic reflection together with recently proposed group-dependent spatial representations. These approximations are implemented as standard options in APOLLO 2 assembly transport code. Presented example calculations show that significant savings in computational costs are obtained while preserving the overall accuracy
Brown, Gavin T. L.; Harris, Lois R.; O'Quin, Chrissie; Lane, Kenneth E.
2017-01-01
Multi-group confirmatory factor analysis (MGCFA) allows researchers to determine whether a research inventory elicits similar response patterns across samples. If statistical equivalence in responding is found, then scale score comparisons become possible and samples can be said to be from the same population. This paper illustrates the use of…
Molenaar, Dylan; Dolan, Conor V.; Wicherts, Jelle M.
2009-01-01
Research into sex differences in general intelligence, g, has resulted in two opposite views. In the first view, a g-difference is nonexistent, while in the second view, g is associated with a male advantage. Past research using Multi-Group Covariance and Mean Structure Analysis (MG-CMSA) found no sex difference in g. This failure raised the…
Energy Technology Data Exchange (ETDEWEB)
Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)
1968-06-15
Programme MULTI calculates the space energy distribution of thermal neutrons in a multizone, cylindrical, infinitely long reactor lattice by using the multigroup or multipoint P{sub 3} approximation. This report presents a short description of the algorithm and the programme and gives the instructions for its exploitation. (author)
International Nuclear Information System (INIS)
Si, S.
2012-01-01
The Universal Algorithm of Stiffness Confinement Method (UASCM) for neutron kinetics model of multi-dimensional and multi-group transport equations or diffusion equations has been developed. The numerical experiments based on transport theory code MGSNM and diffusion theory code MGNEM have demonstrated that the algorithm has sufficient accuracy and stability. (authors)
International Nuclear Information System (INIS)
Smith, L.A.; Gehin, J.C.; Worley, B.A.; Renier, J.P.
1994-01-01
The FOEHN critical experiments were analyzed to validate the use of multigroup cross sections in the design of the Advanced Neutron Source. Eleven critical configurations were evaluated using the KENO, DORT, and VENTURE neutronics codes. Eigenvalue and power density profiles were computed and show very good agreement with measured values
de Jong, M.G.; Pieters, R.; Stremersch, S.
2012-01-01
Answers to sensitive questions are prone to social desirability bias. If not properly addressed, the validity of the research can be suspect. This article presents multigroup item randomized response theory (MIRRT) to measure self-reported sensitive topics across cultures. The method was
International Nuclear Information System (INIS)
Modak, R.S.; Sahni, D.C.
1996-01-01
Some simple reciprocity-like relations that exist in multi-group neutron diffusion and transport theory over bare homogeneous regions are presented. These relations do not involve the adjoint solutions and are directly related to numerical schemes based on an explicit evaluation of the fission matrix. (author)
A Mathematical Model of the Thermo-Anemometric Flowmeter.
Korobiichuk, Igor; Bezvesilna, Olena; Ilchenko, Andriі; Shadura, Valentina; Nowicki, Michał; Szewczyk, Roman
2015-09-11
A thermo-anemometric flowmeter design and the principles of its work are presented in the article. A mathematical model of the temperature field in a stream of biofuel is proposed. This model allows one to determine the fuel consumption with high accuracy. Numerical modeling of the heater heat balance in the fuel flow of a thermo-anemometric flowmeter is conducted and the results are analyzed. Methods for increasing the measurement speed and accuracy of a thermo-anemometric flowmeter are proposed.
MPI version of NJOY and its application to multigroup cross-section generation
Energy Technology Data Exchange (ETDEWEB)
Alpan, A.; Haghighat, A.
1999-07-01
Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances
MPI version of NJOY and its application to multigroup cross-section generation
International Nuclear Information System (INIS)
Alpan, A.; Haghighat, A.
1999-01-01
Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances, temperatures
Thermo Techno Modern Analytical Equipment for Research and Industrial Laboratories
Directory of Open Access Journals (Sweden)
Khokhlov, S.V.
2014-03-01
Full Text Available A brief overview of some models of Thermo Techno analytical equipment and possible areas of their application is given. Thermo Techno Company was created in 2000 as a part of representative office of international corporation Thermo Fisher Scientific — world leader in manufacturing analytical equipments. Thermo Techno is a unique company in its integrated approach in solving the problems of the user, which includes a series of steps: setting the analytical task, selection of effective analysis methods, sample delivery and preparation as well as data transmitting and archiving.
Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm
Energy Technology Data Exchange (ETDEWEB)
Tavares, Matheus G.; Petersen, Claudio Z., E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), Capao do Leao, RS (Brazil). Departamento de Matematica e Estatistica; Schramm, Marcelo, E-mail: schrammmarcelo@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Centro de Engenharias; Zanette, Rodrigo, E-mail: rodrigozanette@hotmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Instituto de Matematica e Estatistica
2017-07-01
In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)
Approximate analytical solution of two-dimensional multigroup P-3 equations
International Nuclear Information System (INIS)
Matausek, M.V.; Milosevic, M.
1981-01-01
Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (orig./RW) [de
Energy Technology Data Exchange (ETDEWEB)
Zanette, Rodrigo; Petersen, Caudio Zen [Univ. Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Schramm, Marcello [Univ. Federal de Pelotas (Brazil). Centro de Engenharias; Zabadal, Jorge Rodolfo [Univ. Federal do Rio Grande do Sul, Tramandai (Brazil)
2017-05-15
In this paper a solution for the one-dimensional steady state Multilayer Multigroup Neutron Diffusion Equation in cartesian geometry by Fictitious Borders Power Method and a perturbative analysis of this solution is presented. For each new iteration of the power method, the neutron flux is reconstructed by polynomial interpolation, so that it always remains in a standard form. However when the domain is long, an almost singular matrix arises in the interpolation process. To eliminate this singularity the domain segmented in R regions, called fictitious regions. The last step is to solve the neutron diffusion equation for each fictitious region in analytical form locally. The results are compared with results present in the literature. In order to analyze the sensitivity of the solution, a perturbation in the nuclear parameters is inserted to determine how a perturbation interferes in numerical results of the solution.
Program to solve the multigroup discrete ordinates transport equation in (x,y,z) geometry
International Nuclear Information System (INIS)
Lathrop, K.D.
1976-04-01
Numerical formulations and programming algorithms are given for the THREETRAN computer program which solves the discrete ordinates, multigroup transport equation in (x,y,z) geometry. An efficient, flexible, and general data-handling strategy is derived to make use of three hierarchies of storage: small core memory, large core memory, and disk file. Data management, input instructions, and sample problem output are described. A six-group, S 4 , 18 502 mesh point, 2 800 zone, k/sub eff/ calculation of the ZPPR-4 critical assembly required 144 min of CDC-7600 time to execute to a convergence tolerance of 5 x 10 -4 and gave results in good qualitative agreement with experiment and other calculations. 6 references
MINX: a multigroup interpretation of nuclear X-sections from ENDF/B
International Nuclear Information System (INIS)
Weisbin, C.R.; Soran, P.D.; MacFarlane, R.E.; Harris, D.R.; LaBauve, R.J.; Hendricks, J.S.; White, J.E.; Kidman, R.B.
1976-09-01
MINX calculates fine-group averaged infinitely dilute cross sections, self-shielding factors, and group-to-group transfer matrices from ENDF/B-IV data. Its primary purpose is to generate pseudo-composition independent multigroup libraries in the standard CCCC-III interface formats for use in the design and analysis of nuclear systems. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX and ENDRUN and the high-Legendre-order transfer matrices of ETOG and SUPERTOG. Group structure, Legendre order, weight function, temperature, dilutions, and processing tolerances are all under user control. Paging and variable dimensioning allow very large problems to be run. Both CDC and IBM versions of MINX are available
Solution of the Multigroup-Diffusion equation by the response matrix method
International Nuclear Information System (INIS)
Oliveira, C.R.E.
1980-10-01
A preliminary analysis of the response matrix method is made, considering its application to the solution of the multigroup diffusion equations. The one-dimensional formulation is presented and used to test some flux expansions, seeking the application of the method to the two-dimensional problem. This formulation also solves the equations that arise from the integro-differential synthesis algorithm. The slow convergence of the power method, used to solve the eigenvalue problem, and its acceleration by means of the Chebyshev polynomial method, are also studied. An algorithm for the estimation of the dominance ratio is presented, based on the residues of two successive iteration vectors. This ratio, which is not known a priori, is fundamental for the efficiency of the method. Some numerical problems are solved, testing the 1D formulation of the response matrix method, its application to the synthesis algorithm and also, at the same time, the algorithm to accelerate the source problem. (Author) [pt
Global dynamics of a novel multi-group model for computer worms
International Nuclear Information System (INIS)
Gong Yong-Wang; Song Yu-Rong; Jiang Guo-Ping
2013-01-01
In this paper, we study worm dynamics in computer networks composed of many autonomous systems. A novel multi-group SIQR (susceptible-infected-quarantined-removed) model is proposed for computer worms by explicitly considering anti-virus measures and the network infrastructure. Then, the basic reproduction number of worm R 0 is derived and the global dynamics of the model are established. It is shown that if R 0 is less than or equal to 1, the disease-free equilibrium is globally asymptotically stable and the worm dies out eventually, whereas, if R 0 is greater than 1, one unique endemic equilibrium exists and it is globally asymptotically stable, thus the worm persists in the network. Finally, numerical simulations are given to illustrate the theoretical results. (general)
Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm
International Nuclear Information System (INIS)
Tavares, Matheus G.; Petersen, Claudio Z.; Schramm, Marcelo; Zanette, Rodrigo
2017-01-01
In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)
The Nodal Polynomial Expansion method to solve the multigroup diffusion equations
International Nuclear Information System (INIS)
Ribeiro, R.D.M.
1983-03-01
The methodology of the solutions of the multigroup diffusion equations and uses the Nodal Polynomial Expansion Method is covered. The EPON code was developed based upon the above mentioned method for stationary state, rectangular geometry, one-dimensional or two-dimensional and for one or two energy groups. Then, one can study some effects such as the influence of the baffle on the thermal flux by calculating the flux and power distribution in nuclear reactors. Furthermore, a comparative study with other programs which use Finite Difference (CITATION and PDQ5) and Finite Element (CHD and FEMB) Methods was undertaken. As a result, the coherence, feasibility, speed and accuracy of the methodology used were demonstrated. (Author) [pt
Hong, Yong-Rock; Holcomb, Derek; Ballard, Michael; Schwartz, Laurel
Winds of change have been blowing in the U.S. healthcare system since passage of the Affordable Care Act. Examining differences between individuals covered by different types of insurance is essential if healthcare executives are to develop new strategies in response to the emerging health insurance market. In this study, we used multigroup path analysis models to examine the moderating effects of health insurance on direct and indirect associations with general health status, satisfaction with received care, financial burden, and perceived value of the healthcare system. Data were obtained from the 2012 Medical Expenditure Panel Survey and analyzed according to the types of insurance: private, public, and military. With the satisfactory fit of the model (χ = 2,532.644, df = 96, p spending.
Approximate analytical solution of two-dimensional multigroup P-3 equations
International Nuclear Information System (INIS)
Matausek, M.V.; Milosevic, M.
1981-01-01
Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (author)
ETOA, ABBN Multigroup Constants from ENDF/B for Fast Reactors
International Nuclear Information System (INIS)
Nishimura, Hideo
1977-01-01
1 - Nature of physical problem solved: Production of ABBN type group constants up to 70 groups for fast reactor calculations, reading ENDF/B library as input. 2 - Method of solution: The multigroup method of Bondarenko et al. is used for processing basic nuclear data. Calculational algorithms for an unresolved resonance region are the same as those in the MC 2 code. For a resolved resonance region, an ultrafine energy structure dependent on a level scheme is adopted. 3 - Restrictions on the complexity of the problem: Maximum number of: energy groups: 70; sigma 0 values: 6; temperatures: 5. Self-shielding factors for an unrealistically low value of sigma 0 are not guaranteed because of the approximations used in the unresolved resonance region
Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model
Energy Technology Data Exchange (ETDEWEB)
Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)
2006-07-01
Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)
AMPX: a modular system for multigroup cross-section generation and manipulation
International Nuclear Information System (INIS)
Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Diggs, B.R.; Webster, C.C.; Lucius, J.L.; White, J.E.; Wright, R.Q.; Westfall, R.M.
1978-01-01
The AMPX system, developed at the Oak Ridge National Laboratory over the past seven years, is a collection of computer programs in a modular arrangement. Starting with ENDF-formatted nuclear data files, the system includes a full range of features needed to produce and use multigroup neutron, gamma-ray production, and gamma-ray interaction cross-section data. The balance between production and analysis is roughly even; thus, the system serves a wide variety of needs. The modularity is particularly attractive, since it allows the user to choose an arbitrary execution sequence from the approximately 40 to 50 modules available in the system. The modularity also allows selection from different treatments; e.g., the Nordheim method, a full-blown integral transport calculation, the Bondarenko method, or other alternative can be selected for resonance shielding. 2 figures
Analytic solutions of the multigroup space-time reactor kinetics equations
International Nuclear Information System (INIS)
Lee, C.E.; Rottler, S.
1986-01-01
The development of analytical and numerical solutions to the reactor kinetics equations is reviewed. Analytic solutions of the multigroup space-time reactor kinetics equations are developed for bare and reflected slabs and spherical reactors for zero flux, zero current and extrapolated endpoint boundary conditions. The material properties of the reactors are assumed constant in space and time, but spatially-dependent source terms and initial conditions are investigated. The system of partial differential equations is reduced to a set of linear ordinary differential equations by the Laplace transform method. These equations are solved by matrix Green's functions yielding a general matrix solution for the neutron flux and precursor concentration in the Laplace transform space. The detailed pole structure of the Laplace transform matrix solutions is investigated. The temporally- and spatially-dependent solutions are determined from the inverse Laplace transform using the Cauchy residue theorem, the theorem of Frobenius, a knowledge of the detailed pole structure and matrix operators. (author)
A multi-region boundary element method for multigroup neutron diffusion calculations
International Nuclear Information System (INIS)
Ozgener, H.A.; Ozgener, B.
2001-01-01
For the analysis of a two-dimensional nuclear system consisting of a number of homogeneous regions (termed cells), first the cell matrices which depend solely on the material composition and geometrical dimension of the cell (hence on the cell type) are constructed using a boundary element formulation based on the multigroup boundary integral equation. For a particular nuclear system, the cell matrices are utilized in the assembly of the global system matrix in block-banded form using the newly introduced concept of virtual side. For criticality calculations, the classical fission source iteration is employed and linear system solutions are by the block Gaussian-elimination algorithm. The numerical applications show the validity of the proposed formulation both through comparison with analytical solutions and assessment of benchmark problem results against alternative methods
Spectrum of the multigroup neutron transport operator for bounded spatial domains
International Nuclear Information System (INIS)
Larsen, E.W.
1979-01-01
The spectrum of the multigroup neutron transport operator A is studied for bounded spatial regions D which consist of a finite number of material subregions. Our main results provide simple conditions on the material cross sections which guarantee that (1) A possesses eigenvalues in the finite plane; (2) A possesses a ''leading'' eigenvalue lambda 0 which is real, not less than the real part of any other eigenvalue, and to which there corresponds at least one nonnegative eigenfunction psi/sub lambda/0; and (3) A possesses a ''dominant'' eigenvalue lambda 0 which is real, simple, greater than the real part of any other eigenvalue, and whose eigenfunction psi/sub lambda/0 satisfies psi/sub lambda/0> or =0 and ∫psi/sub lambda/0d 2 Ω>0. We give examples to illustrate the results and to show that a leading eigenvalue need not be simple, nor its eigenfunction(s) positive
The solution of the multigroup neutron transport equation using spherical harmonics
International Nuclear Information System (INIS)
Fletcher, K.
1981-01-01
A solution of the multi-group neutron transport equation in up to three space dimensions is presented. The flux is expanded in a series of unnormalised spherical harmonics. Using the various recurrence formulae a linked set of first order differential equations is obtained for the moments psisup(g)sub(lm)(r), γsup(g)sub(lm)(r). Terms with odd l are eliminated resulting in a second order system which is solved by two methods. The first is a finite difference formulation using an iterative procedure, secondly, in XYZ and XY geometry a finite element solution is given. Results for a test problem using both methods are exhibited and compared. (orig./RW) [de
Recent validation experience with multigroup cross-section libraries and scale
International Nuclear Information System (INIS)
Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Parks, C.V.; Petrie, L.M.
1995-01-01
This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment
LABAN-PEL: a two-dimensional, multigroup diffusion, high-order response matrix code
International Nuclear Information System (INIS)
Mueller, E.Z.
1991-06-01
The capabilities of LABAN-PEL is described. LABAN-PEL is a modified version of the two-dimensional, high-order response matrix code, LABAN, written by Lindahl. The new version extends the capabilities of the original code with regard to the treatment of neutron migration by including an option to utilize full group-to-group diffusion coefficient matrices. In addition, the code has been converted from single to double precision and the necessary routines added to activate its multigroup capability. The coding has also been converted to standard FORTRAN-77 to enhance the portability of the code. Details regarding the input data requirements and calculational options of LABAN-PEL are provided. 13 refs
International Nuclear Information System (INIS)
Moraes, Pedro Gabriel B.; Leite, Michel C.A.; Barros, Ricardo C.
2013-01-01
In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation
International Nuclear Information System (INIS)
Erradi, L.; Karouani, K.
1994-01-01
Many multigroup neutron cross section libraries have been processed from basic evaluated nuclear data for use in neutron dosimetry, reactor shielding calculation and in the development of fusion reactors. Most of these libraries have been tested only for fission spectra and were not validated for fusion spectra. Fifteen of these libraries such as DOSCROS84, IRDF85 and ENDFB5 have been used along with the neutron spectra unfolding code SAND II to evaluate about fifteen threshold detector saturated activities. The comparison between these computed activities and the measured ones of a set of foils placed in different places along the axis of a paraffin cylinder and irradiated by 14 MeV neutrons generated by a D-T source, hence giving rise to complex spectra, leads to different types of discrepancies. The analysis of these discrepancies allows to select from these libraries the ones that can be recommended. 1 fig., 4 refs. (author)
SIRIUS - A one-dimensional multigroup analytic nodal diffusion theory code
Energy Technology Data Exchange (ETDEWEB)
Forslund, P. [Westinghouse Atom AB, Vaesteraas (Sweden)
2000-09-01
In order to evaluate relative merits of some proposed intranodal cross sections models, a computer code called Sirius has been developed. Sirius is a one-dimensional, multigroup analytic nodal diffusion theory code with microscopic depletion capability. Sirius provides the possibility of performing a spatial homogenization and energy collapsing of cross sections. In addition a so called pin power reconstruction method is available for the purpose of reconstructing 'heterogeneous' pin qualities. consequently, Sirius has the capability of performing all the calculations (incl. depletion calculations) which are an integral part of the nodal calculation procedure. In this way, an unambiguous numerical analysis of intranodal cross section models is made possible. In this report, the theory of the nodal models implemented in sirius as well as the verification of the most important features of these models are addressed.
Zhuo, Tao; Li, Yuan-Yuan; Xiang, Hai-Ying; Wu, Zhan-Yu; Wang, Xian-Bin; Wang, Ying; Zhang, Yong-Liang; Li, Da-Wei; Yu, Jia-Lin; Han, Cheng-Gui
2014-06-01
Polerovirus P0 suppressors of host gene silencing contain a consensus F-box-like motif with Leu/Pro (L/P) requirements for suppressor activity. The Inner Mongolian Potato leafroll virus (PLRV) P0 protein (P0(PL-IM)) has an unusual F-box-like motif that contains a Trp/Gly (W/G) sequence and an additional GW/WG-like motif (G139/W140/G141) that is lacking in other P0 proteins. We used Agrobacterium infiltration-mediated RNA silencing assays to establish that P0(PL-IM) has a strong suppressor activity. Mutagenesis experiments demonstrated that the P0(PL-IM) F-box-like motif encompasses amino acids 76-LPRHLHYECLEWGLLCG THP-95, and that the suppressor activity is abolished by L76A, W87A, or G88A substitution. The suppressor activity is also weakened substantially by mutations within the G139/W140/G141 region and is eliminated by a mutation (F220R) in a C-terminal conserved sequence of P0(PL-IM). As has been observed with other P0 proteins, P0(PL-IM) suppression is correlated with reduced accumulation of the host AGO1-silencing complex protein. However, P0(PL-IM) fails to bind SKP1, which functions in a proteasome pathway that may be involved in AGO1 degradation. These results suggest that P0(PL-IM) may suppress RNA silencing by using an alternative pathway to target AGO1 for degradation. Our results help improve our understanding of the molecular mechanisms involved in PLRV infection.
Dynamic Modeling of ThermoFluid Systems
DEFF Research Database (Denmark)
Jensen, Jakob Munch
2003-01-01
The objective of the present study has been to developed dynamic models for two-phase flow in pipes (evaporation and condensation). Special attention has been given to modeling evaporators for refrigeration plant particular dry-expansion evaporators. Models of different complexity have been...... formulated. The different models deviate with respect to the detail¿s included and calculation time in connection with simulation. The models have been implemented in a new library named ThermoTwoPhase to the programming language Modelica. A test rig has been built with an evaporator instrumented in a way...
Thermo field theory versus imaginary time formalism
International Nuclear Information System (INIS)
Fujimoto, Y.; Nishino, H.; Grigjanis, R.
1983-11-01
We calculate a two-loop diagram at finite temperature to compare Thermo Field Theory (=Th.F.Th.) with the conventional imaginary time formalism (=Im.T.F.). The summation over the Matsubara frequency in Im.T.F. is carried out at two-loop level, and the result is shown to coincide with that of Th.F.Th. We confirm that in Im.T.F. the temperature dependent divergences cancel out at least in the calculation of effective potential of phi 4 theory, as in Th.F.Th. (author)
Thermo Scientific Ozone Analyzer Instrument Handbook
Energy Technology Data Exchange (ETDEWEB)
Springston, S. R. [Brookhaven National Lab. (BNL), Upton, NY (United States)
2016-03-01
The primary measurement output from the Thermo Scientific Ozone Analyzer is the concentration of the analyte (O3) reported at 1-s resolution in units of ppbv in ambient air. Note that because of internal pneumatic switching limitations the instrument only makes an independent measurement every 4 seconds. Thus, the same concentration number is repeated roughly 4 times at the uniform, monotonic 1-s time base used in the AOS systems. Accompanying instrument outputs include sample temperatures, flows, chamber pressure, lamp intensities and a multiplicity of housekeeping information. There is also a field for operator comments made at any time while data is being collected.
Introduction to thermo-fluids systems design
Garcia McDonald, André
2012-01-01
A fully comprehensive guide to thermal systems design covering fluid dynamics, thermodynamics, heat transfer and thermodynamic power cycles Bridging the gap between the fundamental concepts of fluid mechanics, heat transfer and thermodynamics, and the practical design of thermo-fluids components and systems, this textbook focuses on the design of internal fluid flow systems, coiled heat exchangers and performance analysis of power plant systems. The topics are arranged so that each builds upon the previous chapter to convey to the reader that topics are not stand-alone i
Thermo-Physical Properties of Selected Inconel
Directory of Open Access Journals (Sweden)
Krajewski P.K.
2014-10-01
Full Text Available The paper brings results of examinations of main thermo-physical properties of selected Inconel alloys, i.e. their heat diffusivity, thermal conductivity and heat capacity, measured in wide temperature range of 20 – 900 oC. Themathematical relationships of the above properties vs. temperature were obtained for the IN 100 and IN 713C alloys. These data can be used when modelling the IN alloys solidification processes aimed at obtaining required structure and properties as well as when designing optimal work temperature parameters.
XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections
International Nuclear Information System (INIS)
Ganesan, S.; Jagannathan, V.; Thiyagarajan, T.K.
2005-01-01
1 - Description of program or function: XnWlup is a computer program with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualisation of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. IAEA1395/05: New features of version 3.0: - Plotting absorption and fission cross sections of resonant nuclide after applying the self-shielding cross section. - Plotting the data of Resonant Integral table, as a function of dilution cross section for a selected temperature and for a given energy group. - Plotting the data of Resonant Integral table, as a function of temperature for a selected background dilution cross section and for a given energy group. - Clearing all the graphs except one graph from the display screen is easily done by using a tool bar button. - Displaying the coordinate of the cursor point with appropriate units. 2 - Methods: XnWlup helps to obtain histogram plots of the values of cross section data of an element/isotope available as 69-group WIMS-D library as a function of energy bins. The software XnWlup is developed with this graphical user interface in order to help those users who frequently refer to the WIMS-D library cross section data of neutron-nuclear reactions. The software also helps to produce handbook of WIMS-D cross sections
An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons
International Nuclear Information System (INIS)
Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.
2013-01-01
Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons
Testing a new multigroup inference approach to reconstructing past environmental conditions
Directory of Open Access Journals (Sweden)
Maria RIERADEVALL
2008-08-01
Full Text Available A new, quantitative, inference model for environmental reconstruction (transfer function, based for the first time on the simultaneous analysis of multigroup species, has been developed. Quantitative reconstructions based on palaeoecological transfer functions provide a powerful tool for addressing questions of environmental change in a wide range of environments, from oceans to mountain lakes, and over a range of timescales, from decades to millions of years. Much progress has been made in the development of inferences based on multiple proxies but usually these have been considered separately, and the different numeric reconstructions compared and reconciled post-hoc. This paper presents a new method to combine information from multiple biological groups at the reconstruction stage. The aim of the multigroup work was to test the potential of the new approach to making improved inferences of past environmental change by improving upon current reconstruction methodologies. The taxonomic groups analysed include diatoms, chironomids and chrysophyte cysts. We test the new methodology using two cold-environment training-sets, namely mountain lakes from the Pyrenees and the Alps. The use of multiple groups, as opposed to single groupings, was only found to increase the reconstruction skill slightly, as measured by the root mean square error of prediction (leave-one-out cross-validation, in the case of alkalinity, dissolved inorganic carbon and altitude (a surrogate for air-temperature, but not for pH or dissolved CO2. Reasons why the improvement was less than might have been anticipated are discussed. These can include the different life-forms, environmental responses and reaction times of the groups under study.
International Nuclear Information System (INIS)
Schriewer, J.; Hehn, G.; Mattes, M.; Pfister, G.; Keinert, J.
1978-01-01
Calculations were made for different benchmark experiments in order to test the coupled multigroup neutron and gamma library EURLIB-3 with 100 neutron groups and 20 gamma groups. In cooperation with EURATOM, Ispra, we produced this shielding library recently from ENDF/B-IV data for application in fission and fusion technology. Integral checks were performed for natural lithium, carbon, oxygen, and iron. Since iron is the most important structural material in nuclear technology, we started with calculations of iron benchmark experiments. Most of them are integral experiments of INR, Karlsruhe, but comparisons were also done with benchmark experiments from USA and Japan. For the experiments with fission sources we got satisfying results. All details of the resonances cannot be checked with flux measurements and multigroup cross sections used. But some averaged resonance behaviour of the measured and calculated fluxes can be compared and checked within the error limits given. We get greater differences in the calculations of benchmark experiments with 14 MeV neutron sources. For iron the group cross sections of EURLIB-3 produce an underestimation of the neutron flux in a broad energy region below the source energy. The conclusion is that the energy degradation by inelastic scattering is too strong. For fusion application the anisotropy of the inelastic scatter process must be taken into account, which isn't done by the processing codes at present. If this effect isn't enough, additional corrections have to be applied to the inelastic cross sections of iron in ENDF/B-IV. (author)
DEFF Research Database (Denmark)
Moldovan, M.; Rosberg, M. R.; Alvarez Herrero, Susana
2016-01-01
Mice deficient of myelin protein zero (P0) are established models of demyelinating Charcot-Marie-Tooth (CMT) disease. Recent work form our laboratory indicated that in severely affected P0−/− as well as in P0+/− (modeling CMT1B), the neuropathy is aggravated by associated changes in voltage...... function up to 2 hours after the blockers. Overall, the baseline excitability measures were much more abnormal in P0−/− at 4 months as compared to P0+/− at 20 months. Nevertheless, in both models, the NaV1.8 blockers produced similar deviations in excitability at a dose of 100 mg/Kg. Most notably...
Preparation of thermo-responsive membranes. II.
Nozawa, I; Suzuki, Y; Sato, S; Sugibayashi, K; Morimoto, Y
1991-05-01
Two types of liquid crystal (LC)-immobilized membranes were prepared by a soaking method and sandwich method to control the permeation of indomethacin, as a model drug, in response to local and systemic fever. Monooxyethylene trimethylolpropane tristearate (MTTS) was used as a model LC because it has a gel-liquid crystal phase transition temperature near the body temperature, 39-40 degrees C in phosphate buffered saline (pH 7.4). Two porous polypropylene (PP) membranes were soaked into 20% MTTS chloroform solution in the soaking method, and two PP membranes were poured with the melted MTTS and pressed in the sandwich method. Thermo-response efficacy of the soaked membrane was dependent upon the content of MTTS in MTTS membrane, and the MTTS content above the void volume of PP membrane (38%) was needed for high efficacy. On the other hand, the sandwich membrane exhibited higher thermo-response efficacy than the soaked membrane, because more LC was embedded in the pores of sandwich membrane than that of the soaked membrane. The sandwich membrane permeation of indomethacin was sharply controlled by temperature changes between 32 and 38 degrees C.
Thermo-optical Properties of Nanofluids
International Nuclear Information System (INIS)
Ortega, Maria Alejandra; Echevarria, Lorenzo; Rodriguez, Luis; Castillo, Jimmy; Fernandez, Alberto
2008-01-01
In this work, we report thermo-optical properties of nanofluids. Spherical gold nanoparticles obtained by laser ablation in condensed media were characterized using thermal lens spectroscopy in SDS-water solution pumping at 532 nm with a 10 ns pulsed laser-Nd-YAG system. Nanoparticles obtained by laser ablation were stabilized in the time by surfactants (Sodium Dodecyl-Sulfate or SDS) in different molar concentrations. The morphology and size of the gold nanoparticles were determined by transmission electron microscopy (TEM). The plasmonic resonance bands in gold nanoparticles are responsible of the light optical absorption of this wavelength. The position of the absorption maximum and width band in the UV-Visible spectra is given by the morphological characteristics of these systems. The thermo-optical constant such as thermal diffusion, thermal conductivity and dn/dT are functions of nanoparticles sizes and dielectric constant of the media. The theoretical model existents do not describe completely this relations because is not possible separate the contributions due to nanoparticles size, factor form and dielectric constant. The thermal lens signal obtained is also dependent of nanoparticles sizes. This methodology can be used in order to evaluate nanofluids and characterizing nanoparticles in different media. These results are expected to have an impact in bioimaging, biosensors and other technological applications such as cooler system
On the general theory of thermo-elastic friction
Alblas, J.B.
1961-01-01
A theory of the thermo-elastic dissipation in vibrating bodies is developed, starting from the three-dimensional thermo-elastic equations. After a discussion of the basic thermodynamical foundations, some general considerations on the problem of the conversion of mechanical energy into heat are
MC2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data
International Nuclear Information System (INIS)
2001-01-01
1 - Description of program or function: MC 2 -2 solves the neutron slowing-down equations using basic neutron data derived from ENDF/B data files to determine fundamental mode spectra for use in generating multigroup neutron cross sections. The current edition includes the ability to treat all ENDF/B-V and -VI data representations. It accommodates high-order P scattering representations and provides numerous capabilities such as isotope mixing, delayed neutron processing, free-format input, and flexibility in output data selection. This edition supersedes previous releases of the MC22 program and the earlier MC2 program. Improved physics algorithms and increased computational efficiency are incorporated. Input data files required by MC2-2 may be generated from ENDF/B data by the code ETOE-2. The hyper-fine-group integral transport theory module of MC2-2, RABANL, is an improved version of the RABBLE/RABID codes. Many of the MC2-2 modules are used in the SDX code. 2 - Methods: The extended transport P1, B1, consistent P1, and consistent B1 fundamental mode ultra-fine-group equations are solved using continuous slowing-down theory and multigroup methods. Fast and accurate resonance integral methods are used in the narrow resonance resolved and unresolved resonance treatments. A fundamental mode homogeneous unit cell calculation is performed using either a multigroup or a continuous slowing-down treatment. Multigroup neutron homogeneous cross sections are generated in an ISOTXS format for an arbitrary group structure. A hyper-fine-group integral transport slowing down calculation (RABANL) is available as an option. RABANL performs a homogeneous or heterogeneous (pin or slab) unit cell calculation over the resonance region (resolved and unresolved) and generates multigroup neutron cross sections in an ISOTXS format. Neutron cross sections are generated by RABANL for the homogeneous unit cell and for each heterogeneous region in the pin or slab unit cell calculation
Energy Technology Data Exchange (ETDEWEB)
Li, M
1998-08-01
In this thesis, two methods for solving the multigroup Boltzmann equation have been studied: the interface-current method and the Monte Carlo method. A new version of interface-current (IC) method has been develop in the TDT code at SERMA, where the currents of interface are represented by piecewise constant functions in the solid angle space. The convergence of this method to the collision probability (CP) method has been tested. Since the tracking technique is used for both the IC and CP methods, it is necessary to normalize he collision probabilities obtained by this technique. Several methods for this object have been studied and implemented in our code, we have compared their performances and chosen the best one as the standard choice. The transfer matrix treatment has been a long-standing difficulty for the multigroup Monte Carlo method: when the cross-sections are converted into multigroup form, important negative parts will appear in the angular transfer laws represented by low-order Legendre polynomials. Several methods based on the preservation of the first moments, such as the discrete angles methods and the equally-probable step function method, have been studied and implemented in the TRIMARAN-II code. Since none of these codes has been satisfactory, a new method, the non equally-probably step function method, has been proposed and realized in our code. The comparisons for these methods have been done in several aspects: the preservation of the moments required, the calculation of a criticality problem and the calculation of a neutron-transfer in water problem. The results have showed that the new method is the best one in all these comparisons, and we have proposed that it should be a standard choice for the multigroup transfer matrix. (author) 76 refs.
International Nuclear Information System (INIS)
Peduzzi, Emanuela; Tock, Laurence; Boissonnet, Guillaume; Maréchal, François
2013-01-01
In a carbon and resources constrained world, thermo-chemical conversion of lignocellulosic biomass into fuels and chemicals is regarded as a promising alternative to fossil resources derived products. Methanol is one potential product which can be used for the synthesis of various chemicals or as a fuel in fuel cells and internal combustion engines. This study focuses on the evaluation and optimization of the thermodynamic and economic performance of methanol production from biomass by applying process integration and optimization techniques. Results reveal the importance of the energy integration and in particular of the cogeneration of electricity for the efficient use of biomass. - Highlights: • A thermo-economic model for biomass conversion into methanol is developed. • Process integration and multi-objective optimization techniques are applied. • Results reveal the importance of energy integration for electricity co-generation
International Nuclear Information System (INIS)
Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.
1995-05-01
The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are ∼ 13%, while the average differences are < 8%
International Nuclear Information System (INIS)
Chalhoub, E.S.; Moraes, M. de.
1984-01-01
A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author) [pt
International Nuclear Information System (INIS)
Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.
1976-07-01
The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)
International Nuclear Information System (INIS)
Petersen, Claudio Zen; Vilhena, Marco T.; Barros, Ricardo C.
2009-01-01
In this paper the application of the Laplace transform method is described in order to determine the energy-dependent albedo matrix that is used in the boundary conditions multigroup neutron diffusion eigenvalue problems in slab geometry for nuclear reactor global calculations. In slab geometry, the diffusion albedo substitutes without approximation the baffle-reflector system around the active domain. Numerical results to typical test problems are shown to illustrate the accuracy and the efficiency of the Chebysheff acceleration scheme. (orig.)
International Nuclear Information System (INIS)
Santos, R.S. dos
1993-01-01
This paper presents a computational program to solve numerically the reactor kinetics equations in the multigroup diffusion theory. One or two-dimensional problems in cylindrical or Cartesian geometries, with any number of energy and delayed-neutron precursors groups are dealt with. The main input and output of the program are briefly discussed. Various results demonstrate the accuracy and versatility of the program, when compared with other kinetics programs. (author)
A multi-group and preemptable scheduling of cloud resource based on HTCondor
Jiang, Xiaowei; Zou, Jiaheng; Cheng, Yaodong; Shi, Jingyan
2017-10-01
and LHAASO. The result indicates that multi-group and preemptable resource scheduling is efficient to support multi-group and soft preemption. Additionally, the permission controlling component has been used in the local computing cluster, supporting for experiment JUNO, CMS and LHAASO, and the scale will be expanded to more experiments at the first half year, including DYW, BES and so on. Its evidence that the permission controlling is efficient.
Obtaining incremental multigroup cross sections for CANDU super cells with reactivity devices
International Nuclear Information System (INIS)
Balaceanu, V.; Constantin, M.
2001-01-01
In the last 20 years a multigroup methodology WIMS - PIJXYZ (WP) was developed and validated at INR Pitesti for obtaining incremental cross sections for reactivity devices in CANDU reactors. This is an alternate methodology to the CANDU classic methodology (experimentally adjusted) based on the POWDERPUFS and MULTICELL computer codes. The 2D supercell calculation performed with the WIMS code, that is a NEA Data Bank transport code, and which produces multigroup cross sections (on 18 energy groups) for CANDU supercell material (standard and perturbed, with and without reactivity devices). To obtain an as correct as possible 3D modelling for the CANDU supercells containing reactivity devices, the WIMS cross sections are used as input data for the PIJXYZ code, thus obtaining homogenized cross sections for CANDU supercells. PIJXYZ is an integral transport code based on the formalism of the first collision probabilities. It is analogue to the SHETAN code and it was created for neutron analyzes at cell level for CANDU type reactors were the reactivity devices are perpendicular to the fuel channels. The coordinate system used in PIJXYZ is a mixed one, namely a rectangular-cylindrical system. The geometric model used in PIJXYZ is presented. The fuel beam is represented by a horizontal cylinder and the reactivity device by a vertical one both cylinders being immersed in the moderator. Two supercell types were considered: a perturbed supercell (containing a reactivity device) and the standard supercell were the place of reactivity device is occupied by the moderator. The incremental cross sections for reactivity device are obtained as differences between the homogenized over supercell cross sections (with reactivity device) and homogenized over standards supercell (without device) cross sections. The PIJXYZ computation may be done on an energy cutting with 2 up to 18 groups. The validation of VIMS - PIJXYZ was done on the basis of several benchmark and by comparison with
Thermo-mechanical ratcheting in jointed rock masses
Pasten, C.; Garcí a, M.; Santamarina, Carlos
2015-01-01
Thermo-mechanical coupling takes place in jointed rock masses subjected to large thermal oscillations. Examples range from exposed surfaces under daily and seasonal thermal fluctuations to subsurface rock masses affected by engineered systems such as geothermal operations. Experimental, numerical and analytical results show that thermo-mechanical coupling can lead to wedging and ratcheting mechanisms that result in deformation accumulation when the rock mass is subjected to a biased static-force condition. Analytical and numerical models help in identifying the parameter domain where thermo-mechanical ratcheting can take place.
Thermo-driven microcrawlers fabricated via a microfluidic approach
International Nuclear Information System (INIS)
Wang Wei; Yao Chen; Zhang Maojie; Ju Xiaojie; Xie Rui; Chu Liangyin
2013-01-01
A novel thermo-driven microcrawler that can transform thermal stimuli into directional mechanical motion is developed by a simple microfluidic approach together with emulsion-template synthesis. The microcrawler is designed with a thermo-responsive poly(N-isopropylacrylamide) (PNIPAM) hydrogel body and a bell-like structure with an eccentric cavity. The asymmetric shrinking–swelling circulation of the microcrawlers enables a thermo-driven locomotion responding to repeated temperature changes, which provides a novel model with symmetry breaking principle for designing biomimetic soft microrobots. The microfluidic approach offers a novel and promising platform for design and fabrication of biomimetic soft microrobots. (paper)
Thermo-mechanical ratcheting in jointed rock masses
Pasten, C.
2015-09-01
Thermo-mechanical coupling takes place in jointed rock masses subjected to large thermal oscillations. Examples range from exposed surfaces under daily and seasonal thermal fluctuations to subsurface rock masses affected by engineered systems such as geothermal operations. Experimental, numerical and analytical results show that thermo-mechanical coupling can lead to wedging and ratcheting mechanisms that result in deformation accumulation when the rock mass is subjected to a biased static-force condition. Analytical and numerical models help in identifying the parameter domain where thermo-mechanical ratcheting can take place.
Thermo-cleavable polymers: Materials with enhanced photochemical stability
DEFF Research Database (Denmark)
Manceau, Matthieu; Petersen, Martin Helgesen; Krebs, Frederik C
2010-01-01
Photochemical stability of three thermo-cleavable polymers was investigated as thin films under atmospheric conditions. A significant increase in lifetime was observed once the side-chain was cleaved emphasizing the detrimental effect of solubilizing groups on the photochemical stability of conju......Photochemical stability of three thermo-cleavable polymers was investigated as thin films under atmospheric conditions. A significant increase in lifetime was observed once the side-chain was cleaved emphasizing the detrimental effect of solubilizing groups on the photochemical stability...... of conjugated polymers. In addition to their ease of processing, thermo-cleavable polymers thus also offer a greater intrinsic stability under illumination....
International Nuclear Information System (INIS)
Wang Xuan; Chen Xiaofei; Dong Weihua
2007-01-01
Objective: To evaluate the clinical efficacy of thermo-chemotherapy and thermo-lipiodol embolization in treatment of primary hepatocellular carcinoma(PHC). Methods: One hundred and sixteen cases of PHC were divided into three groups. Group A (38 cases)was treated with normal temperature chemotherapy and normal temperature lipiodol, Group B(40 cases)with thermo-chemotherapy and normal temperature lipiodol and group C (38 cases)with thermo-chemotherapy and thermo-lipiodol. Group B and group C were called the thermotherapy group. Results: In the thermotherapy groups, the rates of tumor size reduction were significantly greater than those in the normal group. There were no significant different in the hepatic function tests among the three groups. The 6-, 12-, 18-, and 24- month survival rates of the normal group and thermotherapy groups were 97%, 58%, 39% and 18%, versus 99%, 79%, 57% and 36%, respectively. No significant differences were found in the rates of reduction of tumor size and survival rates between group B and group C. Conclusion: Thermo-chemotherapy and thermo-embolization possess significant effect on PHC but without conspicuous damage to liver function. (authors)
Acute Genotoxic Effects of Effluent Water of Thermo-Power Plant “Kosova” In Tradescantia Pallida
Directory of Open Access Journals (Sweden)
I. R. Elezaj, L.B.Millaku, R.H. Imeri-Millaku, Q.I. Selimi, and K. Rr. Letaj
2011-09-01
Full Text Available The aim of this study was the evaluation of acute genotoxic effect of effluent water of thermo-power plant by means of Tradescantia root tips micronucleus test (MN, mitotic index and cell aberrations. Tradescantia, was experimentally treated (for 24 h, with effluent water of thermo-power plant in different dilution ratios (negative control – distilled water; primary untreated effluent water and 1:1; 1:2; 1:3; 1:4; 1:5; 1:6 and 1:7 respectively. Number of aberrant cells, and frequency of micronuclei (MN, in meristematic root tip cells of treated plants (Tradescantia, were significantly increased (P<0.001; P<0.001 respectively, while the mitotic index in all treated plants was progressively decreased in comparison to the negative control. The results of present study indicate that Tradescantia root-tip micronucleus assay with direct exposure of intact plants is an appropriate method which enables to detect genotoxic effects of effluent waters.
Development of multi-group xs libraries for the gfr 2400 reactor
International Nuclear Information System (INIS)
Cerba, Š.; Vrban, B.; Lüley, J.; Necas, V.
2016-01-01
GFR 2400 is considered as a conceptual design of the large scale GEN IV Gas-Cooled Fast Reactor. In general, the GEN IV technologies are seen as reliable but also very challenging reactor concepts. Since GFR 2400 lacks any experimental data, the questions on its safety are even more complex and the assessment of its performance could be made only based on computational experience. The paper deals with the development process of multi-group XS libraries based on a hybrid deterministic-Stochastic methodology, using the NJOY99, TRANSX, DIF3D, PARTISN and MCNP5 codes. A new optimized 25 group SBJ E 71 2 5G cross section library was developed based on ENDF/B-VII.1 evaluated data, ZZ-KAFAX-E70 background cross sections and GFR 2400 neutron spectrum. The created library was validated through integral experiments evaluated on the HEX-Z deterministic models in DIF3D. The results were also compared with MCNP5 calculations. (authors)
A consistent multigroup model for radiative transfer and its underlying mean opacities
International Nuclear Information System (INIS)
Turpault, Rodolphe
2005-01-01
In some regimes, such as in plasma physics or in super orbital atmospheric entry of space objects, the effects of radiation are crucial and can tremendously modify the hydrodynamics of the gas. In such cases, it is therefore important to have a good prediction of the radiative variables. However, full transport solutions of these multi-dimensional, time-dependent problems are too expensive to get to be involved in a coupled configuration. It is hence necessary to develop other models for radiation that are cheap, yet accurate enough to give good predictions of the radiative effects. We will herein introduce the multigroup-M1 model and look at its characteristics and in particular try to separate the angular error from the frequential one since these two approximation play very different roles. The angular behaviour of the model will be tested on a case proposed by Su and Olson and used by Olson et al. to compare various moments and (flux-limited) diffusion models. For the frequency behaviour, we use a simplified flame test-case and show the importance of taking good mean opacities
Offensive Strategy in the 2D Soccer Simulation League Using Multi-Group Ant Colony Optimization
Directory of Open Access Journals (Sweden)
Shengbing Chen
2016-02-01
Full Text Available The 2D soccer simulation league is one of the best test beds for the research of artificial intelligence (AI. It has achieved great successes in the domain of multi-agent cooperation and machine learning. However, the problem of integral offensive strategy has not been solved because of the dynamic and unpredictable nature of the environment. In this paper, we present a novel offensive strategy based on multi-group ant colony optimization (MACO-OS. The strategy uses the pheromone evaporation mechanism to count the preference value of each attack action in different environments, and saves the values of success rate and preference in an attack information tree in the background. The decision module of the attacker then selects the best attack action according to the preference value. The MACO-OS approach has been successfully implemented in our 2D soccer simulation team in RoboCup competitions. The experimental results have indicated that the agents developed with this strategy, along with related techniques, delivered outstanding performances.
Stability analysis of multi-group deterministic and stochastic epidemic models with vaccination rate
International Nuclear Information System (INIS)
Wang Zhi-Gang; Gao Rui-Mei; Fan Xiao-Ming; Han Qi-Xing
2014-01-01
We discuss in this paper a deterministic multi-group MSIR epidemic model with a vaccination rate, the basic reproduction number ℛ 0 , a key parameter in epidemiology, is a threshold which determines the persistence or extinction of the disease. By using Lyapunov function techniques, we show if ℛ 0 is greater than 1 and the deterministic model obeys some conditions, then the disease will prevail, the infective persists and the endemic state is asymptotically stable in a feasible region. If ℛ 0 is less than or equal to 1, then the infective disappear so the disease dies out. In addition, stochastic noises around the endemic equilibrium will be added to the deterministic MSIR model in order that the deterministic model is extended to a system of stochastic ordinary differential equations. In the stochastic version, we carry out a detailed analysis on the asymptotic behavior of the stochastic model. In addition, regarding the value of ℛ 0 , when the stochastic system obeys some conditions and ℛ 0 is greater than 1, we deduce the stochastic system is stochastically asymptotically stable. Finally, the deterministic and stochastic model dynamics are illustrated through computer simulations. (general)
Multi-Group Reductions of LTE Air Plasma Radiative Transfer in Cylindrical Geometries
Scoggins, James; Magin, Thierry Edouard Bertran; Wray, Alan; Mansour, Nagi N.
2013-01-01
Air plasma radiation in Local Thermodynamic Equilibrium (LTE) within cylindrical geometries is studied with an application towards modeling the radiative transfer inside arc-constrictors, a central component of constricted-arc arc jets. A detailed database of spectral absorption coefficients for LTE air is formulated using the NEQAIR code developed at NASA Ames Research Center. The database stores calculated absorption coefficients for 1,051,755 wavelengths between 0.04 µm and 200 µm over a wide temperature (500K to 15 000K) and pressure (0.1 atm to 10.0 atm) range. The multi-group method for spectral reduction is studied by generating a range of reductions including pure binning and banding reductions from the detailed absorption coefficient database. The accuracy of each reduction is compared to line-by-line calculations for cylindrical temperature profiles resembling typical profiles found in arc-constrictors. It is found that a reduction of only 1000 groups is sufficient to accurately model the LTE air radiation over a large temperature and pressure range. In addition to the reduction comparison, the cylindrical-slab formulation is compared with the finite-volume method for the numerical integration of the radiative flux inside cylinders with varying length. It is determined that cylindrical-slabs can be used to accurately model most arc-constrictors due to their high length to radius ratios.
Study on the Control Strategy of Shifting Time Involving Multigroup Clutches
Directory of Open Access Journals (Sweden)
Zhen Zhu
2016-01-01
Full Text Available This paper focuses on the control strategy of shifting time involving multigroup clutches for a hydromechanical continuously variable transmission (HMCVT. The dynamic analyses of mathematical models are presented in this paper, and the simulation models are used to study the control strategy of HMCVT. Simulations are performed in Simulation X platform to investigate the shifting time of clutches under different operating conditions. On this basis, simulation analysis and test verification of two typical conditions, which play the decisive roles for the shifting quality, are carried out. The results show that there are differences in the shifting time of the two typical conditions. In the shifting process from the negative transmission of hydromechanical ranges to the positive transmission of hydromechanical ranges, the control strategy based on the shifting time is switching the clutches of shifting mechanism firstly and then disengaging a group of clutches of planetary gear mechanism and engaging another group of the clutches of planetary gear mechanism lastly. In the shifting process from the hydraulic range to the hydromechanical range, the control strategy based on the shifting time is switching the clutches of hydraulic shifting mechanism and planetary gear mechanism at first and then engaging the clutch of shifting mechanism.
International Nuclear Information System (INIS)
Jones, D.B.
1986-01-01
EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated
JSD1000: multi-group cross section sets for shielding materials
International Nuclear Information System (INIS)
Yamano, Naoki
1984-03-01
A multi-group cross section library for shielding safety analysis has been produced by using ENDF/B-IV. The library consists of ultra-fine group cross sections, fine-group cross sections, secondary gamma-ray production cross sections and effective macroscopic cross sections for typical shielding materials. Temperature dependent data at 300, 560 and 900 K have been also provided. Angular distributions of the group to group transfer cross section are defined by a new method of ''Direct Angular Representation'' (DAR) instead of the method of finite Legendre expansion. The library designated JSD1000 are stored in a direct access data base named DATA-POOL and data manipulations are available by using the DATA-POOL access package. The 3824 neutron group data of the ultra-fine group cross sections and the 100 neutron, 20 photon group cross sections are applicable to shielding safety analyses of nuclear facilities. This report provides detailed specifications and the access method for the JSD1000 library. (author)
Three-dimensional h-adaptivity for the multigroup neutron diffusion equations
Wang, Yaqi
2009-04-01
Adaptive mesh refinement (AMR) has been shown to allow solving partial differential equations to significantly higher accuracy at reduced numerical cost. This paper presents a state-of-the-art AMR algorithm applied to the multigroup neutron diffusion equation for reactor applications. In order to follow the physics closely, energy group-dependent meshes are employed. We present a novel algorithm for assembling the terms coupling shape functions from different meshes and show how it can be made efficient by deriving all meshes from a common coarse mesh by hierarchic refinement. Our methods are formulated using conforming finite elements of any order, for any number of energy groups. The spatial error distribution is assessed with a generalization of an error estimator originally derived for the Poisson equation. Our implementation of this algorithm is based on the widely used Open Source adaptive finite element library deal.II and is made available as part of this library\\'s extensively documented tutorial. We illustrate our methods with results for 2-D and 3-D reactor simulations using 2 and 7 energy groups, and using conforming finite elements of polynomial degree up to 6. © 2008 Elsevier Ltd. All rights reserved.
MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1
International Nuclear Information System (INIS)
Conlin, Jeremy Lloyd; Parsons, Donald Kent; Gardiner, Steven J.; Gray, Mark Girard; Lee, Mary Beth; White, Morgan Curtis
2015-01-01
A new multi-group neutron cross section library has been released along with the release of NDI version 2.0.20. The library is named MENDF71x and is based upon the evaluations released in ENDF/B-VII.1 which was made publicly available in December 2011. ENDF/B-VII.1 consists of 423 evaluations of which ten are excited states evaluations and 413 are ground state evaluations. MENDF71x was created by processing the 423 evaluations into 618-group, downscatter only NDI data tables. The ENDF/B evaluation files were processed using NJOY version 99.393 with the exception of 35 Cl and 233 U. Those two isotopes had unique properties that required that we process the evaluation using NJOY version 2012. The MENDF71x library was only processed to room temperature, i.e., 293.6 K. In the future, we plan on producing a multi-temperature library based on ENDF/B-VII.1 and compatible with MENDF71x.
General solution of the multigroup spherical harmonics equations in R-Z geometry
International Nuclear Information System (INIS)
Matausek, M.
1983-01-01
In the present paper the generalization is performed of the procedure to solve multigroup spherical harmonics equations, which has originally been proposed and developed foe one-dimensional systems in cylindrical or spherical geometry, and later extended for special case of a two-dimensional system in r-z geometry. The expressions are derived for the axial and the radial dependence of the group values of the neutron flux moments, in the P-3 approximation of the spherical harmonics method, in a cylindrically symmetrical system with an arbitrary number of material regions in both r and z directions. In the special case of an axially homogeneous system, these expressions reduce to the relations derived previously. The analysis is performed of the possibilities to satisfy the boundary conditions in the case when the system considered represents an elementary reactor lattice cell and in the case when the system represents a reactor as a whole. The computational effort is estimated for system of a given configuration. (author)
Young Adults’ Attitude Towards Advertising: a multi-group analysis by ethnicity
Directory of Open Access Journals (Sweden)
Hiram Ting
2015-08-01
Full Text Available Objective – This study aims to investigate the attitude of Malaysian young adults towards advertising. How this segment responds to advertising, and how ethnic/cultural differences moderate are assessed. Design/methodology/approach – A quantitative questionnaire is used to collect data at two universities. Purposive sampling technique is adopted to ensure the sample represents the actual population. Structural equation modelling (SEM and multi-group analysis (MGA are utilized in analysis. Findings - The findings show that product information, hedonism, and good for economy are significant predictors of attitude towards advertising among young adults. Additionally, falsity is found to be significant among the Chinese, while social role and materialism among the Dayaks. No difference is observed in the effect of attitude on intention towards advertising by ethnicity. While homogeneity in advertising beliefs is assumed across ethnic groups, the Chinese and Dayak young adults are different in some of their advertising beliefs. Practical implications – Despite cultural effect being well-documented, young adults today seem to have similar beliefs and attitude towards advertising. Knowing what is shared and what is not for this segment is essential. Hence, it is imperative to keep track of their values in diversified communities to ensure effective communication process in advertising. Originality/value – In addition to the theory of reasoned action, MGA is utilized to assess the moderating effect of ethnic/culture on the whole model. This affords a more comprehensive understanding on the subject matter in multi-ethnic and cultural countries.
Two-dimensional semi-analytic nodal method for multigroup pin power reconstruction
International Nuclear Information System (INIS)
Seung Gyou, Baek; Han Gyu, Joo; Un Chul, Lee
2007-01-01
A pin power reconstruction method applicable to multigroup problems involving square fuel assemblies is presented. The method is based on a two-dimensional semi-analytic nodal solution which consists of eight exponential terms and 13 polynomial terms. The 13 polynomial terms represent the particular solution obtained under the condition of a 2-dimensional 13 term source expansion. In order to achieve better approximation of the source distribution, the least square fitting method is employed. The 8 exponential terms represent a part of the analytically obtained homogeneous solution and the 8 coefficients are determined by imposing constraints on the 4 surface average currents and 4 corner point fluxes. The surface average currents determined from a transverse-integrated nodal solution are used directly whereas the corner point fluxes are determined during the course of the reconstruction by employing an iterative scheme that would realize the corner point balance condition. The outgoing current based corner point flux determination scheme is newly introduced. The accuracy of the proposed method is demonstrated with the L336C5 benchmark problem. (authors)
International Nuclear Information System (INIS)
Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.
1977-03-01
TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided
International Nuclear Information System (INIS)
Alpan, F. Arzu; Haghighat, Alireza
2008-01-01
Multigroup (i.e., broad-group) libraries play a significant role in the accuracy of transport calculations. There are several broad-group libraries available for particular applications. For example the 47-neutron (26 fast groups), 20-gamma-group BUGLE libraries are commonly used for light water reactor shielding and pressure vessel dosimetry problems. However, there is no publicly available methodology to construct group structures for a problem and objective of interest. Therefore, we have developed the Contribution and Point-wise Cross-Section Driven (CPXSD) methodology, which constructs effective fine-and broad-group structures. In this paper, we use the CPXSD methodology to construct broad-group structures for fast neutron dosimetry problems. It is demonstrated that the broad-group libraries generated from CPXSD constructed group structures, while only 14 groups (rather than 26 groups) in the fast energy range are in good agreement (similar to 1 %-2 %) with the fine-group library from which they were derived, in reaction rate calculations.
Review of uncertainty files and improved multigroup cross section files for FENDL
International Nuclear Information System (INIS)
Ganesan, S.
1994-03-01
The IAEA Nuclear Data Section, in co-operation with several national nuclear data centers and research groups, is creating an internationally available Fusion Evaluated Nuclear Data Library (FENDL), which will serve as a comprehensive source of processed and tested nuclear data tailored to the requirements of the Engineering and Development Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project and other fusion-related development projects. The FENDL project of the International Atomic Energy Agency has the task of coordination with the goal of assembling, processing and testing a comprehensive, fusion-relevant Fusion Evaluated Nuclear Data Library with unrestricted international distribution. The present report contains the summary of the IAEA Advisory Group Meeting on ''Review of Uncertainty Files and Improved Multigroup Cross Section Files for FENDL'', held during 8-12 November 1993 at the Tokai Research Establishment, JAERI, Japan, organized in cooperation with the Japan Atomic Energy Research Institute. The report presents the current status of the FENDL activity and the future work plans in the form of conclusions and recommendations of the four Working Groups of the Advisory Group Meeting on (1) experimental and calculational benchmarks, (2) preparation processed libraries for FENDL/ITER, (3) specifying procedures for improving FENDL and (4) selection of activation libraries for FENDL. (author). 1 tab
TASK, 1-D Multigroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron
International Nuclear Information System (INIS)
Buhl, A.R.; Hermann, O.W.; Hinton, R.J.; Dodds, H.L. Jr.; Robinson, J.C.; Lillie, R.A.
1975-01-01
1 - Description of problem or function: TASK solves the one-dimensional multigroup form of the reactor kinetics equations, using either transport or diffusion theory and allowing an arbitrary number of delayed neutron groups. The program can also be used to solve standard static problems efficiently such as eigenvalue problems, distributed source problems, and boundary source problems. Convergence problems associated with sources in highly multiplicative media are circumvented, and such problems are readily calculable. 2 - Method of solution: TASK employs a combination scattering and transfer matrix method to eliminate certain difficulties that arise in classical finite difference approximations. As such, within-group (inner) iterations are eliminated and solution convergence is independent of spatial mesh size. The time variable is removed by Laplace transformation. (A later version will permit direct time solutions.) The code can be run either in an outer iteration mode or in closed (non-iterative) form. The running mode is dictated by the number of groups times the number of angles, consistent with available storage. 3 - Restrictions on the complexity of the problem: The principal restrictions are available storage and computation time. Since the code is flexibly-dimensioned and has an outer iteration option there are no internal restrictions on group structure, quadrature, and number of ordinates. The flexible-dimensioning scheme allows optional use of core storage. The generalized cylindrical geometry option is not complete in Version I of the code. The feedback options and omega-mode search options are not included in Version I
Colaert, Niklaas; Barsnes, Harald; Vaudel, Marc; Helsens, Kenny; Timmerman, Evy; Sickmann, Albert; Gevaert, Kris; Martens, Lennart
2011-08-05
The Thermo Proteome Discoverer program integrates both peptide identification and quantification into a single workflow for peptide-centric proteomics. Furthermore, its close integration with Thermo mass spectrometers has made it increasingly popular in the field. Here, we present a Java library to parse the msf files that constitute the output of Proteome Discoverer. The parser is also implemented as a graphical user interface allowing convenient access to the information found in the msf files, and in Rover, a program to analyze and validate quantitative proteomics information. All code, binaries, and documentation is freely available at http://thermo-msf-parser.googlecode.com.
International Nuclear Information System (INIS)
Abo, Yohichi; Hagiya, Akiko; Naganuma, Takao; Tohkairin, Yukiko; Shiomi, Kunihiro; Kajiura, Zenta; Hachimori, Akira; Uchiumi, Toshio; Nakagaki, Masao
2004-01-01
We constructed an overexpression system for human ribosomal phosphoprotein P0, together with P1 and P2, which is crucially important for translation. Genes for these proteins, fused with the glutathione S-transferase (GST)-tag at the N-terminus, were inserted into baculovirus and introduced to insect cells. The fusion proteins, but not the proteins without the tag, were efficiently expressed into cells as soluble forms. The fusion protein GST.P0 as well as GST.P1/GST.P2 was phosphorylated in cells as detected by incorporation of 32 P and reactivity with monoclonal anti-phosphoserine antibody. GST.P0 expressed in insect cells, but not the protein obtained in Escherichia coli, had the ability to form a complex with P1 and P2 proteins and to bind to 28S rRNA. Moreover, the GST.P0-P1-P2 complex participated in high eEF-2-dependent GTPase activity. Baculovirus expression systems appear to provide recombinant human P0 samples that can be used for studies on the structure and function
Shinkai, Masayuki; Imano, Motohiro; Hiraki, Yoko; Kato, Hiroaki; Iwama, Mitsuru; Shiraishi, Osamu; Yasuda, Atsushi; Kimura, Yutaka; Imamoto, Haruhiko; Furukawa, Hiroshi; Yasuda, Takushi
2017-11-01
We evaluate the feasibility and efficacy of combination chemotherapy including single intraperitoneal( IP)administration of paclitaxel(PTX), followed by triplet chemotherapy(PTX, cisplatin[CDDP]and S-1: PCS)for CY1P0 gastric cancer. First of all, we performed staging laparoscopy and confirmed CY1P0, and secondary, administrated PTX intraperitoneally. Thirdly, patients received PCS chemotherapy for 2 courses. After antitumor effect had been confirmed, we performed second look laparoscopy. In the case of CY0P0, we performed gastrectomy with D2 lymph nodes dissection. Total 4 patients were enrolled. Grade 3 leukopenia and neutropenia were observed in one patient while intraperitoneal and systemic-chemotherapy. One patients showed PR and 3 patients showed SD. All patients underwent second look laparoscopy. CY0P0 was observed in all patients and gastrectomy with D2 dissection was performed for all patients. Postoperative complications were observed in 2 patients. Two patients were still alive without recurrence, while the remaining 2 had died of liver metastasis and #16 LN metastasis. Combination chemotherapy including single IP PTX followed by PCS systemic-chemotherapy for CY1P0 gastric cancer is feasible and efficient.
Wigner Function of Thermo-Invariant Coherent State
International Nuclear Information System (INIS)
Xue-Fen, Xu; Shi-Qun, Zhu
2008-01-01
By using the thermal Winger operator of thermo-field dynamics in the coherent thermal state |ξ) representation and the technique of integration within an ordered product of operators, the Wigner function of the thermo-invariant coherent state |z,ℵ> is derived. The nonclassical properties of state |z,ℵ> is discussed based on the negativity of the Wigner function. (general)
Library of neutron cross sections of the Thermos code
International Nuclear Information System (INIS)
Alonso V, G.; Hernandez L, H.
1991-10-01
The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)
An oral Na(V)1.8 blocker improves motor function in mice completely deficient of myelin protein P-0
DEFF Research Database (Denmark)
Rosberg, Mette R.; Alvarez Herrero, Susana; Krarup, Christian
2016-01-01
Mice deficient of myelin protein P0 are established models of demyelinating Charcot-Marie-Tooth (CMT) disease. Dysmyelination in these mice is associated with an ectopic expression of the sensory neuron specific sodium channel isoform NaV1.8 on motor axons. We reported that in P0+/−, a model of CMT......1B, the membrane dysfunction could be acutely improved by a novel oral NaV1.8 blocker referred to as Compound 31 (C31, Bioorg. Med. Chem. Lett. 2010, 20, 6812; AbbVie Inc.). The aim of this study was to investigate the extent to which C31 treatment could also improve the motor axon function in P0......-of-concept that treatment with oral subtype-selective NaV1.8 blockers could be used to improve the motor function in severe forms of demyelinating CMT....
Thermo-pneumatic canning; Le gainage thermopneumatique
Energy Technology Data Exchange (ETDEWEB)
Gauthron, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1958-07-01
In the thermo-pneumatic canning, the fuel is enclosed in its can with a clearance that must be reduced by external heated gas pressure. The principal applications are: a) binding magnesium cans on to uranium in fuel elements of reactors cooled by CO{sub 2} under pressure, b) application of a can to a hollow bar of uranium too thin to resist the pressure of cold hydraulic canning, c) application of an aluminium can to a bar, with an initial diametrical clearance between uranium and can too great to sustain cold hydraulic canning without buckling, d) detection of major leakage in the slugs. (author) [French] Ce procede consiste a appliquer une gaine sur une barre d'uranium par pression hydrostatique d'un gaz chaud. Les principales applications sont: a) le frettage des gaines de magnesium des elements combustibles des piles refroidies au CO{sub 2} sous pression, b) le gainage d'un barreau creux qui serait ecrase a froid, c) le gainage avec un jeu initial trop fort pour etre effectue a froid sans plisser, d) la detection des fuites de cartouches. (auteur)
Energy Technology Data Exchange (ETDEWEB)
Alonso V, G; Hernandez L, H [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)
1991-10-15
The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)
APPLE, Plot of 1-D Multigroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN
International Nuclear Information System (INIS)
Kawasaki, Hiromitsu; Seki, Yasushi
1983-01-01
A - Description of problem or function: The APPLE-2 code has the following functions: (1) It plots multi-group energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT-3.5, and MORSE. (2) It gives an overview plot of multi-group neutron fluxes calculated by ANISN and DOT-3.5. The scalar neutron flux phi(r,E) is plotted with the spatial parameter r linear along the Y-axis, logE along the X-axis and log phi(r,E) in the Z direction. (3) It calculates the spatial distribution and region volume integrated values of reaction rates using the scalar flux calculated with ANISN and DOT-3.5. (4) Reaction rate distribution along the R or Z direction may be plotted. (5) An overview plot of reaction rates or scalar fluxes summed over specified groups may be plotted. R(ri,zi) or phi(ri,zi) is plotted with spatial parameters r and z along the X- and Y-axes in an orthogonal coordinate system. (6) Angular flux calculated by ANISN is rearranged and a shell source at any specified spatial mesh point may be punched out in FIDO format. The shell source obtained may be employed in solving deep penetration problems with ANISN, when the entire reactor system is divided into two or more parts and the neutron fluxes in two adjoining parts are connected by using the shell source. B - Method of solution: (a) The input data specification is made as simple as possible by making use of the input data required in the radiation transport code. For example, geometry related data in ANISN and DOT are transmitted to APPLE-2 along with scalar flux data so as to reduce duplicity and errors in reproducing these data. (b) Most the input data follow the free form FIDO format developed at Oak Ridge National Laboratory and used in the ANISN code. Furthermore, the mixture specifying method used in ANISN is also employed by APPLE-2. (c) Libraries for some standard response functions required in fusion reactor design have been prepared and are made available to users of the 42-group neutron
Energy Technology Data Exchange (ETDEWEB)
Aggery, A
1999-12-01
The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)
Energy Technology Data Exchange (ETDEWEB)
Zanette, Rodrigo [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pós-Graduação em Matemática Aplicada; Petersen, Claudio Z.; Tavares, Matheus G., E-mail: rodrigozanette@hotmail.com, E-mail: claudiopetersen@yahoo.com.br, E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Programa de Pós-Graduação em Modelagem Matemática
2017-07-01
We describe in this work the application of the modified power method for solve the multigroup neutron diffusion eigenvalue problem in slab geometry considering two-dimensions for nuclear reactor global calculations. It is well known that criticality calculations can often be best approached by solving eigenvalue problems. The criticality in nuclear reactors physics plays a relevant role since establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve the eigenvalue problem, a modified power method is used to obtain the dominant eigenvalue (effective multiplication factor (K{sub eff})) and its corresponding eigenfunction (scalar neutron flux), which is non-negative in every domain, that is, physically relevant. The innovation of this work is solving the neutron diffusion equation in analytical form for each new iteration of the power method. For solve this problem we propose to apply the Finite Fourier Sine Transform on one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. The inverse Fourier transform is used to reconstruct the solution for the original problem. It is known that the power method is an iterative source method in which is updated by the neutron flux expression of previous iteration. Thus, for each new iteration, the neutron flux expression becomes larger and more complex due to analytical solution what makes propose that it be reconstructed through an polynomial interpolation. The methodology is implemented to solve a homogeneous problem and the results are compared with works presents in the literature. (author)
Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method
International Nuclear Information System (INIS)
Dunley, Leonardo Souza
2002-01-01
The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron
Symmetry breaking in the opinion dynamics of a multi-group project organization
International Nuclear Information System (INIS)
Zhu Zhen-Tao; Zhou Jing; Chen Xing-Guang; Li Ping
2012-01-01
A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces: (i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness. Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes, i.e., a deadlock regime, a convergence regime, and a bifurcation regime in opinion dynamics. The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to. In the case of a three-group project with a symmetric social network, both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord, instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p. 125 Fig. 5), project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations, which urges that apart from divergence in participants' interests, nonlinear interaction can also make conflict inevitable in the PO. The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO. It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO
Symmetry breaking in the opinion dynamics of a multi-group project organization
Zhu, Zhen-Tao; Zhou, Jing; Li, Ping; Chen, Xing-Guang
2012-10-01
A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces: (i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness. Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes, i.e., a deadlock regime, a convergence regime, and a bifurcation regime in opinion dynamics. The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to. In the case of a three-group project with a symmetric social network, both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord, instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p. 125 Fig. 5), project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations, which urges that apart from divergence in participants' interests, nonlinear interaction can also make conflict inevitable in the PO. The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO. It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO.
Reliability generalization of the Multigroup Ethnic Identity Measure-Revised (MEIM-R).
Herrington, Hayley M; Smith, Timothy B; Feinauer, Erika; Griner, Derek
2016-10-01
[Correction Notice: An Erratum for this article was reported in Vol 63(5) of Journal of Counseling Psychology (see record 2016-33161-001). The name of author Erika Feinauer was misspelled as Erika Feinhauer. All versions of this article have been corrected.] Individuals' strength of ethnic identity has been linked with multiple positive indicators, including academic achievement and overall psychological well-being. The measure researchers use most often to assess ethnic identity, the Multigroup Ethnic Identity Measure (MEIM), underwent substantial revision in 2007. To inform scholars investigating ethnic identity, we performed a reliability generalization analysis on data from the revised version (MEIM-R) and compared it with data from the original MEIM. Random-effects weighted models evaluated internal consistency coefficients (Cronbach's alpha). Reliability coefficients for the MEIM-R averaged α = .88 across 37 samples, a statistically significant increase over the average of α = .84 for the MEIM across 75 studies. Reliability coefficients for the MEIM-R did not differ across study and participant characteristics such as sample gender and ethnic composition. However, consistently lower reliability coefficients averaging α = .81 were found among participants with low levels of education, suggesting that greater attention to data reliability is warranted when evaluating the ethnic identity of individuals such as middle-school students. Future research will be needed to ascertain whether data with other measures of aspects of personal identity (e.g., racial identity, gender identity) also differ as a function of participant level of education and associated cognitive or maturation processes. (PsycINFO Database Record (c) 2016 APA, all rights reserved).
Optimization of multi-group cross sections for fast reactor analysis
International Nuclear Information System (INIS)
Chin, M. R.; Manalo, K. L.; Edgar, C. A.; Paul, J. N.; Molinar, M. P.; Redd, E. M.; Yi, C.; Sjoden, G. E.
2013-01-01
The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO 2 -UO 2 with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)
PUFF-III: A Code for Processing ENDF Uncertainty Data Into Multigroup Covariance Matrices
International Nuclear Information System (INIS)
Dunn, M.E.
2000-01-01
PUFF-III is an extension of the previous PUFF-II code that was developed in the 1970s and early 1980s. The PUFF codes process the Evaluated Nuclear Data File (ENDF) covariance data and generate multigroup covariance matrices on a user-specified energy grid structure. Unlike its predecessor, PUFF-III can process the new ENDF/B-VI data formats. In particular, PUFF-III has the capability to process the spontaneous fission covariances for fission neutron multiplicity. With regard to the covariance data in File 33 of the ENDF system, PUFF-III has the capability to process short-range variance formats, as well as the lumped reaction covariance data formats that were introduced in ENDF/B-V. In addition to the new ENDF formats, a new directory feature is now available that allows the user to obtain a detailed directory of the uncertainty information in the data files without visually inspecting the ENDF data. Following the correlation matrix calculation, PUFF-III also evaluates the eigenvalues of each correlation matrix and tests each matrix for positive definiteness. Additional new features are discussed in the manual. PUFF-III has been developed for implementation in the AMPX code system, and several modifications were incorporated to improve memory allocation tasks and input/output operations. Consequently, the resulting code has a structure that is similar to other modules in the AMPX code system. With the release of PUFF-III, a new and improved covariance processing code is available to process ENDF covariance formats through Version VI
Electroproduction of p0 mesons
International Nuclear Information System (INIS)
Cohen, I.; Erickson, R.; Messing, F.; Nordberg, E.; Siemann, R.; Smith-Kintner, J.; Stein, P.; Sadoff, A.; Drews, G.; Gebert, W.
1980-03-01
Cross-sections for rho 0 electroproduction measured in a streamer chamber experiment are separated into elastic (ep → eprho 0 ) and inelastic production channels. For the inelastic channel 1/sigma dsigma/dz, 1/sigma dsigma/dp 2 sub(t) and a density matrix element are shown and compared to quark-parton model predictions. The ratio of rho 0 to π 0 production is measured to be 2.0 +- 0.5. For the elastic channel, the total cross-section and t-dependence are presented. (orig.)
Improved wavelengths for the 1s2s3S1-1s2p3P0,2 transitions in helium-like Si12+
International Nuclear Information System (INIS)
Armour, I.A.; Myers, E.G.; Silver, J.D.; Traebert, E.; Oxford Univ.
1979-01-01
The wavelengths of the 1s2s 3 S 1 -1s2p 3 P 0 , 2 transitions in He-like Si 12+ have been remaesured to be 87.86 +- 0.01 nm and 81.48 +- 0.01 nm. The use of Rydberg lines for the calibration of fast beam spectra is discussed. (orig.)
Vliegenthart, J.F.G.; Gutiérrez Gallego, R.; Jiménez Blanco, J.L.; Thijssen-van Zuylen, C.W.E.M.; Gotfredsen, C.H.; Voshol, H.; Duus, J.Ø.; Schachner, M.
2001-01-01
The carbohydrate structures present on the glycoproteins in the central and peripheral nerve systems are essential in many cell adhesion processes. The P0 glycoprotein, expressed by myelinating Schwann cells, plays an important role during the formation and maintenance of myelin, and it is the most
Background/Introduction. PrPC is highly conserved among mammals, but its natural function is unclear. Prnp ablated mice (PrP0/0) appear to develop normally and are able to reproduce. These observations seem to indicate that the gene is not essential for viability, in spite of it being highly conse...
Csorba, Tibor; Lózsa, Rita; Hutvágner, György; Burgyán, József
2010-05-01
RNA silencing plays an important role in plants in defence against viruses. To overcome this defence, plant viruses encode suppressors of RNA silencing. The most common mode of silencing suppression is sequestration of double-stranded RNAs involved in the antiviral silencing pathways. Viral suppressors can also overcome silencing responses through protein-protein interaction. The poleroviral P0 silencing suppressor protein targets ARGONAUTE (AGO) proteins for degradation. AGO proteins are the core component of the RNA-induced silencing complex (RISC). We found that P0 does not interfere with the slicer activity of pre-programmed siRNA/miRNA containing AGO1, but prevents de novo formation of siRNA/miRNA containing AGO1. We show that the AGO1 protein is part of a high-molecular-weight complex, suggesting the existence of a multi-protein RISC in plants. We propose that P0 prevents RISC assembly by interacting with one of its protein components, thus inhibiting formation of siRNA/miRNA-RISC, and ultimately leading to AGO1 degradation. Our findings also suggest that siRNAs enhance the stability of co-expressed AGO1 in both the presence and absence of P0.
International Nuclear Information System (INIS)
Ragusa, J. C.
2004-01-01
In this paper, a method for performing spatially adaptive computations in the framework of multigroup diffusion on 2-D and 3-D Cartesian grids is investigated. The numerical error, intrinsic to any computer simulation of physical phenomena, is monitored through an a posteriori error estimator. In a posteriori analysis, the computed solution itself is used to assess the accuracy. By efficiently estimating the spatial error, the entire computational process is controlled through successively adapted grids. Our analysis is based on a finite element solution of the diffusion equation. Bilinear test functions are used. The derived a posteriori error estimator is therefore based on the Hessian of the numerical solution. (authors)
International Nuclear Information System (INIS)
Prati, A.; Anaf, J.
1988-09-01
The IBM version of the multigroup diffusion code 2DB was implemented in the IEAv CDC CYBER 170/750 system. It was optimized relative to the use of the central memory, limited to 132 K-words, through the memory manager CMM and its partition into three source codes: rectangular and cylindrical geometries, triangular geometry and hexagonal geometry. The reactangular, triangular and hexagonal geometry nodal options were revised and optimized. A fast reactor and a PWR type thermal reactor sample cases were studied. The results are presented and analized. An updated 2DB code user's manual was written in Portugueses and published separately. (author) [pt
International Nuclear Information System (INIS)
Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.
1986-01-01
For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig
International Nuclear Information System (INIS)
Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.
1986-02-01
Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10 B, 11 B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs
Theory and modeling of cylindrical thermo-acoustic transduction
Energy Technology Data Exchange (ETDEWEB)
Tong, Lihong, E-mail: lhtong@ecjtu.edu.cn [School of Civil Engineering and Architecture, East China Jiaotong University, Nanchang, Jiangxi (China); Lim, C.W. [Department of Architecture and Civil Engineering, City University of Hong Kong, Kowloon, Hong Kong SAR (China); Zhao, Xiushao; Geng, Daxing [School of Civil Engineering and Architecture, East China Jiaotong University, Nanchang, Jiangxi (China)
2016-06-03
Models both for solid and thinfilm-solid cylindrical thermo-acoustic transductions are proposed and the corresponding acoustic pressure solutions are obtained. The acoustic pressure for an individual carbon nanotube (CNT) as a function of input power is investigated analytically and it is verified by comparing with the published experimental data. Further numerical analysis on the acoustic pressure response and characteristics for varying input frequency and distance are also examined both for solid and thinfilm-solid cylindrical thermo-acoustic transductions. Through detailed theoretical and numerical studies on the acoustic pressure solution for thinfilm-solid cylindrical transduction, it is concluded that a solid with smaller thermal conductivity favors to improve the acoustic performance. In general, the proposed models are applicable to a variety of cylindrical thermo-acoustic devices performing in different gaseous media. - Highlights: • Theory and modeling both for solid and thinfilm-solid cylindrical thermo-acoustic transductions are proposed. • The modeling is verified by comparing with the published experimental data. • Acoustic response characteristics of cylindrical thermo-acoustic transductions are predicted by the proposed model.
Energy Technology Data Exchange (ETDEWEB)
Bore, C; Dandeu, Y; Saint-Amand, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1965-07-01
MUDE is a nuclear code written in FORTRAN II for IBM 7090-7094. It resolves a system of difference equations approximating to the one-dimensional multigroup neutron scattering problem. More precisely, this code makes it possible to: 1. Calculate the critical condition of a reactor (k{sub eff}, critical radius, critical composition) and the corresponding fluxes; 2. Calculate the associated fluxes and various subsidiary results; 3. Carry out perturbation calculations; 4. Study the propagation of fluxes at a distance; 5. Estimate the relative contributions of the cross sections (macroscopic or microscopic); 6. Study the changes with time of the composition of the reactor. (authors) [French] MUDE est un code nucleaire ecrit en FORTRAN II pour IBM 7090-7094. Il resout un systeme d'equations aux differences approchant le probleme de diffusion neutronique multigroupe a une dimension. Plus precisement ce code permet de: 1. Calculer la condition critique d'un reacteur (k{sub eff}, rayon critique, composition critique) et les flux correspondants; 2. Calculer les flux adjoints et divers resultats connexes; 3. Effectuer des calculs de perturbation; 4. Etudier la propagation des flux a longue distance; 5. Ponderer des sections efficaces (macroscopiques ou microscopiques); 6. Etudier l'evolution de la composition du reacteur au cours du temps. (auteurs)
Sensitivity of 238U resonance absorption to library multigroup structure as calculated by WIMS-AECL
International Nuclear Information System (INIS)
Laughton, P.J.; Donnelly, J.V.
1995-01-01
In simulations of the TRX-1 experimental lattice, WIMS-AECL overpredicts, relative to MCNP, resonance absorption in neutron-energy groups containing the three large, low-lying resonances of 238 U when a standard ENDF/B-V-based library is used. A total excess in these groups of 4.0 neutron captures by 238 U per thousand fission neutrons has been observed. Similar comparisons are made in this work for the MIT-4 experimental lattice and simplified CANDU lattice cells containing 37-element fuel, with and without heavy-water coolant. Eleven different 89-group cross-section libraries were constructed for WIMS-AECL from ENDF/B-V data: only the neutron-energy-group boundaries used in generating multigroup cross sections and the Goldstein-Cohen correction factors differ from one library to the next. The first library uses the original 89-group structure, and the other ten involve energy groups of varying widths centred on the three large, low-lying resonances of 238 U. For TRX-1, some reduction in total discrepancy in 238 U capture can be achieved by using a new structure, although the improvement is small. The discrepancies in 238 U capture are of the same order for the MIT-4 case as those observed for TRX-1 for both the original group structure and the ten new structures. The WIMS-AECL calculation of 238 U resonance absorption in the same ranges of energy for the simplified CANDU 37-element lattice are in better agreement with MCNP than they are for TRX-1 and MIT-4: when the original structure is used, WIMS-AECL underpredicts total capture rate by 238 U in the energy range of interest by only 0.56 per thousand fission neutrons (coolant present) and 0.88 per thousand fission neutrons (voided coolant channel). The discrepancies are reduced when some of the new structures are used. For almost all of the cases considered here-TRX-1, MIT-4 and CANDU with coolant-better group-by-group agreement of 238 U capture around the 6.67-eV resonance is achieved by using a new library
Analytical Expressions for Thermo-Osmotic Permeability of Clays
Gonçalvès, J.; Ji Yu, C.; Matray, J.-M.; Tremosa, J.
2018-01-01
In this study, a new formulation for the thermo-osmotic permeability of natural pore solutions containing monovalent and divalent cations is proposed. The mathematical formulation proposed here is based on the theoretical framework supporting thermo-osmosis which relies on water structure alteration in the pore space of surface-charged materials caused by solid-fluid electrochemical interactions. The ionic content balancing the surface charge of clay minerals causes a disruption in the hydrogen bond network when more structured water is present at the clay surface. Analytical expressions based on our heuristic model are proposed and compared to the available data for NaCl solutions. It is shown that the introduction of divalent cations reduces the thermo-osmotic permeability by one third compared to the monovalent case. The analytical expressions provided here can be used to advantage for safety calculations in deep underground nuclear waste repositories.
Athermalization of resonant optical devices via thermo-mechanical feedback
Rakich, Peter; Nielson, Gregory N.; Lentine, Anthony L.
2016-01-19
A passively athermal photonic system including a photonic circuit having a substrate and an optical cavity defined on the substrate, and passive temperature-responsive provisions for inducing strain in the optical cavity of the photonic circuit to compensate for a thermo-optic effect resulting from a temperature change in the optical cavity of the photonic circuit. Also disclosed is a method of passively compensating for a temperature dependent thermo-optic effect resulting on an optical cavity of a photonic circuit including the step of passively inducing strain in the optical cavity as a function of a temperature change of the optical cavity thereby producing an elasto-optic effect in the optical cavity to compensate for the thermo-optic effect resulting on an optical cavity due to the temperature change.
Fiber Optic Thermo-Hygrometers for Soil Moisture Monitoring.
Leone, Marco; Principe, Sofia; Consales, Marco; Parente, Roberto; Laudati, Armando; Caliro, Stefano; Cutolo, Antonello; Cusano, Andrea
2017-06-20
This work deals with the fabrication, prototyping, and experimental validation of a fiber optic thermo-hygrometer-based soil moisture sensor, useful for rainfall-induced landslide prevention applications. In particular, we recently proposed a new generation of fiber Bragg grating (FBGs)-based soil moisture sensors for irrigation purposes. This device was realized by integrating, inside a customized aluminum protection package, a FBG thermo-hygrometer with a polymer micro-porous membrane. Here, we first verify the limitations, in terms of the volumetric water content (VWC) measuring range, of this first version of the soil moisture sensor for its exploitation in landslide prevention applications. Successively, we present the development, prototyping, and experimental validation of a novel, optimized version of a soil VWC sensor, still based on a FBG thermo-hygrometer, but able to reliably monitor, continuously and in real-time, VWC values up to 37% when buried in the soil.
Ionization of small molecules by state-selected neon (3P0, 3P2) metastable atoms in the 0.06
Berg, van den F.T.M.; Schonenberg, J.H.M.; Beijerinck, H.C.W.
1987-01-01
The velocity dependence and absolute values of the total ionisation cross section for the molecules H2, N2, O2, NO, CO, N2O, CO2, and CH4 by metastable Ne* (3P0) and Ne* (3P2) atoms at collision energies ranging from 0.06 to 6.0 eV have been measured in a crossed beam experiment. State selection of
Measurements of nitrogen depth distribution in the surface of steel with the 14N(d,p0)15N reaction
International Nuclear Information System (INIS)
Didriksson, R.; Goenczi, L.; Sundqvist, B.
1980-01-01
The 14 N(d,p 0 ) 15 N nuclear reaction has been used to measure the nitrogen depth distribution in the surface of steel samples. With a beam energy of 2.5 MeV a depth of 15μm could be analyzed. The depth resolution was 0.7 μm (FWHM) and nitrogen contents down to 0.02 percent could be determined. (author)
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.
1988-11-01
We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)
International Nuclear Information System (INIS)
Kasselmann, S.; Druska, C.; Lauer, A.
2010-01-01
The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)
Solution of multi-group diffusion equation in x-y-z geometry by finite Fourier transformation
International Nuclear Information System (INIS)
Kobayashi, Keisuke
1975-01-01
The multi-group diffusion equation in three-dimensional x-y-z geometry is solved by finite Fourier transformation. Applying the Fourier transformation to a finite region with constant nuclear cross sections, the fluxes and currents at the material boundaries are obtained in terms of the Fourier series. Truncating the series after the first term, and assuming that the source term is piecewise linear within each mesh box, a set of coupled equations is obtained in the form of three-point equations for each coordinate. These equations can be easily solved by the alternative direction implicit method. Thus a practical procedure is established that could be applied to replace the currently used difference equation. This equation is used to solve the multi-group diffusion equation by means of the source iteration method; and sample calculations for thermal and fast reactors show that the present method yields accurate results with a smaller number of mesh points than the usual finite difference equations. (auth.)
Peng, Wei-Ren; Lin, Wen-Piao; Chi, Sien
2006-03-01
The authors propose a novel frequency-overlapping multigroup scheme for a passive all-optical fast-frequency hopped code-division multiple-access (OFFH-CDMA) system based on fiber Bragg grating array (FBGA). In the conventional scheme, the users are assigned those codes constructed on the nonoverlapping frequency slots, and therefore the bandgaps between the adjacent gratings are wasted. To make a more efficient use of the optical spectrum, the proposed scheme divided the users into several groups, and assigned the codes, which interleaved to each other to the different groups. In addition to the higher utilization of the spectrum, the interleaved nature of the frequency allocations of different groups will make the groups less correlated and, hence, lower the multiple-access interference (MAI). The corresponding codeset and its constraints for this new scheme are also developed and analyzed. The performance of the system in terms of the correlation functions and bit error rate (BER) are given in both the conventional and the proposed schemes. The numerical results show that, with the multigroup scheme, performance is much improved compared to the conventional scheme.
International Nuclear Information System (INIS)
Thiyagarajan, T.K.; Ganesan, S.; Jagannathan, V.; Karthikeyan, R.
2002-01-01
As a result of the IAEA Co-ordinated Research Programme entitled 'Final Stage of the WIMS Library Update Project', new and updated WIMS-D libraries based upon ENDF/B-VI.5, JENDL-3.2 and JEF-2.2 have become available. A project to prepare an exhaustive handbook of WIMS-D cross sections from old and new libraries has been taken up by the authors. As part of this project, we have developed a computer program XnWlup with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualization of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. The current features of the software, on-line help manual and future plans for further development are described in this paper
Directory of Open Access Journals (Sweden)
Shane Stimpson
2017-09-01
Full Text Available An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1 a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications Progression Problem 2a and (2 a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Given these performance benefits, these approaches have been adopted as the default in MPACT.
International Nuclear Information System (INIS)
Stimpson, Shane G.; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.
2017-01-01
An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC) is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1) a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications) Progression Problem 2a and (2) a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Furthermore given these performance benefits, these approaches have been adopted as the default in MPACT.
DEFF Research Database (Denmark)
Lehnhoff, Janna; Moldovan, Mihai; Hedegaard, Anne
2014-01-01
Mice deficient for the peripheral myelin binding protein zero (P0-/-) show a progressive dysmyelinating neuropathy phenotypically resembling severe forms of Charcot-Marie-Tooth (CMT) disease. Traditionally, the progression of the disease was attributed to axonal loss, but the effect of chronic...... dysmyelination remains poorly understood. In this study, in vivo electrophysiological recordings were used to assess the function of both central and axonal components of spinal lumbar motoneurones in adult P0-/- mice.Three month old P0-/- mice (n=7) and wild type (WT) littermate controls (n=5) were...... anaesthetized with Hypnorm (0.315 mg/mL fentanyl-citrate + 10 mg/mL fluanisone), Midazolam (5 mg/mL), and sterile water, mixed in the ratio 1:1:2 (induction: 0.15mL/25g, maintenance: 0.05 mL/20 minutes, S.C.). Anaesthesia during surgery was assessed by the lack of reflexes to a short noxious pinch on the hind...
Singlet ground-state fluctuations in praseodymium observed by muon spin relaxation in PrP and PrP0.9
International Nuclear Information System (INIS)
Noakes, D R; Waeppling, R; Kalvius, G M; Jr, M F White; Stronach, C E
2005-01-01
Muon spin relaxation (μSR) in the singlet ground-state compounds PrP and PrP 0.9 reveals the unusual situation of a Lorentzian local field distribution with fast-fluctuation-limit strong-collision dynamics, a case that does not show motional narrowing. Contrary to publications by others, where PrP 0.9 was asserted to have vacancy-induced spin-glass freezing, no spin-glass freezing is seen in PrP 0.9 or PrP down to ≤100mK. This was confirmed by magnetization measurements on these same samples. In both compounds, the muon spin relaxation rate does increase as temperature decreases, demonstrating increasing strength of the paramagnetic response. A Monte Carlo model of fluctuations of Pr ions out of their crystalline-electric-field singlet ground states into their magnetic excited states (and back down again) produces the strong-collision-dynamic Lorentzian relaxation functions observed at each individual temperature but not the observed temperature dependence. This model contains no exchange interaction, and so predicts decreasing paramagnetic response as the temperature decreases, contrary to the temperature dependence observed. Comparison of the simulations to the data suggests that the exchange interaction is causing the system to approach magnetic freezing (by mode softening), but fails to complete the process
Combustion synthesis and characterization of MV0.5P0.5O4: Sm3+, Tm3+ (M = Gd, La, Y)
Motloung, Selepe J.; Lephoto, Mantwa A.; Tshabalala, Kamohelo G.; Ntwaeaborwa, Odireleng M.
2018-04-01
In this paper, GdV0.5P0.5O4: Sm3+, Tm3+, LaV0.5P0.5O4: Sm3+, Tm3+ and YV0.5P0.5O4: Sm3+, Tm3+ phosphor powders were prepared by solution combustion method using urea as a fuel. The phase purity, surface morphology, optical and photoluminescence properties were investigated by X-ray diffraction (XRD), scanning electron microscopy (SEM), UV-vis spectroscopy and photoluminescence spectroscopy. The XRD results indicated that the prepared powders are of a single phase and crystallized in tetragonal structure for Gd and Y systems while monoclinic phase was observed for La system. SEM showed that the samples consisted of mixed structures. The estimated band gaps were 2.2, 2.4 and 2.3 eV for Y, Gd and La systems respectively. The photoluminescence results showed four emission peaks. One peak is assigned to 1G4 - 3H6 transition of Tm3+, and three other emission peaks are attributed to 6G5/2 - 6H5/2, 6G5/2 - 6H7/2 and 6G5/2 - 6H9/2 transitions of Sm3+. The photoluminescent intensity was the highest in the gadolinium system.
Magnetic-entropy change in Mn1.1Fe0.9P0.7As0.3-xGe x
International Nuclear Information System (INIS)
Tegus, O.; Fuquan, B.; Dagula, W.; Zhang, L.; Brueck, E.; Si, P.Z.; Boer, F.R. de; Buschow, K.H.J.
2005-01-01
We have studied the magnetic properties and magnetic-entropy changes of Mn 1.1 Fe 0.9 P 0.7 As 0.3-x Ge x compounds with x = 0, 0.05, 0.1, 0.15 and 0.3. X-ray diffraction (XRD) study shows all the compounds crystallize in the Fe 2 P-type structure. Magnetic measurements show that the Curie temperature increases from 150 K for Mn 1.1 Fe 0.9 P 0.7 As 0.3 to 380 K for Mn 1.1 Fe 0.9 P 0.7 Ge 0.3 . A field-induced first-order magnetic phase transition is observed above the Curie temperature for the compounds with x up to 0.15. There exists an optimal composition in which the first-order phase transition is the sharpest. The optimal composition for this system is x = 0.1. The maximal magnetic-entropy change derived from the magnetization data is about 40 J/(kg K) for a field change from 0 to 3 T
Thermo-Fluid Dynamics of Two-Phase Flow
Ishii, Mamrou
2011-01-01
"Thermo-fluid Dynamics of Two-Phase Flow, Second Edition" is focused on the fundamental physics of two-phase flow. The authors present the detailed theoretical foundation of multi-phase flow thermo-fluid dynamics as they apply to: Nuclear reactor transient and accident analysis; Energy systems; Power generation systems; Chemical reactors and process systems; Space propulsion; Transport processes. This edition features updates on two-phase flow formulation and constitutive equations and CFD simulation codes such as FLUENT and CFX, new coverage of the lift force model, which is of part
Several new thermo-hydraulic test facilities in NPIC
International Nuclear Information System (INIS)
Ye Shurong; Sun Yufa; Ji Fuyun; Zong Guifang; Guo Zhongchuan
1997-01-01
Several new thermo-hydraulic test facilities are under construction in Nuclear Power Institute of Chinese (NPIC) at Chengdu. These facilities include: 1. Nuclear Power Component Comprehensive Test Facility. 2. Reactor Hydraulic Modeling Test Facility. 3. Control Rod Drive Line Hydraulic Test Facility. 4. Large Scale Thermo-Hydraulic Test Facility. The construction of these facilities will make huge progress in the research and development capability of nuclear power technology in CHINA. The author will present a brief description of the design parameters flowchart and test program of these facilities
Preparation of nano-aluminum and studies on thermo-reaction properties
International Nuclear Information System (INIS)
Wei Sheng; Wang Chaoyang; Huang Yong; Wu Weidong; Tang Yongjian; Wei Jianjun
2002-01-01
The author presents the fabrication of nano-aluminum powders by evaporation-condensation method. The thermo gravimetric-differential scanning calorimetry technique is used to characterize the thermo-reaction properties between nano-aluminum powders and N 2 or Ar. The experiment results confirm the different thermo-reaction properties between block- and nano-aluminum
Energy Technology Data Exchange (ETDEWEB)
Fletcher, J K
1973-05-01
CTD is a computer program written in Fortran 4 to solve the multi-group diffusion theory equations in X, Y, Z and triangular Z geometries. A power print- out neutron balance and breeding gain are also produced. 4 references. (auth)
International Nuclear Information System (INIS)
Menezes, Welton A.; Filho, Hermes Alves; Barros, Ricardo C.
2014-01-01
Highlights: • Fixed-source S N transport problems. • Energy multigroup model. • Anisotropic scattering. • Slab-geometry spectral nodal method. - Abstract: A generalization of the spectral Green’s function (SGF) method is developed for multigroup, fixed-source, slab-geometry discrete ordinates (S N ) problems with anisotropic scattering. The offered SGF method with the one-node block inversion (NBI) iterative scheme converges numerical solutions that are completely free from spatial truncation errors for multigroup, slab-geometry S N problems with scattering anisotropy of order L, provided L < N. As a coarse-mesh numerical method, the SGF method generates numerical solutions that generally do not give detailed information on the problem solution profile, as the grid points can be located considerably away from each other. Therefore, we describe in this paper a technique for the spatial reconstruction of the coarse-mesh solution generated by the multigroup SGF method. Numerical results are given to illustrate the method’s accuracy
Verdam, M.G.E.; Oort, F.J.; van der Linden, Y.M.; Sprangers, M.A.G.
2015-01-01
Purpose: Missing data due to attrition present a challenge for the assessment and interpretation of change and response shift in HRQL outcomes. The objective was to handle such missingness and to assess response shift and ‘true change’ with the use of an attrition-based multigroup structural
Verdam, Mathilde G. E.; Oort, Frans J.; van der Linden, Yvette M.; Sprangers, Mirjam A. G.
2015-01-01
Missing data due to attrition present a challenge for the assessment and interpretation of change and response shift in HRQL outcomes. The objective was to handle such missingness and to assess response shift and 'true change' with the use of an attrition-based multigroup structural equation
Sideridis, Georgios D.; Tsaousis, Ioannis; Al-harbi, Khaleel A.
2015-01-01
The purpose of the present study was to extend the model of measurement invariance by simultaneously estimating invariance across multiple populations in the dichotomous instrument case using multi-group confirmatory factor analytic and multiple indicator multiple causes (MIMIC) methodologies. Using the Arabic version of the General Aptitude Test…
VARI-QUIR-3, 2-D Multigroup Steady-State Neutron Diffusion in X-Y R-Z or R-Theta Geometry
International Nuclear Information System (INIS)
Collier, George
1984-01-01
1 - Nature of physical problem solved: The steady-state, multigroup, two-dimensional neutron diffusion equations are solved in x-y, r-z, and r-theta geometry. 2 - Method of solution: A Gauss-Seidel type of solution with inner and outer iterations is used. The source is held constant during the inner iterations
Magnetic and magnetocaloric properties of the alloys Mn2-xFexP0.5As0.5 (0⩽x⩽0.5)
Gribanov, I. F.; Golovchan, A. V.; Varyukhin, D. V.; Val'kov, V. I.; Kamenev, V. I.; Sivachenko, A. P.; Sidorov, S. L.; Mityuk, V. I.
2009-10-01
The results of investigations of the magnetic and magnetocaloric properties of alloys from the system Mn2-xFexP0.5As0.5 (0⩽x⩽0.5) are presented. The magnetization measurements are performed in the temperature interval 4.2-700K in magnetic fields up to 8T. The entropy changes ΔS with the magnetic field changing from 0 to 2, 4, 5, and 8T are determined from the magnetization isotherms obtained near temperatures of the spontaneous appearance of the ferromagnetic state (TC,TAF -FM1), and the curves ΔS(T0) are constructed. It is found that TC and TAF-FM1 decrease monotonically with increasing manganese concentration and that the ferromagnetic phase is completely suppressed in Mn1.5Fe0.5P0.5As0.5. It is found that the concentration dependences of the maximum entropy jump (and the corresponding cold-storage capacity) and the magnitudes of the ferromagnetic moment of the unit cell with maxima for x =0.9 and 0.8 show extremal behavior. The data obtained are compared with the ferromagnetic moments calculated from first principles by the Korringa-Kohn-Rostoker method using the coherent-potential approximation (KKR-CPA)—the discrepancy for 0.5⩽x⩽0.7 is attributed to the appearance of an antiferromagnetic component of the magnetic structure. It is concluded that the alloys Mn2-xFexP0.5As0.5 have promise for use in magnetic refrigerators operating at room temperature.
EXAFS study of Mn1.28Fe0.67P0.46Si0.54 compound with first-order phase transition
International Nuclear Information System (INIS)
L, Yingjie; Huliyageqi, B; Haschaolu, W; Song, Zhiqiang; Tegus, O; Nakai, Ikuo
2014-01-01
Highlights: • We have investigated the Fe and Mn K edge XAFS spectra of the Mn 1.28 Fe 0.67 P 0.46 Si 0.54 compound at 25 K and 295 K. • The site occupation of the Fe and Mn atoms and local structure of Mn 1.28 Fe 0.67 P 0.46 Si 0.54 are determined. • The atomic distances between Fe–Fe in c-plane for the ferromagnetic state are larger than those in the paramagnetic state. - Abstract: The Fe 2 P-type MnFe(P,Si) compounds are investigated by means of magnetic measurements and X-ray absorption fine structure spectroscopy. Magnetic measurements show that the Mn 1.28 Fe 0.67 P 0.46 Si 0.54 compound undergoes a first-order phase transition at the Curie temperature of 254 K. The Fe K-edge and Mn K-edge X-ray absorption fine structure spectra show that Mn atom mainly located at the 3g sites, while the 3f sites are occupied by Fe atoms and Mn atom randomly. The distances between the Fe atom and its nearest neighbor atoms in a triangle Fe–Mn–Fe change from 2.80 Å at 25 K to 2.74 Å at 300 K. On the other hand, the distances between Fe atom and its second neighbor atoms change from 4.06 Å at 25 K to 4.02 Å at 300 K
Complex investigation of thermo-technical parameters of Ruskov andesite
Directory of Open Access Journals (Sweden)
František Krepelka
2006-12-01
Full Text Available The research of thermo-technical parameters of Ruskov andesite was made as a part of the complex research of its properties as well as of rock disintegration by the action of chemical flame on the rock surface, i.e. thermal spalling in particular. Thermal spalling is a process in which thermal stresses are induced in the surface layer of rock whose surface is thereby disintegrated into small parts, the so called spalls, by the brittle manner. The evaluation of thermo-technical properties of the studied rocks is necessary for the qualification and quantification of the thermal spalling process. The measured and evaluated parameters were the coefficient of linear thermal expansion, the coefficient of thermal conductivity, the specific heat capacity and the coefficient of thermal diffusivity. Andesite from the Ruskov locality was chosen as a basic experimental material for the investigation of thermal spalling upon preliminary experiments. The estimated thermo-technical parameters were analyzed regarding the application of thermal spalling for the disintegration of the Ruskov andesite. The outcome as that the values of determine thermo-technical parameters established an expectation for its successful application.
Enhanced thermo-mechanical performance and strain-induced ...
Indian Academy of Sciences (India)
Enhanced thermo-mechanical performance and strain-induced band gap reduction of TiO2@PVC nanocomposite films ... School of Chemical Engineering, Yeungnam University, Gyeongsan 712-749, Republic of Korea; School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749, Republic of Korea ...
Thermo effect of chemical reaction in irreversible electrochemical systems
International Nuclear Information System (INIS)
Tran Vinh Quy; Nguyen Tang
1989-01-01
From first law of thermodynamics the expressions of statistical calculation of 'Fundamental' and 'Thermo-chemical' thermal effects are obtained. Besides, method of calculation of thermal effect of chemical reactions in non-equilibrium electro-chemical systems is accurately discussed. (author). 7 refs
Probabilistic thermo-chemical analysis of a pultruded composite rod
DEFF Research Database (Denmark)
Baran, Ismet; Tutum, Cem Celal; Hattel, Jesper Henri
2012-01-01
In the present study the deterministic thermo-chemical pultrusion simulation of a composite rod taken from the literature [7] is used as a validation case. The predicted centerline temperature and cure degree profiles of the rod match well with those in the literature [7]. Following the validation...
Near-field NanoThermoMechanical memory
International Nuclear Information System (INIS)
Elzouka, Mahmoud; Ndao, Sidy
2014-01-01
In this letter, we introduce the concept of NanoThermoMechanical Memory. Unlike electronic memory, a NanoThermoMechanical memory device uses heat instead of electricity to record, store, and recover data. Memory function is achieved through the coupling of near-field thermal radiation and thermal expansion resulting in negative differential thermal resistance and thermal latching. Here, we demonstrate theoretically via numerical modeling the concept of near-field thermal radiation enabled negative differential thermal resistance that achieves bistable states. Design and implementation of a practical silicon based NanoThermoMechanical memory device are proposed along with a study of its dynamic response under write/read cycles. With more than 50% of the world's energy losses being in the form of heat along with the ever increasing need to develop computer technologies which can operate in harsh environments (e.g., very high temperatures), NanoThermoMechanical memory and logic devices may hold the answer
Effect of Thermo-extrusion Process Parameters on Selected Quality ...
African Journals Online (AJOL)
Effect of Thermo-extrusion Process Parameters on Selected Quality Attributes of Meat Analogue from Mucuna Bean Seed Flour. ... Nigerian Food Journal ... The product functional responses with coefficients of determination (R2) ranging between 0.658 and 0.894 were most affected by changes in barrel temperature and ...
Prediction of thermo-mechanical reliability of wafer backend processes
Gonda, V.; Toonder, den J.M.J.; Beijer, J.G.J.; Zhang, G.Q.; van Driel, W.D.; Hoofman, R.J.O.M.; Ernst, L.J.
2004-01-01
More than 65% of IC failures are related to thermal and mechanical problems. For wafer backend processes, thermo-mechanical failure is one of the major bottlenecks. The ongoing technological trends like miniaturization, introduction of new materials, and function/product integration will increase
Prediction of thermo-mechanical integrity of wafer backend processes
Gonda, V.; Toonder, den J.M.J.; Beijer, J.G.J.; Zhang, G.Q.; Hoofman, R.J.O.M.; Ernst, L.J.; Ernst, L.J.
2003-01-01
More than 65% of IC failures are related to thermal and mechanical problems. For wafer backend processes, thermo-mechanical failure is one of the major bottlenecks. The ongoing technological trends like miniaturization, introduction of new materials, and function/product integration will increase
Effect of Blend Ratio on Thermo-Physical and Sensory ...
African Journals Online (AJOL)
Thermo-physical properties of bread made from wheat, cassava and soybean blends were investigated. During investigation, the organoleptic acceptance of the composite wheat, cassava and soy bread was determined. All the blend ratios were exposed to equal heating rate during baking at set temperature of 230oC. The ...
Moroccan rock phosphate solubilization during a thermo-anaerobic ...
African Journals Online (AJOL)
In order to investigate the presence of thermo-tolerant rock phosphate (RP) solubilizing anaerobic microbes during the fermentation process, we used grassland as sole organic substrate to evaluate the RP solubilization process under anaerobic thermophilic conditions. The result shows a significant decrease of pH from ...
Thermo-aerobic bacteria from geothermal springs in Saudi Arabia ...
African Journals Online (AJOL)
Fifteen isolates of thermo-aerobic bacteria were found. Bacillus cereus, B. licheniformis, B. thermoamylovorans, Pseudomonas sp., Pseudomonas aeruginosa and Enterobacter sp. were dominant in hot springs. Genetic relatedness indicated that eleven Bacillus spp. grouped together formed several clusters within one main ...
ThermoDex An index of selected thermodynamic data handbooks
This database contains records for printed handbooks and compilations of thermodynamic and thermophysical data for chemical compounds and other substances. You can enter both a type of compound and a property, and ThermoDex will return a list of hand
Quantum electron transfer processes induced by thermo-coherent ...
Indian Academy of Sciences (India)
WINTEC
Thermo-coherent state; electron transfer; quantum rate. 1. Introduction. The study ... two surfaces,16 namely, one electron two-centered exchange problem,7–10 many ... temperature classical regime for the single and the two-mode cases have ...
InAs0.45P0.55/InP strained multiple quantum wells intermixed by inductively coupled plasma etching
International Nuclear Information System (INIS)
Cao, Meng; Wu, Hui-Zhen; Lao, Yan-Feng; Cao, Chun-Fang; Liu, Cheng
2009-01-01
The intermixing effect on InAs 0.45 P 0.55 /InP strained multiple quantum wells (SMQWs) by inductively coupled plasma (ICP) etching and rapid thermal annealing (RTA) is investigated. Experiments show that the process of ICP etching followed RTA induces the blue shift of low temperature photoluminescence (PL) peaks of QWs. With increasing etching depth, the PL intensities are firstly enhanced and then diminished. This phenomenon is attributed to the variation of surface roughness and microstructure transformation inside the QW structure during ICP processing.
The 2s2p 4P0sub(5/2) - 2p24Psup(e)sub(5/2)-transition in O VI
International Nuclear Information System (INIS)
Sjoedin, R.; Pihl, J.; Hallin, R.; Lindskog, J.; Marelius, A.
1976-03-01
The Li-like doubly excited transitions 2s2p 4 P 0 sub(5/2) - 2p 2 4 Psup(e)sub(5/2) in O VI has been studied with the beam-foil technique. Oxygen ion beams with energies between 4.5 to 9 MeV were used. The wavelength of the transition was measured to 944.0+-0.5 A and the lifetime for the upper level 2p 2 4 Psup(e) was measured to be 0.51+-0.04 ns. (Auth.)
Energy Technology Data Exchange (ETDEWEB)
Shestakov, A I; Offner, S R
2006-09-21
We present a scheme to solve the nonlinear multigroup radiation diffusion (MGD) equations. The method is incorporated into a massively parallel, multidimensional, Eulerian radiation-hydrodynamic code with adaptive mesh refinement (AMR). The patch-based AMR algorithm refines in both space and time creating a hierarchy of levels, coarsest to finest. The physics modules are time-advanced using operator splitting. On each level, separate 'level-solve' packages advance the modules. Our multigroup level-solve adapts an implicit procedure which leads to a two-step iterative scheme that alternates between elliptic solves for each group with intra-cell group coupling. For robustness, we introduce pseudo transient continuation ({Psi}tc). We analyze the magnitude of the {Psi}tc parameter to ensure positivity of the resulting linear system, diagonal dominance and convergence of the two-step scheme. For AMR, a level defines a subdomain for refinement. For diffusive processes such as MGD, the refined level uses Dirichet boundary data at the coarse-fine interface and the data is derived from the coarse level solution. After advancing on the fine level, an additional procedure, the sync-solve (SS), is required in order to enforce conservation. The MGD SS reduces to an elliptic solve on a combined grid for a system of G equations, where G is the number of groups. We adapt the 'partial temperature' scheme for the SS; hence, we reuse the infrastructure developed for scalar equations. Results are presented. We consider a multigroup test problem with a known analytic solution. We demonstrate utility of {Psi}tc by running with increasingly larger timesteps. Lastly, we simulate the sudden release of energy Y inside an Al sphere (r = 15 cm) suspended in air at STP. For Y = 11 kT, we find that gray radiation diffusion and MGD produce similar results. However, if Y = 1 MT, the two packages yield different results. Our large Y simulation contradicts a long-standing theory
Energy Technology Data Exchange (ETDEWEB)
Shestakov, A I; Offner, S R
2007-03-02
We present a scheme to solve the nonlinear multigroup radiation diffusion (MGD) equations. The method is incorporated into a massively parallel, multidimensional, Eulerian radiation-hydrodynamic code with adaptive mesh refinement (AMR). The patch-based AMR algorithm refines in both space and time creating a hierarchy of levels, coarsest to finest. The physics modules are time-advanced using operator splitting. On each level, separate 'level-solve' packages advance the modules. Our multigroup level-solve adapts an implicit procedure which leads to a two-step iterative scheme that alternates between elliptic solves for each group with intra-cell group coupling. For robustness, we introduce pseudo transient continuation ({Psi}tc). We analyze the magnitude of the {Psi}tc parameter to ensure positivity of the resulting linear system, diagonal dominance and convergence of the two-step scheme. For AMR, a level defines a subdomain for refinement. For diffusive processes such as MGD, the refined level uses Dirichet boundary data at the coarse-fine interface and the data is derived from the coarse level solution. After advancing on the fine level, an additional procedure, the sync-solve (SS), is required in order to enforce conservation. The MGD SS reduces to an elliptic solve on a combined grid for a system of G equations, where G is the number of groups. We adapt the 'partial temperature' scheme for the SS; hence, we reuse the infrastructure developed for scalar equations. Results are presented. We consider a multigroup test problem with a known analytic solution. We demonstrate utility of {Psi}tc by running with increasingly larger timesteps. Lastly, we simulate the sudden release of energy Y inside an Al sphere (r = 15 cm) suspended in air at STP. For Y = 11 kT, we find that gray radiation diffusion and MGD produce similar results. However, if Y = 1 MT, the two packages yield different results. Our large Y simulation contradicts a long-standing theory
Energy Technology Data Exchange (ETDEWEB)
Nguyen-Ngoc, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1969-07-01
In order to reduce computing time, two and three-dimensional multigroup neutron diffusion equations in cylindrical, rectangular (X, Y), (X, Y, Z) and hexagonal geometries are solved by the method of synthesis using an appropriate variational principle (stationary principle). The basic idea is to reduce the number of independent variables by constructing two or three-dimensional solutions from solutions of fewer variables, hence the name 'synthesis method'. Whatever the geometry, we are led to solve a system of ordinary differential equations with matrix coefficients to which one can apply well-known numerical methods: CHEBYSHEV's polynomial method, Gaussian elimination. Numerical results furnished by synthesis programs written for the IBM 7094, the IBM 360-75 and the CDC 6600 computers, are confronted with those which are given by programs employing the classical finite difference method. [French] En vue de reduire le-temps de calcul, les equations de diffusion neutronique, multigroupe, a deux et trois dimensions d'espace dans les geometries cylindrique, rectangulaire (X, Y), (X, Y, Z) et hexagonale sont resolues par la methode de synthese utilisant un principe variationnel approprie (principe stationnaire). L'idee consiste a reduire le nombre de variables independantes par construction d'une solution bi ou tridimensionnelle au moyen de solutions dependant d'un nombre inferieur de variables, d'ou le nom de la methode. Dans tous les cas de geometrie, nous sommes conduits a resoudre un systeme d'equations differentielles a coefficients matriciels auquel peuvent s'appliquer les methodes numeriques courantes; methode polynomiale de TCHEBYCHEFF et methode d'elimination de GAUSS. Les resultats numeriques obtenus par nos codes de synthese programmes sur IBM 7094, IBM 360-75 et CDC 6600, sont confrontes avec ceux que fournissent les programmes adoptant la methode classique des differences finies. (auteur)
Did Life Emerge in Thermo-Acidic Conditions?
Holmes, D. S.
2017-12-01
There is widespread, but not unanimous, agreement that life emerged in hot conditions by exploiting redox and pH disequilibria found on early earth. Although there are several hypotheses to explain the postulated pH disequilibria, few of these consider that life evolved at very low pH (biological evolution. This presentation will evaluate the pros and cons of the hypothesis that the early evolution of life occurred in thermo-acidic conditions. Such environments are thought to have been abundant on early earth and were probably rich in hydrogen and soluble metals including iron and sulfur that could have served as sources and sinks of electrons. Extant thermo-acidophiles thrive in such conditions. Low pH environments are rich in protons that are the major drivers of energy conservation by coupling to phosphorylation in virtually all organisms on earth; this may be a "biochemical fossil" reflecting the use of protons (low pH) in primitive energy conservation. It has also been proposed that acidic conditions favored the evolution of an RNA world with expanded catalytic activities. On the other hand, the idea that life emerged in thermo-acidic conditions can be challenged because of the proposed difficulties of folding and stabilizing proteins simultaneously exposed to high temperature and low pH. In addition, although thermo-acidophiles root to the base of the phylogenetic tree of life, consistent with the proposition that they evolved early, yet there are problems of interpretation of their subsequent evolution that cloud this simplistic phylogenetic view. We propose solutions to these problems and hypothesize that life evolved in thermo-acidic conditions.
Gning, Youssou; Sow, Malick; Traoré, Alassane; Dieng, Matabara; Diakhate, Babacar; Biaye, Mamadi; Wagué, Ahmadou
2015-01-01
In the present work a special computational program Scilab (Scientific Laboratory) in the complex rotation method has been used to calculate resonance parameters of ((2s2) 1Se, (2s2p) 1,3P0) and ((3s2) 1Se, (3s3p) 1,3P0) states of helium-like ions with Z≤10. The purpose of this study required a mathematical development of the Hamiltonian applied to Hylleraas wave function for intrashell states, leading to analytical expressions which are carried out under Scilab computational program. Results are in compliance with recent theoretical calculations.
International Nuclear Information System (INIS)
Zhang, L.; Szargan, R.; Chasse, T.
2004-01-01
ZnS films were grown by molecular beam epitaxy employing a single compound effusion cell on GaP(0 0 1) substrate at different temperatures, and characterised by means of low energy electron diffraction, X-ray and ultra-violet photoelectron spectroscopy, angle-resolved ultra-violet photoelectron spectroscopy and X-ray emission spectroscopy. The GaP(0 0 1) substrate exhibits a (4x2) reconstruction after Ar ion sputtering and annealing at 370 deg. C. Crystal quality of the ZnS films depends on both film thickness and growth temperature. Thinner films grown at higher temperatures and thicker films grown at lower temperatures have better crystal quality. The layer-by-layer growth mode of the ZnS films at lower (25, 80 and 100 deg. C) temperatures changes to layer-by-layer-plus-island mode at higher temperatures (120, 150 and 180 deg. C). A chemical reaction takes place and is confined to the interface. The valence band offset of the ZnS-GaP heterojunction was determined to be 0.8±0.1 eV. Sulphur L 2,3 emission spectra of ZnS powder raw material and the epitaxial ZnS films display the same features, regardless of the existence of the Ga-S bonding in the film samples
Photoluminescence characterization of GaAs/GaAs0.64P0.19Sb0.17/GaAs heterostructure
International Nuclear Information System (INIS)
Chen, J.Y.; Chen, B.H.; Huang, Y.S.; Chin, Y.C.; Tsai, H.S.; Lin, H.H.; Tiong, K.K.
2013-01-01
Interfacial characteristics of GaAs/GaAs 0.64 P 0.19 Sb 0.17 GaAs heterostructures and emission properties of a quaternary GaAs 0.64 P 0.19 Sb 0.17 layer were studied by excitation-power- and temperature-dependent photoluminescence (PL) measurements. The GaAs-to-GaAsPSb upper interface related emission feature and signals from GaAsPSb and GaAs were observed and characterized. The upper interface related emission peak was attributed to the radiative recombination of spatially separated electron–hole pairs and suggesting the type-II alignment at the GaAs/GaAsPSb interface. The localized excitonic emission feature of GaAsPSb revealed a blueshift due to the saturation effect of localized states and showed a fast thermal-quench with the increase of temperature. The temperature variation of the band edge emission signal of GaAsPSb was found to follow that of GaAs closely. -- Highlights: ► PL characterization of GaAs/GaAsPSb/GaAs heterostructure. ► Type-II alignment at the GaAs/GaAsPSb interface. ► Near-band-edge emission lines of GaAsPSb
Thermo-economic optimization of an endoreversible four-heat-reservoir absorption-refrigerator
International Nuclear Information System (INIS)
Qin Xiaoyong; Chen Lingen; Sun Fengrui; Wu Chih
2005-01-01
Based on an endoreversible four-heat-reservoir absorption-refrigeration-cycle model, the optimal thermo-economic performance of an absorption-refrigerator is analyzed and optimized assuming a linear (Newtonian) heat-transfer law applies. The optimal relation between the thermo-economic criterion and the coefficient of performance (COP), the maximum thermo-economic criterion, and the COP and specific cooling load for the maximum thermo-economic criterion of the cycle are derived using finite-time thermodynamics. Moreover, the effects of the cycle parameters on the thermo-economic performance of the cycle are studied by numerical examples
Thermo-hydro-mechanical behavior of argillite
International Nuclear Information System (INIS)
Tran, Duy Thuong; Dormieux, Luc; Lemarchand, Eric; Skoczylas, Frederic
2012-01-01
Document available in extended abstract form only. Argillite is a very low permeability geo-material widely encountered: that is the reason why it is an excellent candidate for the storage of long-term nuclear waste depositories. This study focuses on argillites from Meuse-Haute-Marne (East of France) which forms a geological layer located approximately 400 m and 500 m depth. We know that this material is made up of a mixture of shale, quartz and calcite phases. The multi-scale definition of this material suggests the derivation of micro-mechanics reasonings in order to better account for the mechanisms occurring at the local (nano and micro-) scale and controlling the macroscopic mechanical behavior. In this work, up-scaling techniques are used in the context of thermo-hydro-mechanical couplings. The first step consists in clarifying the morphology of the microstructure at the relevant scales (particles arrangement, pore size distribution) and identifying the mechanisms that take place at those scales. These local informations provide the input data of micro-mechanics based models. Schematic picture of the microstructure where the argillite material behaves as a dual-porosity, with liquid in both micro-pores and interlayer space in between clay solid platelets, seems a reasonable starting point for this micro-mechanical modelling of clay. This allows us to link the physical phenomena (swelling clays) and the mechanical properties (elastic moduli, Poisson's ratio). At the pressure applied by the fluid on the solid platelets appears as the sum of the uniform pressure in the micro-pores and of a swelling overpressure depending on the distance between platelets and on the ion concentration in the micro-pores. The latter is proved to be responsible for a local elastic modulus of physical origin. This additional elastic component may strongly be influenced by both relative humidity and temperature. A first contribution of this study is to analysing this local elastic
International Nuclear Information System (INIS)
Kobayashi, Keisuke; Kikuchi, Hirohiko; Tsutsuguchi, Ken
1993-01-01
A neutron multigroup transport equation in x-y-z geometry is solved by the spherical harmonics method using finite Fourier transformation. Using the first term of the Fourier series for the space variables of spherical harmonics moments, three-point finite difference like equations are derived for x-, y- and z-axis directions, which are more consistent and accurate than those derived using the usual finite difference approximation, and these equations are solved by the iteration method in each axis direction alternatively. A method to find an optimum acceleration factor for this inner iteration is described. It is shown in the numerical examples that the present method gives higher accuracy with less mesh points that the usual finite difference method. (author)
Energy Technology Data Exchange (ETDEWEB)
Díez, C.J., E-mail: cj.diez@upm.es [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Cabellos, O. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Martínez, J.S. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain)
2015-01-15
Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.
International Nuclear Information System (INIS)
Díez, C.J.; Cabellos, O.; Martínez, J.S.
2015-01-01
Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties
Díez, C. J.; Cabellos, O.; Martínez, J. S.
2015-01-01
Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.
Directory of Open Access Journals (Sweden)
Foo Fatt Mee
2017-06-01
Full Text Available This study aims to measure the latent mean difference in perfectionism and marital satisfaction by counseling help-seeking attitudes. The respondents were 327 married graduate students from a research university in Malaysia. An online self-administered questionnaire was used to collect the data. The respondents completed the Almost Perfect Scale- Revised, Dyadic Almost Perfect Scale, Marital Satisfaction Scale, and Attitudes toward Seeking Professional Psychology Help Scale. Confirmatory factor analysis was used to examined the instruments and the results indicated that construct validity were achieved. The latent mean difference in perfectionism and marital satisfaction by counseling help-seeking attitudes were tested using multigroup invariance analysis. The respondents with negative attitudes toward counseling help-seeking (n = 159 reported a higher latent mean in perfectionism but a lower latent mean in marital satisfaction compared to those with positive attitudes toward counseling help-seeking (n = 168. The implications of these findings for counseling services are discussed.
International Nuclear Information System (INIS)
White, J.E.; Ingersoll, D.T.; Slater, C.O.; Roussin, R.W.
1996-01-01
A revised multigroup cross-section library based ON ENDF/B-VI Release 3 has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 library released in February 1994 and is expected to replace te BUGLE-93 data. The cross-section processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. As an added feature, cross-section sets having upscatter data for four thermal neutron groups are included in the BUGLE-96 package available from the Radiation Shielding Information Center. The upscattering data should improve the application of this library to the calculation of more accurate thermal fluences, although more computer time will be required. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs
International Nuclear Information System (INIS)
Grimstone, M.J.
1978-06-01
The WRS Modular Programming System has been developed as a means by which programmes may be more efficiently constructed, maintained and modified. In this system a module is a self-contained unit typically composed of one or more Fortran routines, and a programme is constructed from a number of such modules. This report describes one WRS module, the function of which is to solve a set of multigroup diffusion equations for a system represented in one-dimensional plane, cylindrical or spherical geometry. The information given in this manual is of use both to the programmer wishing to incorporate the module in a programme, and to the user of such a programme. (author)
International Nuclear Information System (INIS)
Matausek, M.V.; Milosevic, M.
1986-01-01
In the present paper a generalization is performed of a procedure to solve multigroup spherical harmonics equations, which has originally been proposed and developed for one-dimensional systems in cylindrical or spherical geometry, and later extended for a special case of a two-dimensional system in r-z geometry. The expressions are derived for the axial and the radial dependence of the group values of the neutron flux moments, in the P-3 approximation of the spherical harmonics method, in a cylindrically symmetrical system with an arbitrary number of material regions in both r- and z-directions. In the special case of an axially homogeneous system, these expressions reduce to the relations derived previously. (author)
International Nuclear Information System (INIS)
Bastos, H.F.B.N.
1979-01-01
In this work a study of the methodology of the adjustment of multigroup cross sections by means of integral data is presented. A synthesis of the principal methods existent and the mathematical development of the adaptation of one of them are made. A calculational system is built from this reference method, with the basic conditions for the operation of the process of adjustment. In order to test the system developed and analyze several problems related to the adjustment, a series of trial adjustments was made with the value of the U 235 fission cross section from the infinite dilution library used in the calculational system for fast reactors of the Instituto de Engenharia Nuclear. (author)
Multigroup analysis of nuclear elastic scattering effects in Cat-D and DD3He fusion plasmas
International Nuclear Information System (INIS)
Nakano, Yasuyuki; Hanada, Takahiro; Hori, Hidetoshi; Kudo, Kazuhiko; Ohta, Masao
1987-01-01
Effects of nuclear elastic scattering (NES) on the slowing down of charged fusion products in a typical deuterium plasma and the burn dynamics of ignited Cat-D and DD 3 He plasmas are investigated. A time-dependent multigroup method is used to take into account the effect of finite (non-zero) slowing-down time as well as the discrete nature of NES. It is shown that adequate treatment of the slowing-down process, especially consideration of NES and slowing-down time delay, is essential for an accurate prediction of the dynamic behavior and thermal instability of the plasmas. NES accelerates the temporal plasma behavior and enhances the thermal instability, leading to 20∼30 keV increase in the critical temperature. (author)
PUFF-IV, Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files
International Nuclear Information System (INIS)
2007-01-01
1 - Description of program or function: The PUFF-IV code system processes ENDF/B-VI formatted nuclear cross section covariance data into multigroup covariance matrices. PUFF-IV is the newest release in this series of codes used to process ENDF uncertainty information and to generate the desired multi-group correlation matrix for the evaluation of interest. This version includes corrections and enhancements over previous versions. It is written in Fortran 90 and allows for a more modular design, thus facilitating future upgrades. PUFF-IV enhances support for resonance parameter covariance formats described in the ENDF standard and now handles almost all resonance parameter covariance information in the resolved region, with the exception of the long range covariance sub-subsections. PUFF-IV is normally used in conjunction with an AMPX master library containing group averaged cross section data. Two utility modules are included in this package to facilitate the data interface. The module SMILER allows one to use NJOY generated GENDF files containing group averaged cross section data in conjunction with PUFF-IV. The module COVCOMP allows one to compare two files written in COVERX format. 2 - Methods: Cross section and flux values on a 'super energy grid,' consisting of the union of the required energy group structure and the energy data points in the ENDF/B-V file, are interpolated from the input cross sections and fluxes. Covariance matrices are calculated for this grid and then collapsed to the required group structure. 3 - Restrictions on the complexity of the problem: PUFF-IV cannot process covariance information for energy and angular distributions of secondary particles. PUFF-IV does not process covariance information in Files 34 and 35; nor does it process covariance information in File 40. These new formats will be addressed in a future version of PUFF
International Nuclear Information System (INIS)
Lourenco, M.J.V.; Santos, F.J.V.; Ramires, M.L.V.; Nieto de Castro, C.A.
2006-01-01
There has been some controversy regarding the uncertainty of measurements of thermal properties using differential scanning calorimeters, namely heat capacity of liquids. A differential scanning calorimeter calibrated in enthalpy and temperature was used to measure the isobaric specific heat capacity of water and aqueous solutions of cesium chloride, in the temperature range 298 K to 370 K, for molalities up 3.2 mol . kg -1 , at p = 0.1 MPa, with an estimated uncertainty (ISO definition) better than 1.1%, at a 95% confidence level. The measurements are completely traceable to SI units of energy and temperature. The results obtained were correlated as a function of temperature and molality and compared with other authors, obtained by different methods and permit to conclude that a DSC calibrated by Joule effect is capable of very accurate measurements of the isobaric heat capacity of liquids, traceable to SI units of measurement
International Nuclear Information System (INIS)
Courtillot, I.
2003-11-01
This thesis reports the first results towards the realization of an optical clock using trapped strontium atoms. This set up would combine advantages of the different approaches commonly used to develop an atomic frequency standard. The first part describes the cold atoms source which is implemented. A magneto-optical trap operating on the 1 S 0 - 1 P 1 transition at 461 nm is loaded from an atomic beam decelerated by a Zeeman slower. The 461 nm laser is obtained by sum-frequency mixing in a potassium titanyl phosphate (KTP) crystal. The second part is devoted to the different stages developed to achieve the direct excitation of the 1 S 0 - 3 P 0 clock transition in 87 Sr. This line has a theoretical natural width of 10 -3 Hz. Before this detection, we obtained an estimate of the resonance frequency by measuring absolute frequencies of several allowed optical transitions. (author)
Thieulot, Cedric
2016-04-01
Many Finite Element geodynamical codes (Fullsack,1995; Zhong et al., 2000; Thieulot, 2011) are based on bi/tri-linear velocity constant pressure element (commonly called Q1P0), because of its ease of programming and rather low memory footprint, despite the presence of (pressure) checkerboard modes. However, it is long known that the Q1P0 is not inf-sup stable and does not lend itself to the use of iterative solvers, which makes it a less than ideal candidate for high resolution 3D models. Other attempts were made more recently (Burstedde et al., 2013; Le Pourhiet et al., 2012) with the use of the stabilised Q1Q1 element (bi/tri-linear velocity and pressure). This element, while also attractive from an implementation and memory standpoint, suffers a major drawback due to the artificial compressibility introduced by the polynomial projection stabilization. These observations have shifted part of the community towards the Finite Difference Method while the remaining part is now embracing infsup stable second order elements [May et al., 2015; Kronbichler,2012). Rather surprinsingly, a third option exists when it comes to first order elements in the form of the stabilised Q1P0 element, but virtually no literature exists concerning its use for geodynamical applications. I will then recall the specificity of the stabilisation and will carry out a series of benchmark experiments and geodynamical tests to assess its performance. While being shown to work as expected in benchmark experiments, the stabilised Q1P0 element turns out to introduce first-order numerical artefacts in the velocity and pressure solutions in the case of buoyancy-driven flows. Burstedde, C., Stadler, G., Alisic, L., Wilcox, L. C., Tan, E., Gurnis, M., & Ghattas, O. (2013). Largescale adaptive mantle convection simulation. Geophysical Journal International, 192(3), 889906. Fullsack, P. (1995). An arbitrary LagrangianEulerian formulation for creeping flows and its application in
Andrade, E.; Canto, C. E.; Rocha, M. F.
2017-09-01
The absolute energy of an ion beam produced by an accelerator is usually determined by an electrostatic or magnetic analyzer, which in turn must be calibrated. Various methods for accelerator energy calibration are extensively reported in the literature, like nuclear reaction resonances, neutron threshold, and time of flight, among others. This work reports on a simple method to calibrate the magnet associated to a vertical 5.5 MV Van de Graaff accelerator. The method is based on bombarding with deuteron beams a thick carbon target and measuring with a surface barrier detector the particle energy spectra produced. The analyzer magnetic field is measured for each spectrum and the beam energy is deduced by the best fit of the simulation of the spectrum with the SIMNRA code that includes 12C(d,p0)13C nuclear cross sections.
Transition probabilities for the 3s2 3p(2P0)-3s3p2(4P) intersystem lines of Si II
Calamai, Anthony G.; Smith, Peter L.; Bergeson, S. D.
1993-01-01
Intensity ratios of lines of the spin-changing 'intersystem' multiplet of S II (4P yields 2P0) at 234 nm have been used to determine electron densities and temperatures in a variety of astrophysical environments. However, the accuracy of these diagnostic calculations have been limited by uncertainties associated with the available atomic data. We report the first laboratory measurement, using an ion-trapping technique, of the radiative lifetimes of the three metastable levels of the 3s3p2 4P term of Si II. Our results are 104 +/- 16, 406 +/- 33, and 811 +/- 77 micro-s for lifetimes of the J = 1/2, 5/2, and 3/2 levels, respectively. A-values were derived from our lifetimes by use of measured branching fractions. Our A-values, which differ from calculated values by 30 percent or more, should give better agreement between modeled and observed Si II line ratios.
Williams, I. D.; Chutjian, A.; Msezane, A. Z.; Henry, R. J. W.
1985-01-01
Angular differential electron scattering cross sections are reported for the unresolved inelastic 3s (2)S to 3p (2)P0 h, k transitions in Mg II for the first time. Relative differential cross sections have been measured at 35 eV and 50 eV in the angular range of Theta between 6 and 17 deg using the newly developed electron energy loss technique in a crossed electron-ion beam geometry. Theoretical values have been calculated in a five-state close-coupling approximation in which 3s, 3p, 3d, 4s, and 4p states were included, and to which measurements were normalized at Theta = 12 deg.
DEFF Research Database (Denmark)
Moldovan, Mihai; Pinchenko, Volodymyr; Dmytriyeva, Oksana
2013-01-01
and mimicked the S100A4-induced neuroprotection in brain trauma. Here, we investigated a possible function of S100A4 and its mimetics in the pathologies of the peripheral nervous system (PNS). We found that S100A4 was expressed in the injured PNS and that its peptide mimetic (H3) affected the regeneration......, these effects were attributed to the modulatory effect of H3 on initial axonal sprouting. In contrast to the modest effect of H3 on the time course of regeneration, H3 had a long-term neuroprotective effect in the myelin protein P0 null mice, a model of dysmyelinating neuropathy (Charcot-Marie-Tooth type 1...... disease), where the peptide attenuated the deterioration of nerve conduction, demyelination and axonal loss. From these results, S100A4 mimetics emerge as a possible means to enhance axonal sprouting and survival, especially in the context of demyelinating neuropathies with secondary axonal loss...
Magnetic properties and magnetocaloric effect of MnFeP0.5Ge0.5-xSix compounds
International Nuclear Information System (INIS)
Song, L.; Wang, G.F.; Ou, Z.Q.; Haschaolu, O.; Tegus, O.; Brueck, E.; Buschow, K.H.J.
2009-01-01
We have studied the magnetic properties and magnetic-entropy changes of the MnFeP 0.5 Ge 0.5-x Si x compounds with x = 0.1, 0.2, 0.3, 0.4 and 0.45. X-ray diffraction shows that the compounds crystallize in the Fe 2 P-type hexagonal structure. The lattice parameter a and the Curie temperature decreases with increasing x. The maximal magnetic-entropy changes for x = 0.4 and 0.45 derived from the magnetization data are about 6.0 J/kg K and 5.8 J/kg K, respectively, for a field change from 0 to 1.5 T
ZZ ANSLV, Multigroup Cross Sections Library for ANS Reactor Design Studies
International Nuclear Information System (INIS)
2000-01-01
A - Description of program or function: - Format: AMPX Master Interface Library format. Number of groups: Fine Group (99 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Broad Group (39 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Gamma-Ray Interaction (GRI) Library in 44-groups. Materials: H, He, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Xe, Sm, Eu, Hf, Ta, Ir, Pb, Th, U, Pu. Origin: ENDF/B-V; LENDL-V evaluations for 12 materials. - Format: AMPX Master Interface Library format. Number of groups: Coupled Library containing (CNG) 99-group neutron and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. Weighting spectrum: Maxwellian 300 K + 1/(E*sigma-total) + fission spectrum4 types of boundaries have been used depending isotope and library type (see report). Pseudo-problem-independent, multigroup cross section libraries were generated to support the Advanced Neutron source (ANS) reactor design studies. The ANS was
NRC wants plant-specific responses on Thermo-Lag
International Nuclear Information System (INIS)
Anon.
1994-01-01
Dissatisfied with recent industry-backed efforts to assure fire safety at nuclear power plants, the Nuclear Regulatory Commission announced on November 24 that it would direct all nuclear plant owners to specify the actions they would take to assure that the use of the Thermo-Lag 330 fire barrier material would not lead to insufficient protection of electrical cables connected to safe-shutdown systems. Previously, the NRC had been content to let the matter wait until tests sponsored by the Nuclear Management and Resources Council (Numarc) could show whether Thermo-Lag, used and installed in certain ways, would provide sufficient protection, but the NRC and Numarc have disagreed over the test methodology, and the Numarc tests are now considered to be several months behind schedule
Thermo-fluid behaviour of periodic cellular metals
Lu, Tian Jian; Wen, Ting
2013-01-01
Thermo-Fluid Behaviour of Periodic Cellular Metals introduces the study of coupled thermo-fluid behaviour of cellular metals with periodic structure in response to thermal loads, which is an interdisciplinary research area that requires a concurrent-engineering approach. The book, for the first time, systematically adopts experimental, numerical, and analytical approaches, presents the fluid flow and heat transfer in periodic cellular metals under forced convection conditions, aiming to establish structure-property relationships for tailoring material structures to achieve properties and performance levels that are customized for defined multifunctional applications. The book, as a textbook and reference book, is intended for both academic and industrial people, including graduate students, researchers and engineers. Dr. Tian Jian Lu is a professor at the School of Aerospace, Xi’an Jiaotong University, Xi’an, China. Dr. Feng Xu is a professor at the Key Laboratory of Biomedical Information Engineering o...
Thermo-cured glass ionomer cements in restorative dentistry.
Gorseta, Kristina; Glavina, Domagoj
2017-01-01
Numerous positive properties of glass ionomer cements including biocompatibility, bioactivity, releasing of fluoride and good adhesion to hard dental tissue even under wet conditions and easy of handling are reasons for their wide use in paediatric and restorative dentistry. Their biggest drawbacks are the weaker mechanical properties. An important step forward in improving GIC's features is thermo-curing with the dental polymerization unit during setting of the material. Due to their slow setting characteristics the GIC is vulnerable to early exposure to moisture. After thermo curing, cements retain all the benefits of GIC with developed better mechanical properties, improved marginal adaptation, increased microhardness and shear bond strength. Adding external energy through thermocuring or ultrasound during the setting of conventional GIC is crucial to achieve faster and better initial mechanical properties. Further clinical studies are needed to confirm these findings.
Black Holes Versus Firewalls and Thermo-Field Dynamics
Chowdhury, Borun D.
2013-09-01
In this paper, we examine the implications of the ongoing black holes versus firewalls debate for the thermo-field dynamics of black holes by analyzing a conformal field theory (CFT) in a thermal state in the context of anti-de Sitter/CFT. We argue that the thermo-field doubled copy of the thermal CFT should be thought of not as a fictitious system, but as the image of the CFT in the heat bath. In case of strong coupling between the CFT and the heat bath, this image allows for free infall through the horizon and the system is described by a black hole. Conversely, firewalls are the appropriate dual description in case of weak interaction of the CFT with its heat bath.
Thermo-electrical systems for the generation of electricity
International Nuclear Information System (INIS)
Bitschi, A.; Froehlich, K.
2010-01-01
This article takes a look at theoretical models concerning thermo-electrical systems for the generation of electricity and demonstrations of technology actually realised. The potentials available and developments are discussed. The efficient use of energy along the whole generation and supply chain, as well as the use of renewable energy sources are considered as being two decisive factors in the attainment of a sustainable energy supply system. The large amount of unused waste heat available today in energy generation, industrial processes, transport systems and public buildings is commented on. Thermo-electric conversion systems are discussed and work being done on the subject at the Swiss Federal Institute of Technology in Zurich is discussed. The findings are discussed and results are presented in graphical form
Dynamic thermo-hydraulic model of district cooling networks
International Nuclear Information System (INIS)
Oppelt, Thomas; Urbaneck, Thorsten; Gross, Ulrich; Platzer, Bernd
2016-01-01
Highlights: • A dynamic thermo-hydraulic model for district cooling networks is presented. • The thermal modelling is based on water segment tracking (Lagrangian approach). • Thus, numerical errors and balance inaccuracies are avoided. • Verification and validation studies proved the reliability of the model. - Abstract: In the present paper, the dynamic thermo-hydraulic model ISENA is presented which can be applied for answering different questions occurring in design and operation of district cooling networks—e.g. related to economic and energy efficiency. The network model consists of a quasistatic hydraulic model and a transient thermal model based on tracking water segments through the whole network (Lagrangian method). Applying this approach, numerical errors and balance inaccuracies can be avoided which leads to a higher quality of results compared to other network models. Verification and validation calculations are presented in order to show that ISENA provides reliable results and is suitable for practical application.
High-coercive garnet films for thermo-magnetic recording
International Nuclear Information System (INIS)
Berzhansky, V N; Danishevskaya, Y V; Nedviga, A S; Milyukova, H T
2016-01-01
The possibility of using high-coercive of garnet films for thermo-magnetic recording is related with the presence of the metastable domain structure, which arises due to a significant mismatch of the lattice parameters of the film and the substrate. In the work the connection between facet crystal structure of elastically strained ferrite garnets films and the domain structure in them is established by methods of phase contrast and polarization microscopy. (paper)
Thermo-Mechanical Fatigue Crack Growth of RR1000
Christopher John Pretty; Mark Thomas Whitaker; Steve John Williams
2017-01-01
Non-isothermal conditions during flight cycles have long led to the requirement for thermo-mechanical fatigue (TMF) evaluation of aerospace materials. However, the increased temperatures within the gas turbine engine have meant that the requirements for TMF testing now extend to disc alloys along with blade materials. As such, fatigue crack growth rates are required to be evaluated under non-isothermal conditions along with the development of a detailed understanding of related failure mechan...
The thermo-electric nature of the Debye temperature
Directory of Open Access Journals (Sweden)
Mithun Bhowmick
2018-05-01
Full Text Available The Debye temperature is typically associated with the heat capacity of a solid and the cut-off of the possible lattice vibrations, but not necessarily to the electric conductivity of the material. By investigating III-V and II-VI compound semiconductors, we reveal that the Debye temperature represents a thermo-electric material parameter, connecting the thermal and electronic properties of a solid via a distinct power law.
Thermal energy storage using thermo-chemical heat pump
International Nuclear Information System (INIS)
Hamdan, M.A.; Rossides, S.D.; Haj Khalil, R.
2013-01-01
Highlights: ► Understanding of the performance of thermo chemical heat pump. ► Tool for storing thermal energy. ► Parameters that affect the amount of thermal stored energy. ► Lithium chloride has better effect on storing thermal energy. - Abstract: A theoretical study was performed to investigate the potential of storing thermal energy using a heat pump which is a thermo-chemical storage system consisting of water as sorbet, and sodium chloride as the sorbent. The effect of different parameters namely; the amount of vaporized water from the evaporator, the system initial temperature and the type of salt on the increase in temperature of the salt was investigated and hence on the performance of the thermo chemical heat pump. It was found that the performance of the heat pump improves with the initial system temperature, with the amount of water vaporized and with the water remaining in the system. Finally it was also found that lithium chloride salt has higher effect on the performance of the heat pump that of sodium chloride.
Thermo-ecological optimization of a solar collector
International Nuclear Information System (INIS)
Szargut, J.; Stanek, W.
2007-01-01
The depletion of non-renewable natural exergy resources (the thermo-ecological cost) has been accepted as the objective function for thermo-ecological optimization. Its general formulation has been cited. A detailed form of the objective function has been formulated for a solar collector producing hot water for household needs. The following design parameters have been accepted as the decision variables: the collector area per unit of the heat demand, the diameter of collector pipes, the distance of the pipe axes in the collector plate. The design parameters of the internal installation (the pipes, the hot water receiver) have not been taken into account, because they are very individual. The accumulation ability of hot water comprising one day has been assumed. The objective function contains the following components: the thermo-ecological cost of copper plate, copper pipes, glass plate, steel box, thermal insulation, heat transfer liquid, electricity for driving the pump of liquid, fuel for the peak boiler. The duration curves of the flux of solar radiation and absorbed heat have been elaborated according to meteorological data and used in the calculations. The objective function for economic optimization may have a similar form, only the cost values would be different
Fundamental topics for thermo-elastic stress analyses
International Nuclear Information System (INIS)
Biermann, M.
1989-01-01
This paper delivers a consistent collection of theoretical fundamentals needed to perform rather sound experimental stress analyses on thermo-elastic materials. An exposition of important concepts of symmetry and so-called peer groups, yielding the very base for a rational description of materials, goes ahead and is followed by an introduction to the constitutive theory of simple materials. Neat distinction is made between stress contributions determined by deformational and thermal impressions, on the one part, and stress constraints not accessible to strain gauging, on the other part. The mathematical formalism required for establishing constitutive equations is coherently developed from scratch and aided, albeit not subrogated, by intuition. The main intention goes to turning some of the recent advances in the nonlinear field theories of thermomechanics to practical account. A full success therein, obviously, results under the restriction to thermo-elasticity. In adverting to more particular subjects, the elementary static effects of nonlinear isotropic elasticity are pointed out. Due allowance is made for thermal effects likely to occur in heat conducting materials also beyond the isothermal or isentropic limit cases. Linearization of the constitutive equations for anisotropic thermo-elastic materials is then shown to entail the formulas of the classical theory. (orig./MM) [de
International Nuclear Information System (INIS)
2005-01-01
A - Description of program or function: (1) Problems to be solved: MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used. - BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism; - BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid; - Arbitrary quadratic surface and torus. The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. (4) Cross sections: The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-3.3, ENDF/B-VI, JEF-3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data. (5) Boundary conditions: Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be
International Nuclear Information System (INIS)
Gning, Youssou; Sow, Malick; Traoré, Alassane; Dieng, Matabara; Diakhate, Babacar; Biaye, Mamadi; Wagué, Ahmadou
2015-01-01
In the present work a special computational program Scilab (Scientific Laboratory) in the complex rotation method has been used to calculate resonance parameters of ((2s 2 ) 1 S e , (2s2p) 1,3 P 0 ) and ((3s 2 ) 1 S e , (3s3p) 1,3 P 0 ) states of helium-like ions with Z≤10. The purpose of this study required a mathematical development of the Hamiltonian applied to Hylleraas wave function for intrashell states, leading to analytical expressions which are carried out under Scilab computational program. Results are in compliance with recent theoretical calculations. - Highlights: • Resonance energy and widths computed for doubly excited states of helium-like ions. • Well-comparable results to the theoretical literature values up to Z=10. • Satisfactory agreements with theoretical calculations for widths
Energy Technology Data Exchange (ETDEWEB)
Le, T.T
2008-01-15
This thesis studied the thermo-hydro-mechanical properties of Boom clay, which was chosen to be the host material for the radioactive waste disposal in Mol, Belgium. Firstly, the research was concentrated on the soil water retention properties and the hydro-mechanical coupling by carrying out axial compression tests with suction monitoring. The results obtained permitted elaborating a rational experimental procedure for triaxial tests. Secondly, the systems for high pressure triaxial test at controlled temperature were developed to carry out compression, heating, and shearing tests at different temperatures. The obtained results showed clear visco-elasto-plastic behaviour of the soil. This behaviour was modelled by extending the thermo-elasto-plastic model of Cui et al. (2000) to creep effect. (author)
International Nuclear Information System (INIS)
Rastas, A.
1985-01-01
1 - Description of problem or function: THERMLIB is a code that generates, revises and expands the input data library to the lattice cell code THERMOS-OTA. It can be used to: - create an entirely new library; - modify the data of library materials, remove materials, add materials; - list the library. 2 - Restrictions on the complexity of the problem: Max. of 30 materials may be modified or removed. Max. of 30 new materials may be created. Max. of 50 velocity groups
Thermo-Hydraulic Modelling of Buffer and Backfill
International Nuclear Information System (INIS)
Pintado, X.; Rautioaho, E.
2013-09-01
The temporal evolution of saturation, liquid pressure and temperature in the components of the engineered barrier system was studied using numerical methods. A set of laboratory tests was conducted to calibrate the parameters employed in the models. The modelling consisted of thermal, hydraulic and thermo-hydraulic analysis in which the significant thermo-hydraulic processes, parameters and features were identified. CODE B RIGHT was used for the finite element modelling and supplementary calculations were conducted with analytical methods. The main objective in this report is to improve understanding of the thermo-hydraulic processes and material properties that affect buffer behaviour in the Olkiluoto repository and to determine the parametric requirements of models for the accurate prediction of this behaviour. The analyses consisted of evaluating the influence of initial canister temperature and gaps in the buffer, and the role played by fractures and the rock mass located between fractures in supplying water for buffer and backfill saturation. In the thermo-hydraulic analysis, the primary processes examined were the effects of buffer drying near the canister on temperature evolution and the manner in which heat flow affects the buffer saturation process. Uncertainties in parameters and variations in the boundary conditions, modelling geometry and thermo-hydraulic phenomena were assessed with a sensitivity analysis. The material parameters, constitutive models, and assumptions made were carefully selected for all the modelling cases. The reference parameters selected for the simulations were compared and evaluated against laboratory measurements. The modelling results highlight the importance of understanding groundwater flow through the rock mass and from fractures in the rock in order to achieve reliable predictions regarding buffer saturation, since saturation times could range from a few years to tens of thousands of years depending on the hydrogeological
International Nuclear Information System (INIS)
Kokkoris, M.; Tsaris, A.; Misaelides, P.; Sokaras, D.; Lagoyannis, A.; Harissopulos, S.; Vlastou, R.; Papadopoulos, C.T.
2010-01-01
In the present work, new, differential cross-section values are presented for the nat K(p, p 0 ) reaction in the energy range E lab = 3000-5000 keV (with an energy step of 25 keV) and for detector angles between 140 o and 170 o (with an angular step of 10 o ). A qualitative discussion of the observed cross-section variations through the influence of strong, closely spaced resonances in the p + 39 K system is also presented. Information has also been extracted concerning the 39 K(p,α 0 ) reaction for E lab = 4000-5000 keV in the same angular range. As a result, more than ∼500 data points will soon be available to the scientific community through IBANDL (Ion Beam Analysis Nuclear Data Library - (http://www-nds.iaea.org/ibandl/)) and could thus be incorporated in widely used IBA algorithms (e.g. SIMNRA, WINDF, etc.) for potassium depth profiling at relatively high proton beam energies.
Extended x-ray absorption fine structure study of MnFeP0.56Si0.44 compound
International Nuclear Information System (INIS)
Li Ying-Jie; Haschaolu W; Wurentuya; Song Zhi-Qiang; Ou Zhi-Qiang; Tegus O; Nakai Ikuo
2015-01-01
The MnFeP 0.56 Si 0.44 compound is investigated by x-ray diffraction, magnetic measurements, and x-ray absorption fine structure spectroscopy. It crystallizes in Fe 2 P-type structure with the lattice parameters a = b = 5.9823(0) Å and c = 3.4551(1) Å and undergoes a first-order phase transition at the Curie temperature of 255 K. The Fe K edge and Mn K edge x-ray absorption fine structure spectra show that Mn atoms mainly reside at 3g sites, while 3f sites are occupied by Fe atoms. The distances between the absorbing Fe atom and the first and second nearest neighbor Fe atoms in a 3f-layer shift from 2.65 Å and 4.01 Å in the ferromagnetic state to 2.61 Å and 3.96 Å in the paramagnetic phase. On the other hand, the distance between the 3g-layer and 3f-layer changes a little as 2.66 Å–2.73 Å below the Curie temperature and 2.68 Å–2.75 Å above it. (paper)
International Nuclear Information System (INIS)
Yang, W.S.; Lee, C.H.
2008-01-01
Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT
Energy Technology Data Exchange (ETDEWEB)
Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)
2008-05-16
Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies
Energy Technology Data Exchange (ETDEWEB)
Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)
2005-07-01
At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.
PROF-DD, Generator of Multigroup Cross-Sections Library DDX for MORSE-DD, ANISN-DD, DOT-DD
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki; Ishiguro, Yukio
2002-01-01
1 - Description of program or function: The code system PROF-DD generates a multi-group double-differential cross section library DDX from evaluated data in ENDF/B-IV or ENDF/B-V format. The system consists of the following five modules: PROF-DDX is the main module of the system. It calculates the multigroup DDX and stores them on a master PDS file. MCFILEF generates a control file for PROF-DDX, which contains energy group and angle bin structures. SPINPTF prepares an input data file for PROF-DDX by combining the control file with other input data. DDXLIBMK edits a DDX library from the master PDS file for transport calculations. RESENDD performs resonance cross section and Doppler broadening calculations. 2 - Restrictions on the complexity of the problem: The numbers of energy groups and angle bins are less than 150 and 40, respectively
International Nuclear Information System (INIS)
Stamatelatos, M.G.; England, T.R.
1977-05-01
FPDCYS and FPSPEC are two FORTRAN computer programs used at the Los Alamos Scientific Laboratory (LASL), in conjunction with the CINDER-10 program, for calculating cumulative fission-product beta and/or gamma multigroup spectra in arbitrary energy structures, and for arbitrary neutron irradiation periods and cooling times. FPDCYS processes ENDF/B-IV fission-product decay energy data to generate multigroup beta and gamma spectra from individual ENDF/B-IV fission-product nuclides. FPSPEC further uses these spectra and the corresponding nuclide activities calculated by the CINDER-10 code to produce cumulative beta and gamma spectra in the same energy grids in which FPDCYS generates individual isotope decay spectra. The code system consisting of CINDER-10, FPDCYS, and FPSPEC has been used for comparisons with experimental spectra and continues to be used at LASL for generating spectra in special user-oriented group structures. 3 figures
Niang, Pape Momar; Huang, Zhiwei; Dulong, Virginie; Souguir, Zied; Le Cerf, Didier; Picton, Luc
2016-03-30
Several thermo-sensitive polyelectrolyte complexes were prepared by ionic self-association between an anionic polysaccharide (alginate) and a monocationic copolymer (polyether amine, Jeffamine®-M2005) with a 'Low Critical Solubility Temperature' (LCST). We show that electro-association must be established below the aggregation temperature of the free Jeffamine®, after which the organization of the system is controlled by the thermo-association of Jeffamine® that was previously electro-associated with the alginate. Evidence for this comes primarily from the rheology in the semi-dilute region. Electro- and thermo-associative behaviours are optimal at a pH corresponding to maximum ionization of both compounds (around pH 7). High ionic strength could prevent the electro-association. The reversibility of the transition is possible only at temperatures lower than the LCST of Jeffamine®. Similar behaviour has been obtained with carboxymethyl cellulose (CMC), which suggests that this behaviour can be observed using a range of anionic polyelectrolytes. In contrast, no specific properties have been found for pullulan, which is a neutral polysaccharide. Copyright © 2015 Elsevier Ltd. All rights reserved.
Fu, Li; Merabia, Samy; Joly, Laurent
2017-11-01
Thermo-osmotic and related thermophoretic phenomena can be found in many situations from biology to colloid science, but the underlying molecular mechanisms remain largely unexplored. Using molecular dynamics simulations, we measure the thermo-osmosis coefficient by both mechanocaloric and thermo-osmotic routes, for different solid-liquid interfacial energies. The simulations reveal, in particular, the crucial role of nanoscale interfacial hydrodynamics. For nonwetting surfaces, thermo-osmotic transport is largely amplified by hydrodynamic slip at the interface. For wetting surfaces, the position of the hydrodynamic shear plane plays a key role in determining the amplitude and sign of the thermo-osmosis coefficient. Finally, we measure a giant thermo-osmotic response of the water-graphene interface, which we relate to the very low interfacial friction displayed by this system. These results open new perspectives for the design of efficient functional interfaces for, e.g., waste-heat harvesting.
Fu, Li; Merabia, Samy; Joly, Laurent
2017-11-24
Thermo-osmotic and related thermophoretic phenomena can be found in many situations from biology to colloid science, but the underlying molecular mechanisms remain largely unexplored. Using molecular dynamics simulations, we measure the thermo-osmosis coefficient by both mechanocaloric and thermo-osmotic routes, for different solid-liquid interfacial energies. The simulations reveal, in particular, the crucial role of nanoscale interfacial hydrodynamics. For nonwetting surfaces, thermo-osmotic transport is largely amplified by hydrodynamic slip at the interface. For wetting surfaces, the position of the hydrodynamic shear plane plays a key role in determining the amplitude and sign of the thermo-osmosis coefficient. Finally, we measure a giant thermo-osmotic response of the water-graphene interface, which we relate to the very low interfacial friction displayed by this system. These results open new perspectives for the design of efficient functional interfaces for, e.g., waste-heat harvesting.
International Nuclear Information System (INIS)
Dejonghe, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.
1983-05-01
The THEMIS cross section processing system has been developped to produce punctual data for MONTE CARLO and coherent multigroup data for SN codes from ENDF/B. The THEMIS-4 data base has been generated from ENDF/B4 using the system and can be accessed by the 3-D Monte Carlo system TRIPOLI-2 and by the SN codes ANISN and DOT. An interpretation of ORNL fusion shielding benchmark is presented
International Nuclear Information System (INIS)
Hong, Ser Gi; Lee, Deokjung
2015-01-01
A highly accurate S 4 eigenfunction-based nodal method has been developed to solve multi-group discrete ordinate neutral particle transport problems with a linearly anisotropic scattering in slab geometry. The new method solves the even-parity form of discrete ordinates transport equation with an arbitrary S N order angular quadrature using two sub-cell balance equations and the S 4 eigenfunctions of within-group transport equation. The four eigenfunctions from S 4 approximation have been chosen as basis functions for the spatial expansion of the angular flux in each mesh. The constant and cubic polynomial approximations are adopted for the scattering source terms from other energy groups and fission source. A nodal method using the conventional polynomial expansion and the sub-cell balances was also developed to be used for demonstrating the high accuracy of the new methods. Using the new methods, a multi-group eigenvalue problem has been solved as well as fixed source problems. The numerical test results of one-group problem show that the new method has third-order accuracy as mesh size is finely refined and it has much higher accuracies for large meshes than the diamond differencing method and the nodal method using sub-cell balances and polynomial expansion of angular flux. For multi-group problems including eigenvalue problem, it was demonstrated that the new method using the cubic polynomial approximation of the sources could produce very accurate solutions even with large mesh sizes. (author)
International Nuclear Information System (INIS)
Chang, Jonghwa
2014-01-01
Today, we can use a computer cluster consist of a few hundreds CPUs with reasonable budget. Such computer system enables us to do detailed modeling of reactor core. The detailed modeling will improve the safety and the economics of a nuclear reactor by eliminating un-necessary conservatism or missing consideration. To take advantage of such a cluster computer, efficient parallel algorithms must be developed. Mechanical structure analysis community has studied the domain decomposition method to solve the stress-strain equation using the finite element methods. One of the most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multi-group diffusion problem in previous study. In this study, we report the result of recent modification to handle the three-dimensional subdomain partitioning, and the sub-domain multi-group problem. Modified FETI-DP algorithm has been successfully applied for the eigenvalue problem of multi-group neutron diffusion equation. The overall CPU time is decreasing as number of sub-domains (partitions) is increasing. However, there may be a limit in decrement due to increment of the number of primal points will increase the CPU time spent by the solution of the global equation. Even distribution of computational load (criterion a) is important to achieve fast computation. The subdomain partition can be effectively performed using suitable graph theory partition package such as MeTIS
Energy Technology Data Exchange (ETDEWEB)
Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
Today, we can use a computer cluster consist of a few hundreds CPUs with reasonable budget. Such computer system enables us to do detailed modeling of reactor core. The detailed modeling will improve the safety and the economics of a nuclear reactor by eliminating un-necessary conservatism or missing consideration. To take advantage of such a cluster computer, efficient parallel algorithms must be developed. Mechanical structure analysis community has studied the domain decomposition method to solve the stress-strain equation using the finite element methods. One of the most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multi-group diffusion problem in previous study. In this study, we report the result of recent modification to handle the three-dimensional subdomain partitioning, and the sub-domain multi-group problem. Modified FETI-DP algorithm has been successfully applied for the eigenvalue problem of multi-group neutron diffusion equation. The overall CPU time is decreasing as number of sub-domains (partitions) is increasing. However, there may be a limit in decrement due to increment of the number of primal points will increase the CPU time spent by the solution of the global equation. Even distribution of computational load (criterion a) is important to achieve fast computation. The subdomain partition can be effectively performed using suitable graph theory partition package such as MeTIS.
Vortex-glass state in the isovalent optimally doped pnictide superconductor BaFe2(As0.68P0.32)2
Salem-Sugui, S., Jr.; Mosqueira, J.; Alvarenga, A. D.; Sóñora, D.; Crisan, A.; Ionescu, A. M.; Sundar, S.; Hu, D.; Li, S.-L.; Luo, H.-Q.
2017-05-01
We report on isochamp magneto-resistivity and ac susceptibility curves obtained in a high-quality single crystal of the isovalent optimally doped pnictide BaFe2(As{}0.68P{}0.32)2 with superconducting temperature T c = 27.8 K for H∥c-axis. Plots of the logarithmic derivative of the resistivity curves allowed the identification of a vortex-glass (VG) phase and to obtain the values of the critical glass temperature T g, the temperature T * marking the transition to the liquid phase and of the critical exponent s. The presence of the VG phase is confirmed by detailed measurements of the third harmonic signal of the ac magnetic susceptibility. The modified VG model was successfully applied to the data allowing the obtention of the temperature independent VG activation energy U b . The activation energy U 0 obtained from the Arrhenius plots in the flux-flow region are compared with U b and with U 0 obtained from flux-creep measurements on a M(H) isothermal in the same sample. A phase diagram of the studied sample is constructed showing the T g glass line, the T * line representing a transition (melting) to the liquid phase, the mean field temperature T c(H) line and the H p line obtained from the peaks in isothermal critical current, J c(H) curves, which are explained in terms of a softening of the vortex lattice. The glass line was fitted by a theory presented in the literature which considers the effect of disorder.
Nuclear materials thermo-physical property database and property analysis using the database
International Nuclear Information System (INIS)
Jeong, Yeong Seok
2002-02-01
It is necessary that thermo-physical properties and understand of nuclear materials for evaluation and analysis to steady and accident states of commercial and research reactor. In this study, development of nuclear materials thermo-properties database and home page. In application of this database, it is analyzed of thermal conductivity, heat capacity, enthalpy, and linear thermal expansion of fuel and cladding material and compared thermo-properties model in nuclear fuel performance evaluation codes with experimental data in database. Results of compare thermo-property model of UO 2 fuel and cladding major performance evaluation code, both are similar
Kunnikuruvan, Sooraj; Parandekar, Priya V; Prakash, Om; Tsotsis, Thomas K; Nair, Nisanth N
2016-06-02
The growing requisite for materials having high thermo-oxidative stability makes the design and development of high performance materials an active area of research. Fluorination of the polymer backbone is a widely applied strategy to improve various properties of the polymer, most importantly the thermo-oxidative stability. Many of these fluorinated polymers are known to have thermo-oxidative stability up to 700 K. However, for space and aerospace applications, it is important to improve its thermo-oxidative stability beyond 700 K. Molecular-level details of the thermo-oxidative degradation of such polymers can provide vital information to improve the polymer. In this spirit, we have applied quantum mechanical and microkinetic analysis to scrutinize the mechanism and kinetics of the thermo-oxidative degradation of a fluorinated polymer with phenylethenyl end-cap, HFPE. This study gives an insight into the thermo-oxidative degradation of HFPE and explains most of the experimental observations on the thermo-oxidative degradation of this polymer. Thermolysis of C-CF3 bond in the dianhydride component (6FDA) of HFPE is found to be the rate-determining step of the degradation. Reaction pathways that are responsible for the experimentally observed weight loss of the polymer is also scrutinized. On the basis of these results, we propose a modification of HFPE polymer to improve its thermo-oxidative stability.
International Nuclear Information System (INIS)
Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto
2015-01-01
Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon
Energy Technology Data Exchange (ETDEWEB)
Oliva, Amaury M.; Filho, Hermes A.; Silva, Davi M.; Garcia, Carlos R., E-mail: aoliva@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: davijmsilva@yahoo.com.br, E-mail: cgh@instec.cu [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional; Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba)
2017-07-01
In this paper, we propose a numerical methodology for the development of a method of the spectral nodal class that will generate numerical solutions free from spatial truncation errors. This method, denominated Spectral Deterministic Method (SDM), is tested as an initial study of the solutions (spectral analysis) of neutron transport equations in the discrete ordinates (S{sub N}) formulation, in one-dimensional slab geometry, multigroup approximation, with linearly anisotropic scattering, considering homogeneous and heterogeneous domains with fixed source. The unknowns in the methodology are the cell-edge, and cell average angular fluxes, the numerical values calculated for these quantities coincide with the analytic solution of the equations. These numerical results are shown and compared with the traditional ne- mesh method Diamond Difference (DD) and the coarse-mesh method spectral Green's function (SGF) to illustrate the method's accuracy and stability. The solution algorithms problems are implemented in a computer simulator made in C++ language, the same that was used to generate the results of the reference work. (author)
TWOTRAN-2, 2-D Multigroup Transport in X-Y, R-Z, R-Theta Geometry with Anisotropic Scattering
International Nuclear Information System (INIS)
Lathrop, K.D.; Brinkley, F.W.
1995-01-01
1 - Description of problem or function: TWOTRAN2 solves the two-dimensional multigroup transport equation in (x,y), (r,theta), and (r,z) geometries. Both regular and adjoint, inhomogeneous and homogeneous (k eff and eigenvalue searches) problems subject to vacuum, reflective, periodic, white or input-specified boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. 2 - Method of solution: The discrete ordinates approximation for the angular variable is used in finite difference form which is solved with the central (diamond) difference approximation. Negative fluxes are eliminated by a local set-to zero and correct algorithm. Standard inner (within-group) and outer iterative cycles are accelerated by a coarse-mesh re-balancing on a coarse mesh which may be independent of the material mesh. 3 - Restrictions on the complexity of the problem: Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXLEN can be accommodated. On IBM machines, TWOTRAN2 will execute in the 4-byte mode so that any combination of problem parameters leading to a container array less than MAXLEN can be accommodated. MAXLEN can be several hundred thousand and most problems can be core-contained. On the CDC machines MAXLEN can be slightly greater than 40,000 words and peripheral storage is used for most group-dependent data
Merz, Erin L.; Malcarne, Vanessa L.; Roesch, Scott C.; Riley, Natasha; Sadler, Georgia Robins
2014-01-01
Depression is a significant problem for ethnic minorities that remains understudied partly due to a lack of strong measures with established psychometric properties. One screening tool, the Patient Health Questionnaire-9 (PHQ-9), which was developed for use in primary care has also gained popularity in research settings. The reliability and validity of the PHQ-9 has been well established among predominantly Caucasian samples, in addition to many minority groups. However, there is little evidence regarding its utility among Hispanic Americans, a large and growing cultural group in the United States. In this study, we investigated the reliability and structural validity of the PHQ-9 in Hispanic American women. A community sample of 479 Latina women from southern California completed the PHQ-9 in their preferred language of English or Spanish. Cronbach’s alphas suggested that there was good internal consistency for both the English- and Spanish-language versions. Structural validity was investigated using multigroup confirmatory factor analysis (CFA). Results support a similar one-factor structure with equivalent response patterns and variances among English- and Spanish-speaking Latinas. These results suggest that the PHQ-9 can be used with confidence in both English and Spanish versions to screen Latinas for depression. PMID:21787063
International Nuclear Information System (INIS)
Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.
1989-03-01
A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)
Energy Technology Data Exchange (ETDEWEB)
Kim, Myung Hyun [Kyunghee University, Seoul (Korea, Republic of)
1996-07-01
A development project of 3-dimensional kinetics code for ALMR has three level of works. In the first level, a multi-group, nodal kinetics code for the HEX-Z geometry has been developed. A code showed very good results for the static analysis as well as for the kinetics problems. At the second level, a core thermal-hydraulic analysis code was developed for the temperature feedback calculation in ALMR transients analysis. This code is coupled with kinetics code. A sodium property table was programmed and tested to the KAERI data and thermal feedback model was developed and coupled in code. Benchmarking of T/H calculation has been performed and showed fairly good results. At the third level of research work, reactivity feedback model for structure thermal expansion is developed and added to the code. At present, basic model was studied. However, code development in now on going. Benchmarking of this model developed can not be done because of lack of data. 31 refs., 17 tabs., 38 figs. (author)
International Nuclear Information System (INIS)
Ceolin, Celina
2010-01-01
The objective of this work is to obtain an analytical solution of the neutron diffusion kinetic equation in one-dimensional cartesian geometry, to monoenergetic and multigroup problems. These equations are of the type stiff, due to large differences in the orders of magnitude of the time scales of the physical phenomena involved, which make them difficult to solve. The basic idea of the proposed method is applying the spectral expansion in the scalar flux and in the precursor concentration, taking moments and solving the resulting matrix problem by the Laplace transform technique. Bearing in mind that the equation for the precursor concentration is a first order linear differential equation in the time variable, to enable the application of the spectral method we introduce a fictitious diffusion term multiplied by a positive value which tends to zero. This procedure opened the possibility to find an analytical solution to the problem studied. We report numerical simulations and analysis of the results obtained with the precision controlled by the truncation order of the series. (author)
International Nuclear Information System (INIS)
Hill, T. R.; Reed, W. H.
1980-01-01
1 - Description of problem or function: TIMEX solves the time- dependent, one-dimensional multigroup transport equation with delayed neutrons in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous problems subject to vacuum, reflective, periodic, white, albedo or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. 2 - Method of solution: The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. Negative fluxes are eliminated by a local set-to-zero and correct algorithm. The time variable is differenced by an explicit technique that is unconditionally stable so that arbitrarily large time-steps can be taken. Two acceleration methods, exponential extrapolation and re-balance, are utilized to improve the accuracy of the time differencing scheme. 3 - Restrictions on the complexity of the problem: Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXCOR can be accommodated. In addition, the CDC version permits the use of extended core storage less than MAXECS
Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi
1993-02-01
A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)
Energy Technology Data Exchange (ETDEWEB)
Ossant, G. [Societe Syrec (France)
1997-12-31
The main principles, performances and constraints of the various types of ground source thermo-pumps for individual houses, i.e. ground/ground thermo-pumps, glycol water/water thermo-pumps and ground/water thermo-pumps are reviewed, and their energy consumptions are discussed. The design and operating conditions of a reverse ground source thermo-pump (Syrec) for space heating and air conditioning through a hot and cold floor system and a Syrec ground source thermo-pump for water heating, are presented
Thermal and thermo-mechanical simulation of laser assisted machining
International Nuclear Information System (INIS)
Germain, G.; Dal Santo, P.; Lebrun, J. L.; Bellett, D.; Robert, P.
2007-01-01
Laser Assisted Machining (LAM) improves the machinability of materials by locally heating the workpiece just prior to cutting. The heat input is provided by a high power laser focused several millimeters in front of the cutting tool. Experimental investigations have confirmed that the cutting force can be decreased, by as much as 40%, for various materials (tool steel, titanium alloys and nickel alloys). The laser heat input is essentially superficial and results in non-uniform temperature profiles within the depth of the workpiece. The temperature field in the cutting zone is therefore influenced by many parameters. In order to understand the effect of the laser on chip formation and on the temperature fields in the different deformation zones, thermo-mechanical simulation were undertaken. A thermo-mechanical model for chip formation with and without the laser was also undertaken for different cutting parameters. Experimental tests for the orthogonal cutting of 42CrMo4 steel were used to validate the simulation via the prediction of the cutting force with and without the laser. The thermo-mechanical model then allowed us to highlight the differences in the temperature fields in the cutting zone with and without the laser. In particular, it was shown that for LAM the auto-heating of the material in the primary shear zone is less important and that the friction between the tool and chip also generates less heat. The temperature fields allow us to explain the reduction in the cutting force and the resulting residual stress fields in the workpiece
Thermo-electro-chemical storage (TECS) of solar energy
International Nuclear Information System (INIS)
Wenger, Erez; Epstein, Michael; Kribus, Abraham
2017-01-01
Highlights: • A solar plant with thermally regenerative battery unifies energy conversion and storage. • Storage is a flow battery with thermo-chemical charging and electro-chemical discharging. • Sodium-sulfur and zinc-air systems are investigated as candidate storage materials. • Theoretical solar to electricity efficiencies of over 60% are predicted. • Charging temperature can be lowered with hybrid carbothermic reduction. - Abstract: A new approach for solar electricity generation and storage is proposed, based on the concept of thermally regenerative batteries. Concentrated sunlight is used for external thermo-chemical charging of a flow battery, and electricity is produced by conventional electro-chemical discharge of the battery. The battery replaces the steam turbine, currently used in commercial concentrated solar power (CSP) plants, potentially leading to much higher conversion efficiency. This approach offers potential performance, cost and operational advantages compared to existing solar technologies, and to existing storage solutions for management of an electrical grid with a significant contribution of intermittent solar electricity generation. Here we analyze the theoretical conversion efficiency for new thermo-electro-chemical storage (TECS) plant schemes based on the electro-chemical systems of sodium-sulfur (Na-S) and zinc-air. The thermodynamic upper limit of solar to electricity conversion efficiency for an ideal TECS cycle is about 60% for Na-S at reactor temperature of 1550 K, and 65% for the zinc-air system at 1750 K, both under sunlight concentration of 3000. A hybrid process with carbothermic reduction in the zinc-air system reaches 60% theoretical efficiency at the more practical conditions of reaction temperature <1200 K and concentration <1000. Practical TECS plant efficiency, estimated from these upper limits, may then be much higher compared to existing solar electricity technologies. The technical and economical
Filingeri, Davide; Fournet, Damien; Hodder, Simon; Havenith, George
2014-10-15
Sensing skin wetness is linked to inputs arising from cutaneous cold-sensitive afferents. As thermosensitivity to cold varies significantly across the torso, we investigated whether similar regional differences in wetness perception exist. We also investigated the regional differences in thermal pleasantness and whether these sensory patterns are influenced by ambient temperature. Sixteen males (20 ± 2 yr) underwent a quantitative sensory test under thermo-neutral [air temperature (Tair) = 22°C; relative humidity (RH) = 50%] and warm conditions (Tair = 33°C; RH = 50%). Twelve regions of the torso were stimulated with a dry thermal probe (25 cm(2)) with a temperature of 15°C below local skin temperature (Tsk). Variations in Tsk, thermal, wetness, and pleasantness sensations were recorded. As a result of the same cold-dry stimulus, the skin-cooling response varied significantly by location (P = 0.003). The lateral chest showed the greatest cooling (-5 ± 0.4°C), whereas the lower back showed the smallest (-1.9 ± 0.4°C). Thermal sensations varied significantly by location and independently from regional variations in skin cooling with colder sensations reported on the lateral abdomen and lower back. Similarly, the frequency of perceived skin wetness was significantly greater on the lateral and lower back as opposed to the medial chest. Overall wetness perception was slightly higher under warm conditions. Significantly more unpleasant sensations were recorded when the lateral abdomen and lateral and lower back were stimulated. We conclude that humans present regional differences in skin wetness perception across the torso, with a pattern similar to the regional differences in thermosensitivity to cold. These findings indicate the presence of a heterogeneous distribution of cold-sensitive thermo-afferent information. Copyright © 2014 the American Physiological Society.
The Thermos program for nuclear reactors specialized in district heating
International Nuclear Information System (INIS)
Lerouge, B.
1976-01-01
Many studies have been made in France on the use of nuclear heat for district heating. After a brief account of the problems raised by the use of thermal waste from big nuclear power stations, the quantitative and qualitative needs of heating networks are analyzed and the Thermos project described. This is a very robust reactor of the pool type, with an output of 100MW, supplying low-pressure water at 100 deg C. The advantages from the aspects of safety and economy are described, and the present state of the project and its possible developments summarized [fr
Thermo field dynamics: a quantum field theory at finite temperature
International Nuclear Information System (INIS)
Mancini, F.; Marinaro, M.; Matsumoto, H.
1988-01-01
A brief review of the theory of thermo field dynamics (TFD) is presented. TFD is introduced and developed by Umezawa and his coworkers at finite temperature. The most significant concept in TFD is that of a thermal vacuum which satisfies some conditions denoted as thermal state conditions. The TFD permits to reformulate theories at finite temperature. There is no need in an additional principle to determine particle distributions at T ≠ 0. Temperature and other macroscopic parameters are introduced in the definition of the vacuum state. All operator formalisms used in quantum field theory at T=0 are preserved, although the field degrees of freedom are doubled. 8 refs
Prediction of thermo-physical properties of liquid formulated products
DEFF Research Database (Denmark)
Mattei, Michele; Conte, Elisa; Kontogeorgis, Georgios
2013-01-01
The objective of this chapter is to give an overview of the models, methods and tools that may be used for the estimation of liquid formulated products. First a classification of the products is given and the thermo-physical properties needed to represent their functions are listed. For each...... property, a collection of the available models are presented according to the property type and the model type. It should be noted, however, that the property models considered or highlighted in this chapter are only examples and are not necessarily the best and most accurate for the corresponding property....
Thermo-osmosis in Membrane Systems: A Review
Barragán, V. María; Kjelstrup, Signe
2017-06-01
We give a first review of experimental results for a phenomenon little explored in the literature, namely thermal osmosis or thermo-osmosis. Such systems are now getting increased attention because of their ability to use waste heat for separation purposes. We show that this volume transport of a solution or a pure liquid caused by a temperature difference across a membrane can be understood as a property of the membrane system, i. e. the membrane with its adjacent solutions. We present experimental values found in the literature of thermo-osmotic coefficients of neutral and hydrophobic as well as charged and hydrophilic membranes, with water and other permeant fluids as well as electrolyte solutions. We propose that the coefficient can be qualitatively explained by a formula that contains the entropy of adsorption of permeant into the membrane, the hydraulic permeability, and a factor that depends on the interface resistance to heat transfer. A variation in the entropy of adsorption with hydrophobic/hydrophilic membranes and structure breaking/structure making cations could then explain the sign of the permeant flux. Systematic experiments in the field are lacking and we propose an experimental program to mend this situation.
Effects of physical properties on thermo-fluids cavitating flows
Chen, T. R.; Wang, G. Y.; Huang, B.; Li, D. Q.; Ma, X. J.; Li, X. L.
2015-12-01
The aims of this paper are to study the thermo-fluid cavitating flows and to evaluate the effects of physical properties on cavitation behaviours. The Favre-averaged Navier-Stokes equations with the energy equation are applied to numerically investigate the liquid nitrogen cavitating flows around a NASA hydrofoil. Meanwhile, the thermodynamic parameter Σ is used to assess the thermodynamic effects on cavitating flows. The results indicate that the thermodynamic effects on the thermo-fluid cavitating flows significantly affect the cavitation behaviours, including pressure and temperature distribution, the variation of physical properties, and cavity structures. The thermodynamic effects can be evaluated by physical properties under the same free-stream conditions. The global sensitivity analysis of liquid nitrogen suggests that ρv, Cl and L significantly influence temperature drop and cavity structure in the existing numerical framework, while pv plays the dominant role when these properties vary with temperature. The liquid viscosity μl slightly affects the flow structure via changing the Reynolds number Re equivalently, however, it hardly affects the temperature distribution.
Determination and Scaling of Thermo Acoustic Characteristics of Premixed Flames
Directory of Open Access Journals (Sweden)
P. R. Alemela
2010-06-01
Full Text Available The paper investigates the determination and the scaling of thermo acoustical characteristics of lean premixed flames as used in gas turbine combustion systems. In the first part, alternative methods to characterize experimentally the flame dynamics are outlined and are compared on the example of a scaled model of an industrial gas turbine burner. Transfer matrix results from the most general direct method are contrasted with data obtained from the hybrid method, which is based on Rankine-Hugoniot relations and the experimental flame transfer function obtained from OH*-chemiluminescence measurements. Also the new network model based regression method is assessed, which is based on a n – τ – σ dynamic flame model. The results indicate very good consistency between the three techniques, providing a global check of the methods/tools used for analyzing the thermo acoustic mechanisms of flames. In the second part, scaling rules are developed that allow to calculate the dynamic flame characteristics at different operation points. Towards this a geometric flame length model is formulated. Together with the other operational data of the flame it provides the dynamic flame model parameters at these points. The comparison between the measured and modeled flame lengths as well as the n – τ – σ parameters shows an excellent agreement.
Operating experience with the Harwell thermo-mechanical generators
International Nuclear Information System (INIS)
Cooke-Yarborough, E.H.
1980-06-01
The Stirling-cycle thermo-mechanical generator (TMG) provides small amounts of electrical power continuously over long periods, while requiring much less fuel than other power sources running from hydrocarbon fuel or radio-isotopes. Two of these 25-watt generators, fuelled by propane, have been used to power the UK National Buoy on two successive missions. A total of more than three years experience at sea has now been accumulated. In addition, a 60-watt version has provided the power for a major lighthouse for more than a year. An early development version of the Thermo-mechanical Generator, adapted to run from the heat of a radio-isotope source, was loaded with strontium 90 titanate in October 1974 and has run continuously in the laboratory ever since. The improvements and changes found necessary in the course of 90,000 generator-hours of running time are described, and the improvements in operational performance and reliability which have resulted are outlined. (author)
Thermo-Chemical Conversion of Microwave Activated Biomass Mixtures
Barmina, I.; Kolmickovs, A.; Valdmanis, R.; Vostrikovs, S.; Zake, M.
2018-05-01
Thermo-chemical conversion of microwave activated wheat straw mixtures with wood or peat pellets is studied experimentally with the aim to provide more effective application of wheat straw for heat energy production. Microwave pre-processing of straw pellets is used to provide a partial decomposition of the main constituents of straw and to activate the thermo-chemical conversion of wheat straw mixtures with wood or peat pellets. The experimental study includes complex measurements of the elemental composition of biomass pellets (wheat straw, wood, peat), DTG analysis of their thermal degradation, FTIR analysis of the composition of combustible volatiles entering the combustor, the flame temperature, the heat output of the device and composition of the products by comparing these characteristics for mixtures with unprocessed and mw pre-treated straw pellets. The results of experimental study confirm that mw pre-processing of straw activates the thermal decomposition of mixtures providing enhanced formation of combustible volatiles. This leads to improvement of the combustion conditions in the flame reaction zone, completing thus the combustion of volatiles, increasing the flame temperature, the heat output from the device, the produced heat energy per mass of burned mixture and decreasing at the same time the mass fraction of unburned volatiles in the products.
Development of thermo-plastic heating and compaction facility
International Nuclear Information System (INIS)
Ko, Dae Hak; Lim, Suk Nam
1998-01-01
Low- and intermediate-level radioactive wastes consist of spent resin, spent filter, concentrated waste and dry active waste(DAW) and they are solidified or packaged into drums or high integrated containers(HICs). DAWs occupy 50 percent of all low- and intermediate-level radioactive wastes generated from nuclear power plants in Korea. Incinerable wastes in the DAWs are about 60 percent. Therefore, it is very important for us to reduce the volume of incinerable wastes in DAWs. Experience of supercompaction turned out that thermo-plastic wastes have a swelling effect after supercompaction process due to their repulsive power. And the thermo-plastic heating and compaction facility has been developed by KEPCO. In conclusion, heating and compaction facility can reduce the volume of DAWs as well as upgrade the quality of treated wastes, because the swelling effect by repulsive power after compaction is removed, final wastes form the shape of block and they have no free-standing water in the wastes. Plan for practical use is that this facility will be installed in other nuclear power plants in Korea in 1999. (Cho, G. S.). 1 tab., 2 figs
Thermo-Fluid Verification of Fuel Column with Crossflow Gap
International Nuclear Information System (INIS)
Lee, Sung Nam; Tak, Nam Il; Kim, Min Hwan; Noh, Jae Man
2013-01-01
Korea Atomic Energy Research Institute (KAERI) has been developing thermal-hydraulic code to design a safe and effective VHTR. Core reliable Optimization and Network thermo-fluid Analysis (CORONA) is a code that solves the fluid region as 1-D and the solid domain as 3-D. The postulated event is modeled to secure safety during design process. The reactor core of VHTR is piled with multi-fuel block layers. The helium gas goes through coolant channel holes after distributed from upper plenum. The fuel blocks are irradiated during operation and there might be cross gaps between blocks. These cross gaps change the passage of coolant channels and could affect the temperature of fuel compact. Therefore, two types of single fuel assembly (i. e., standard and Reserved Shutdown Control (RSC) hole fuel assemblies) were investigated in this study. The CORONA, thermo-fluid analysis code, has been developing to compute the reactor core of VHTR. Crossflow model was applied to predict temperature and flow distribution between fuel blocks in this study. The calculated results are compared with the data of commercial software, CFX. The temperature variations along the axial direction well agree for both standard / RSC fuel assemblies. The flow redistribution due to crossflow matches well. The hot spot temperature and locations might differ depending on the cross gap size. This research will be done in detail for further study
Thermo-Optic Characterization of Silicon Nitride Resonators for Cryogenic Photonic Circuits
Elshaari, A.W.A.; Esmaeil Zadeh, I.; Jöns, K.D.; Zwiller, Val
2016-01-01
In this paper, we characterize the Thermo-optic properties of silicon nitride ring resonators between 18 and 300 K. The Thermo-optic coefficients of the silicon nitride core and the oxide cladding are measured by studying the temperature dependence of the resonance wavelengths. The resonant modes
Enhanced pathway efficiency of Saccharomyces cerevisiae by introducing thermo-tolerant devices.
Liu, Yueqin; Zhang, Genli; Sun, Huan; Sun, Xiangying; Jiang, Nisi; Rasool, Aamir; Lin, Zhanglin; Li, Chun
2014-10-01
In this study, thermo-tolerant devices consisting of heat shock genes from thermophiles were designed and introduced into Saccharomyces cerevisiae for improving its thermo-tolerance. Among ten engineered thermo-tolerant yeasts, T.te-TTE2469, T.te-GroS2 and T.te-IbpA displayed over 25% increased cell density and 1.5-4-fold cell viability compared with the control. Physiological characteristics of thermo-tolerant strains revealed that better cell wall integrity, higher trehalose content and enhanced metabolic energy were preserved by thermo-tolerant devices. Engineered thermo-tolerant strain was used to investigate the impact of thermo-tolerant device on pathway efficiency by introducing β-amyrin synthesis pathway, showed 28.1% increased β-amyrin titer, 28-35°C broadened growth temperature range and 72h shortened fermentation period. The results indicated that implanting heat shock proteins from thermophiles to S. cerevisiae would be an efficient approach to improve its thermo-tolerance. Copyright © 2014 Elsevier Ltd. All rights reserved.
Fernandes, T.; Klaasse Bos, G.J.; Zeeman, G.; Sanders, J.P.M.; Lier, van J.B.
2009-01-01
The effects of different thermo-chemical pre-treatment methods were determined on the biodegradability and hydrolysis rate of lignocellulosic biomass. Three plant species, hay, straw and bracken were thermo-chemically pre-treated with calcium hydroxide, ammonium carbonate and maleic acid. After
Study of the structural integrity of thermo-wells. Application to Class I components
International Nuclear Information System (INIS)
Gavilan Moreno, C. J.
2010-01-01
This paper provides a methodology to determine a thermo-well failure. The practical application will be made on a thermo-well in Cofrentes Nuclear Power Plant. This will be designed by the existence of a spare one and it will be determined the eigenfrequencies, the vortex emission frequencies in the flow, the susceptibility to fatigue, the loads, etc.
Directory of Open Access Journals (Sweden)
Ameer Khusro
2016-07-01
Conclusions: The cellulase-free xylanase showed an alkali-tolerant and thermo-stable property with potentially applicable nature at industrial scale. This statistical approach established a major contribution in enzyme production from the isolate by optimizing independent factors and represents a first reference on the enhanced production of thermo-alkali stable cellulase-free xylanase from B. tequilensis.
Thermo-mechanical properties of SOFC components investigated by a combined method
DEFF Research Database (Denmark)
Teocoli, Francesca; Esposito, Vincenzo; Ramousse, Severine
, and differential thermo-mechanical behavior at each layer. The combination of such factors can have a critical effect on the final shape and microstructure, and on the mechanical integrity. Thermo-mechanical properties and sintering mechanisms of important SOFC materials (CGO, YSZ, ScYSZ) were systematically...
Jin, Shuguang; Zhang, Guangming; Zhang, Panyue; Li, Fan; Fan, Shiyang; Li, Juan
2016-04-01
To improve the reducing sugar production from catalpa sawdust, thermo-chemical pretreatments were examined and the chemicals used including NaOH, Ca(OH)2, H2SO4, and HCl. The hemicellulose solubilization and cellulose crystallinity index (CrI) were significantly increased after thermo-alkaline pretreatments, and the thermo-Ca(OH)2 pretreatment showed the best improvement for reducing sugar production comparing to other three pretreatments. The conditions of thermo-Ca(OH)2 pretreatment and enzymatic hydrolysis were systematically optimized. Under the optimal conditions, the reducing sugar yield increased by 1185.7% comparing to the control. This study indicates that the thermo-Ca(OH)2 pretreatment is ideal for the saccharification of catalpa sawdust and that catalpa sawdust is a promising raw material for biofuel. Copyright © 2016 Elsevier Ltd. All rights reserved.
Analysis of Thermo-Acoustic Emission from Damage in Composite Laminates under Thermal Cyclic Loading
International Nuclear Information System (INIS)
Kim, Young Bok; Min, Dae Hong; Lee, Deok Bo; Choi, Nak Sam
2001-01-01
An investigation on nondestructive evaluation of thermal stress-reduced damage in the composite laminates (3mm in thickness and [+45 6 /-45 6 ] S lay-up angles) has been performed using the thermo-acoustic emission technique. Reduction of thermo-AE events due to repetitive thermal load cycles showed a Kaiser effect. An analysis of the thermo-AE behavior determined the stress free temperature of composite laminates. Fiber fracture and matrix cracks were observed using the optical microscopy, scanning electron microscopy and ultrasonic C-sean. Short-Time Fourier Transform of thermo-AE signals offered the time-frequency characteristics which might classify the thermo-AE as three different types to estimate the damage processes of the composites
Thermo-emf of cermet films based on rare earth borides
International Nuclear Information System (INIS)
Islamgaliev, R.K.; Zyrin, A.V.; Shulishova, O.I.; Shcherbak, I.A
1987-01-01
Thermo-emf and electric conductivity of granulated films which contain a solid solution of europium and praseodymium borides Eu 0.5 Pr 0.5 B 6 as a conducting phase, and glass-crystal binder on the base of alummomagnesial fluosilicates as a dielectric phase are studied within the temperature range of 100-1100 K. Thermo-emf of films has a negative sign within the temperature range of 100-500 K and does not exceed 5 μkV/K according to the absolute value which is close to the value of the conducting phase thermo-emf. A negative sign and a small value of thermo-emf are indicative of the charge transfer in granulated films by electrons. Contribution of each of the components into the general thermo-emf is different at high temperatures in different temperature ranges and depends on the individual physico-chemical properties of the used materials
MINX, Multigroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX
International Nuclear Information System (INIS)
Soran, P.D.; MacFarlane, R.E.; Harris, D.R.; LaBauve, R.J.; Hendricks, J.S.; Kidman, R.B.; Weisbin, C.R.; White, J.E.
1977-01-01
1 - Description of problem or function: MINX calculates fine-group averaged infinitely diluted cross sections and self-shielding factors from ENDF/B-IV data. Its primary purpose is to generate a pseudo-composition-independent multigroup library which is input to the SPHINX space-energy collapse program (2) (PSR-0129) through standard CCCC-III (8) interfaces. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX (5) (NESC0388) and ENDRUN (9) and the high-order group-to-group transfer matrices of SUPERTOG (10) (PSR-0013) and ETOG (11). Fine group energy boundaries, Legendre expansion order, gross spectral shape component (in the Bondarenko flux model), temperatures and dilutions can all be used specifically. 2 - Method of solution: Infinitely dilute, un-broadened point cross sections are obtained from resolved resonance parameters using a modified version of the RESEND program (3) (NESC0465). The SIGMA1 (4) (IAEA0854) kernel-broadening method is used to Doppler broaden and thin the tabulated linearized pointwise cross sections at 0 K (outside of the unresolved energy region). Effective temperature- dependent self-shielded pointwise cross sections are derived from the formulation in the ETOX code. The primary modification to the ETOX algorithm is associated with the numerical quadrature scheme used to establish the mean values of the fluctuation intervals. The selection of energy mesh points, at which the effective cross sections are calculated, has been modified to include the energy points given in the ENDF/B file or, if the energy-independent formalism was employed, points at half-lethargy intervals. Infinitely dilute group cross sections and self-shielding factors are generated using the Bondarenko flux weighting model with the gross spectral shape under user control. The integral over energy for each group is divided into a set of panels defined by the union of the grid points describing the total cross section, the
International Nuclear Information System (INIS)
Ritchie, A.I.M.; Wilson, D.J.
1984-12-01
A multigroup diffusion code has been used to predict the count rate from a neutron moisture meter for a range of values of soil water content ω, thermal neutron absorption cross section Ssub(a) (defined as Σsub(a)/rho) of the soil matrix and soil matrix density rho. Two dimensions adequately approximated the geometry of the source, detector and soil surrounding the detector. Seven energy groups, the data for which were condensed from 128 group data set over the neutron energy spectrum appropriate to the soil-water mixture under study, proved adequate to describe neutron slowing-down and diffusion. The soil-water mixture was an SiO 2 →water mixture, with the absorption cross section of SiO 2 increased to cover the range of Σsub(a) required. The response to changes in matrix density is, in general, linear but the response to changes in water content is not linear over the range of parameter values investigated. Tabular results are presented which allow interpolation of the response for a particular ω, Ssub(a) and rho. It is shown that R(ω, Ssub(a), rho) rho M(Ssub(a)) + C(ω) is a crude representation of the response over a very limited range of variation of ω, and Ssub(a). As the response is a slowly varying function of rho, Ssub(a) and ω, a polynomial fit will provide a better estimate of the response for values of rho, Ssub(a) and ω not tabulated
International Nuclear Information System (INIS)
Lozano, Juan-Andres; Garcia-Herranz, Nuria; Ahnert, Carol; Aragones, Jose-Maria
2008-01-01
In this work we address the development and implementation of the analytic coarse-mesh finite-difference (ACMFD) method in a nodal neutron diffusion solver called ANDES. The first version of the solver is implemented in any number of neutron energy groups, and in 3D Cartesian geometries; thus it mainly addresses PWR and BWR core simulations. The details about the generalization to multigroups and 3D, as well as the implementation of the method are given. The transverse integration procedure is the scheme chosen to extend the ACMFD formulation to multidimensional problems. The role of the transverse leakage treatment in the accuracy of the nodal solutions is analyzed in detail: the involved assumptions, the limitations of the method in terms of nodal width, the alternative approaches to implement the transverse leakage terms in nodal methods - implicit or explicit -, and the error assessment due to transverse integration. A new approach for solving the control rod 'cusping' problem, based on the direct application of the ACMFD method, is also developed and implemented in ANDES. The solver architecture turns ANDES into an user-friendly, modular and easily linkable tool, as required to be integrated into common software platforms for multi-scale and multi-physics simulations. ANDES can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. The verification and performance of the solver are demonstrated using both proof-of-principle test cases and well-referenced international benchmarks
International Nuclear Information System (INIS)
Van Geemert, Rene
2008-01-01
For satisfaction of future global customer needs, dedicated efforts are being coordinated internationally and pursued continuously at AREVA NP. The currently ongoing CONVERGENCE project is committed to the development of the ARCADIA R next generation core simulation software package. ARCADIA R will be put to global use by all AREVA NP business regions, for the entire spectrum of core design processes, licensing computations and safety studies. As part of the currently ongoing trend towards more sophisticated neutronics methodologies, an SP 3 nodal transport concept has been developed for ARTEMIS which is the steady-state and transient core simulation part of ARCADIA R . For enabling a high computational performance, the SP N calculations are accelerated by applying multi-level coarse mesh re-balancing. In the current implementation, SP 3 is about 1.4 times as expensive computationally as SP 1 (diffusion). The developed SP 3 solution concept is foreseen as the future computational workhorse for many-group 3D pin-by-pin full core computations by ARCADIA R . With the entire numerical workload being highly parallelizable through domain decomposition techniques, associated CPU-time requirements that adhere to the efficiency needs in the nuclear industry can be expected to become feasible in the near future. The accuracy enhancement obtainable by using SP 3 instead of SP 1 has been verified by a detailed comparison of ARTEMIS 16-group pin-by-pin SP N results with KAERI's DeCart reference results for the 2D pin-by-pin Purdue UO 2 /MOX benchmark. This article presents the accuracy enhancement verification and quantifies the achieved ARTEMIS-SP 3 computational performance for a number of 2D and 3D multi-group and multi-box (up to pin-by-pin) core computations. (authors)
International Nuclear Information System (INIS)
Hill, T.R.; Reed, W.H.
1976-01-01
TIMEX solves the time-dependent, one-dimensional multigroup transport equation with delayed neutrons in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous problems subject to vacuum, reflective, periodic, white, albedo or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. The time variable is differenced by an explicit technique that is unconditionally stable so that arbitrarily large time steps can be taken. Because no iteration is performed the method is exceptionally fast in terms of computing time per time step. Two acceleration methods, exponential extrapolation and rebalance, are utilized to improve the accuracy of the time differencing scheme. Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXCOR can be accommodated. The running time for TIMEX is highly problem-dependent, but varies almost linearly with the total number of unknowns and time steps. Provision is made for creation of standard interface output files for angular fluxes and angle-integrated fluxes. Five interface units (use of interface units is optional), five output units, and two system input/output units are required. A large bulk memory is desirable, but may be replaced by disk, drum, or tape storage. 13 tables, 9 figures
International Nuclear Information System (INIS)
Mehlhorn, Thomas Alan; Kurecka, Christopher J.; McClarren, Ryan; Brunner, Thomas A.; Holloway, James Paul
2005-01-01
The original LDRD proposal was to use a nonlinear diffusion solver to compute estimates for the material temperature that could then be used in a Implicit Monte Carlo (IMC) calculation. At the end of the first year of the project, it was determined that this was not going to be effective, partially due to the concept, and partially due to the fact that the radiation diffusion package was not as efficient as it could be. The second, and final year, of the project focused on improving the robustness and computational efficiency of the radiation diffusion package in ALEGRA. To this end, several new multigroup diffusion methods have been developed and implemented in ALEGRA. While these methods have been implemented, their effectiveness of reducing overall simulation run time has not been fully tested. Additionally a comprehensive suite of verification problems has been developed for the diffusion package to ensure that it has been implemented correctly. This process took considerable time, but exposed significant bugs in both the previous and new diffusion packages, the linear solve packages, and even the NEVADA Framework's parser. In order to manage this large suite of problem, a new tool called Tampa has been developed. It is a general tool for automating the process of running and analyzing many simulations. Ryan McClarren, at the University of Michigan has been developing a Spherical Harmonics capability for unstructured meshes. While still in the early phases of development, this promises to bridge the gap in accuracy between a full transport solution using IMC and the diffusion approximation
Rennert, Knut; Nitschke, Mirko; Wallert, Maria; Keune, Natalie; Raasch, Martin; Lorkowski, Stefan; Mosig, Alexander S
2017-01-01
Harvesting cultivated macrophages for tissue engineering purposes by enzymatic digestion of cell adhesion molecules can potentially result in unintended activation, altered function, or behavior of these cells. Thermo-responsive polymer is a promising tool that allows for gentle macrophage detachment without artificial activation prior to subculture within engineered tissue constructs. We therefore characterized different species of thermo-responsive polymers for their suitability as cell substrate and to mediate gentle macrophage detachment by temperature shift. Primary human monocyte- and THP-1-derived macrophages were cultured on thermo-responsive polymers and characterized for phagocytosis and cytokine secretion in response to lipopolysaccharide stimulation. We found that both cell types differentially respond in dependence of culture and stimulation on thermo-responsive polymers. In contrast to THP-1 macrophages, primary monocyte-derived macrophages showed no signs of impaired viability, artificial activation, or altered functionality due to culture on thermo-responsive polymers compared to conventional cell culture. Our study demonstrates that along with commercially available UpCell carriers, two other thermo-responsive polymers based on poly(vinyl methyl ether) blends are attractive candidates for differentiation and gentle detachment of primary monocyte-derived macrophages. In summary, we observed similar functionality and viability of primary monocyte-derived macrophages cultured on thermo-responsive polymers compared to standard cell culture surfaces. While this first generation of custom-made thermo-responsive polymers does not yet outperform standard culture approaches, our results are very promising and provide the basis for exploiting the unique advantages offered by custom-made thermo-responsive polymers to further improve macrophage culture and recovery in the future, including the covalent binding of signaling molecules and the reduction of
Thermo-aeraulics of high level waste storage facilities
International Nuclear Information System (INIS)
Lagrave, Herve; Gaillard, Jean-Philippe; Laurent, Franck; Ranc, Guillaume; Duret, Bernard
2006-01-01
This paper discusses the research undertaken in response to axis 3 of the 1991 radioactive waste management act, and possible solutions concerning the processes under consideration for conditioning and long-term interim storage of long-lived radioactive waste. The notion of 'long-term' is evaluated with respect to the usual operating lifetime of a basic nuclear installation, about 50 years. In this context, 'long-term' is defined on a secular time scale: the lifetime of the facility could be as long as 300 years. The waste package taken into account is characterized notably by its high thermal power release. Studies were carried out in dedicated facilities for vitrified waste and for spent UOX and MOX fuel. The latter are not considered as wastes, owing to the value of the reusable material they contain. Three primary objectives have guided the design of these long-term interim storage facilities: - ensure radionuclide containment at all times; - permit retrieval of the containers at any time; - minimize surveillance; - maintenance costs. The CEA has also investigated surface and subsurface facilities. It was decided to work on generic sites with a reasonable set of parameters values that should be applicable at most sites in France. All the studies and demonstrations to date lead to the conclusion that long-term interim storage is technically feasible. The paper addresses the following items: - Long-term interim storage concepts for high-level waste; - Design principles and options for the interim storage facilities; - General architecture; - Research topics, Storage facility ventilation, Dimensioning of the facility; - Thermo-aeraulics of a surface interim storage facility; - VALIDA surface loop, VALIDA single container test campaign, Continuation of the VALIDA program; - Thermo-aeraulics of a network of subsurface interim storage galleries; - SIGAL subsurface loop; - PROMETHEE subsurface loop; - Temperature behaviour of the concrete structures; - GALATEE
Thermo-aeraulics of high level waste storage facilities
Energy Technology Data Exchange (ETDEWEB)
Lagrave, Herve; Gaillard, Jean-Philippe; Laurent, Franck; Ranc, Guillaume [CEA/Valrho, B.P. 17171, F-30207 Bagnols-sur-Ceze (France); Duret, Bernard [CEA Grenoble, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)
2006-07-01
This paper discusses the research undertaken in response to axis 3 of the 1991 radioactive waste management act, and possible solutions concerning the processes under consideration for conditioning and long-term interim storage of long-lived radioactive waste. The notion of 'long-term' is evaluated with respect to the usual operating lifetime of a basic nuclear installation, about 50 years. In this context, 'long-term' is defined on a secular time scale: the lifetime of the facility could be as long as 300 years. The waste package taken into account is characterized notably by its high thermal power release. Studies were carried out in dedicated facilities for vitrified waste and for spent UOX and MOX fuel. The latter are not considered as wastes, owing to the value of the reusable material they contain. Three primary objectives have guided the design of these long-term interim storage facilities: - ensure radionuclide containment at all times; - permit retrieval of the containers at any time; - minimize surveillance; - maintenance costs. The CEA has also investigated surface and subsurface facilities. It was decided to work on generic sites with a reasonable set of parameters values that should be applicable at most sites in France. All the studies and demonstrations to date lead to the conclusion that long-term interim storage is technically feasible. The paper addresses the following items: - Long-term interim storage concepts for high-level waste; - Design principles and options for the interim storage facilities; - General architecture; - Research topics, Storage facility ventilation, Dimensioning of the facility; - Thermo-aeraulics of a surface interim storage facility; - VALIDA surface loop, VALIDA single container test campaign, Continuation of the VALIDA program; - Thermo-aeraulics of a network of subsurface interim storage galleries; - SIGAL subsurface loop; - PROMETHEE subsurface loop; - Temperature behaviour of the concrete
Energy Technology Data Exchange (ETDEWEB)
Silva, Davi Jose M.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: halves@iprj.uerj.br, E-mail: rcbarros@pq.cnpq.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Programa de Pos-Graduacao em Modelagem Computacional
2015-07-01
A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (S{sub N}) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The S{sub N} discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the S{sub N} transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup S{sub N} eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)
International Nuclear Information System (INIS)
Silva, Davi Jose M.; Alves Filho, Hermes; Barros, Ricardo C.
2015-01-01
A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (S N ) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The S N discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the S N transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup S N eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)
Verdam, Mathilde G E; Oort, Frans J; van der Linden, Yvette M; Sprangers, Mirjam A G
2015-03-01
Missing data due to attrition present a challenge for the assessment and interpretation of change and response shift in HRQL outcomes. The objective was to handle such missingness and to assess response shift and 'true change' with the use of an attrition-based multigroup structural equation modeling (SEM) approach. Functional limitations and health impairments were measured in 1,157 cancer patients, who were treated with palliative radiotherapy for painful bone metastases, before [time (T) 0], every week after treatment (T1 through T12), and then monthly for up to 2 years (T13 through T24). To handle missing data due to attrition, the SEM procedure was extended to a multigroup approach, in which we distinguished three groups: short survival (3-5 measurements), medium survival (6-12 measurements), and long survival (>12 measurements). Attrition after third, sixth, and 13th measurement occasions was 11, 24, and 41 %, respectively. Results show that patterns of change in functional limitations and health impairments differ between patients with short, medium, or long survival. Moreover, three response-shift effects were detected: recalibration of 'pain' and 'sickness' and reprioritization of 'physical functioning.' If response-shift effects would not have been taken into account, functional limitations and health impairments would generally be underestimated across measurements. The multigroup SEM approach enables the analysis of data from patients with different patterns of missing data due to attrition. This approach does not only allow for detection of response shift and assessment of true change across measurements, but also allow for detection of differences in response shift and true change across groups of patients with different attrition rates.
International Nuclear Information System (INIS)
Kodeli, I.; Aldama, D. L.; De Leege, P. F. A.; Legrady, D.; Hoogenboom, J. E.; Cowan, P.
2004-01-01
As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) project of the EU community's 5. framework program a special purpose multigroup cross-section library was prepared for use in deterministic and Monte Carlo oil well logging particle transport calculations. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (authors)
International Nuclear Information System (INIS)
Trkov, A.; Budnar, M.; Copic, M.; Perdan, A.; Ravnik, M.
1982-01-01
Multigroup constants for 1-H-1, 92-U-235, and 92-U-238 have been calculated. Averaged cross-sections and other constants have been prepared in the WIMS 69-group format. Comparison has been made between group constants obtained with several evaluated libraries (KEDAK-3 1975, 1979, ENDF/B-4, ENDF/B-5) and the WIMS-D library. Observed differences are most pronounced in the resonance and fast region. From test runs on fuel cell with the WIMS program it can be deduced that these differences affect the fewgroup constants significantly. (author)
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.
International Nuclear Information System (INIS)
Woznicki, Z.I.
1983-07-01
This report presents the HEXAGA-III-programme solving multi-group time-independent real and/or adjoint neutron diffusion equations for three-dimensional-triangular-z-geometry. The method of solution is based on the AGA two-sweep iterative method belonging to the family of factorization techniques. An arbitrary neutron scattering model is permitted. The report written for users provides the description of the programme input and output and the use of HEXAGA-III is illustrated by a sample reactor problem. (orig.) [de
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1975-10-01
The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level
International Nuclear Information System (INIS)
Woznicki, Z.
1979-06-01
This report presents the AGA two-sweep iterative methods belonging to the family of factorization techniques in their practical application in the HEXAGA-II two-dimensional programme to obtain the numerical solution to the multi-group, time-independent, (real and/or adjoint) neutron diffusion equations for a fine uniform triangular mesh. An arbitrary group scattering model is permitted. The report written for the users provides the description of input and output. The use of HEXAGA-II is illustrated by two sample reactor problems. (orig.) [de
International Nuclear Information System (INIS)
Pashchenko, A.B.; Wienke, H.; Ganesan, S.
1996-01-01
Selected neutron reaction nuclear data evaluations and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into GENDF and MATXS format using the NJOY system by R.E. MacFarlane, in VITAMIN-J group structure with VITAMIN-E weighting spectrum. This document summarizes the resulting multigroup data library FENDL/MG version 1.1. The data are available costfree, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 7 refs, 1 tab
Fast reactor safety and computational thermo-fluid dynamics approaches
International Nuclear Information System (INIS)
Ninokata, Hisashi; Shimizu, Takeshi
1993-01-01
This article provides a brief description of the safety principle on which liquid metal cooled fast breeder reactors (LMFBRs) is based and the roles of computations in the safety practices. A number of thermohydraulics models have been developed to date that successfully describe several of the important types of fluids and materials motion encountered in the analysis of postulated accidents in LMFBRs. Most of these models use a mixture of implicit and explicit numerical solution techniques in solving a set of conservation equations formulated in Eulerian coordinates, with special techniques included to specific situations. Typical computational thermo-fluid dynamics approaches are discussed in particular areas of analyses of the physical phenomena relevant to the fuel subassembly thermohydraulics design and that involve describing the motion of molten materials in the core over a large scale. (orig.)
Measurement of water activity from shales through thermo hygrometer
Energy Technology Data Exchange (ETDEWEB)
Rabe, Claudio [Pontificia Univ. Catolica do Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Civil. Grupo de Tecnologia e Engenharia de Petroleo (GTEP)
2004-07-01
This paper presents a campaign of lab tests to obtain the water activity from shales and its pore fluid originated from offshore and onshore basin. The results of water activity from shales indicate that the values rang from 0.754 to 0.923 and for the pore fluid are between 0.987 and 0.940. The results show that the water activity of interstitial water can be obtained in 6 days and the rock in 10 days using the thermo hygrometer used. The degree of saturation, water content, kind and tenor of expansible and hydratable clay mineral, total and interconnected porosity, salinity of interstitial fluid and the capillary pressure of shale samples affected the results of water activity. (author)
Geometric Optimization of Thermo-electric Coolers Using Simulated Annealing
International Nuclear Information System (INIS)
Khanh, D V K; Vasant, P M; Elamvazuthi, I; Dieu, V N
2015-01-01
The field of thermo-electric coolers (TECs) has grown drastically in recent years. In an extreme environment as thermal energy and gas drilling operations, TEC is an effective cooling mechanism for instrument. However, limitations such as the relatively low energy conversion efficiency and ability to dissipate only a limited amount of heat flux may seriously damage the lifetime and performance of the instrument. Until now, many researches were conducted to expand the efficiency of TECs. The material parameters are the most significant, but they are restricted by currently available materials and module fabricating technologies. Therefore, the main objective of finding the optimal TECs design is to define a set of design parameters. In this paper, a new method of optimizing the dimension of TECs using simulated annealing (SA), to maximize the rate of refrigeration (ROR) was proposed. Equality constraint and inequality constraint were taken into consideration. This work reveals that SA shows better performance than Cheng's work. (paper)
Contribution of thermo-fluid analyses to the LHC experiments
Gasser, G
2003-01-01
The big amount of electrical and electronic equipment that will be installed in the four LHC experiments will cause important heat dissipation into the detectors’ volumes. This is a major issue for the experimental groups, as temperature stability is often a fundamental requirement for the different sub-detectors to be able to provide a good measurement quality. The thermofluid analyses that are carried out in the ST/CV group are a very efficient tool to understand and predict the thermal behaviour of the detectors. These studies are undertaken according to the needs of the experimental groups; they aim at evaluate the thermal stability for a proposed design, or to compare different technical solutions in order to choose the best one for the final design. The usual approach to carry out these studies is first presented and then, some practical examples of thermo-fluid analyses are presented focusing on the main results in order to illustrate their contribution.
A conjugate thermo-electric model for a composite medium.
Directory of Open Access Journals (Sweden)
Oscar Chávez
Full Text Available Electrical transmission signals have been used for decades to characterize the internal structure of composite materials. We theoretically analyze the transmission of an electrical signal through a composite material which consists of two phases with different chemical compositions. We assume that the temperature of the biphasic system increases as a result of Joule heating and its electrical resistivity varies linearly with temperature; this last consideration leads to simultaneously study the electrical and thermal effects. We propose a nonlinear conjugate thermo-electric model, which is solved numerically to obtain the current density and temperature profiles for each phase. We study the effect of frequency, resistivities and thermal conductivities on the current density and temperature. We validate the prediction of the model with comparisons with experimental data obtained from rock characterization tests.
Thermo-mechanical process for treatment of welds
International Nuclear Information System (INIS)
Malik, R.K.
1980-03-01
Benefits from thermo-mechanical processing (TMP) of austenitic stainless steel weldments, analogous to hot isostatic pressing (HIP) of castings, most likely result from compressive plastic deformation, enhanced diffusion, and/or increased dislocation density. TMP improves ultrasonic inspectability of austenitic stainless steel welds owing to: conversion of cast dendrites into equiaxed austenitic grains, reduction in size and number of stringers and inclusions, and reduction of delta ferrite content. TMP induces structural homogenization and healing of void-type defects and thus contributes to an increase in elongation, impact strength, and fracture toughness as well as a significant reduction in data scatter for these properties. An optimum temperature for TMP or HIP of welds is one which causes negligible grain growth and an acceptable reduction in yield strength, and permits healing of porosity
Investigation of Thermo-regulating Properties of Multilayer Textile Package
Directory of Open Access Journals (Sweden)
Julija Baltušnikaitė
2015-09-01
Full Text Available Thermal comfort of a clothing system is one of the important goals of the developer that require an engineering approach. In this research work a thermo-regulating textile packages were developed and a wearing comfort of protective clothing consisting from those packages was improved. The microcapsules were added on the fabric surface using pad-dry-cure method. The thermal properties and stabilities were measured using differential scanning calorimetry. The results suggest that higher values of thermal resistance were obtained after incorporation of fabric, coated by PCMs, into inert layer of multilayer textile package. DOI: http://dx.doi.org/10.5755/j01.ms.21.3.6920
Using homogenization, sonication and thermo-sonication to inactivate fungi
Bevilacqua, Antonio; Sinigaglia, Milena; Corbo, Maria Rosaria
2016-01-01
Ultrasound (US), Thermo-sonication (TS) and High Pressure Homogenization (HPH) were studied as tools to inactivate the spores of Penicillium spp. and Mucor spp. inoculated in distilled water. For US, the power ranged from 40% to 100%, pulse from 2 to 10 s, and duration of the treatment from 2 to 10 min. TS was performed combining US (40–80% of power, for 8 min and pulse of 2 s) with a thermal treatment (50, 55 and 60°C at 4, 8 and 12 min). Homogenization was done at 30–150 MPa for 1, 2 and 3 times. Power was the most important factors to determine the antifungal effect of US and TS towards the conidia of Penicillium spp.; on the other hand, in US treatments Mucor spp. was also affected by pulse and time. HPH exerted a significant antifungal effect only if the highest pressures were applied for 2–3 times. PMID:27375964
Thermo-climatic cost of the domestic consumption products
Energy Technology Data Exchange (ETDEWEB)
Szargut, Jan; Stanek, Wojciech [Institute of Thermal Technology, Silesian University of Technology, Konarskiego 22, 44-100 Gliwice (Poland)
2010-02-15
The thermo-climatic cost (TCC) expresses the cumulative emission of CO{sub 2} burdening all the steps of production processes connected with the fabrication of particular consumption products. The TCC of the considered product results from the consumption of semi-finished products and energy carriers. The TCC of hydrocarbon fuels contains three components: the immediate emission of CO{sub 2} resulting from the combustion of carbon, the TCC of delivery and processing and the TCC resulting from import of fuels. The TCC-component connected with import results from the TCC of the domestic products exported in order to gain the financial means for import. The values of the TCC can be used for the minimization of climatic damages by the selection of the production technology or the design and operation parameters of new processes. (author)
Simulation of Thermo-viscoplastic Behaviors for AISI 4140 Steel
Li, Hong-Bin; Feng, Yun-Li
2016-04-01
The thermo-viscoplastic behaviors of AISI 4140 steel are investigated over wide ranges of strain rate and deformation temperature by isothermal compression tests. Based on the experimental results, a unified viscoplastic constitutive model is proposed to describe the hot compressive deformation behaviors of the studied steel. In order to reasonably evaluate the work hardening behaviors, a strain hardening material constant (h0) is expressed as a function of deformation temperature and strain rate in the proposed constitutive model. Also, the sensitivity of initial value of internal variable s to the deformation temperature is discussed. Furthermore, it is found that the initial value of internal variable s can be expressed as a linear function of deformation temperature. Comparisons between the measured and predicted results confirm that the proposed constitutive model can give an accurate and precise estimate of the inelastic stress-strain relationships for the studied high-strength steel.
Experimental and theoretical studies of buoyant-thermo capillary flow
International Nuclear Information System (INIS)
Favre, E.; Blumenfeld, L.; Soubbaramayer
1996-01-01
In the AVLIS process, uranium metal is evaporated using a high power electron gun. We have prior discussed the power balance equation in the electron beam evaporation process and pointed out, among the loss terms, the importance of the power loss due to the convective flow in the molten pool driven by buoyancy and thermo capillarity. An empirical formula has been derived from model experiments with cerium, to estimate the latter power loss and that formula can be used practically in engineering calculations. In order to complete the empirical approach, a more fundamental research program of theoretical and experimental studies have been carried out in Cea-France, with the objective of understanding the basic phenomena (heat transport, flow instabilities, turbulence, etc.) occurring in a convective flow in a liquid layer locally heated on its free surface
Study of a Piezo-Thermo-Elastic Materials Console
Directory of Open Access Journals (Sweden)
hamza madjid berrabah
2015-09-01
Full Text Available In the first part of this work, analytical expressions were determined for the stresses through the thickness of a composite beam submitted to electrical excitation. In the second part of this study we are interested in the theory of elasticity, which is used to obtain exact solutions of piezo-thermo-elastic consoles gradually coupled evaluated under different loads. These solutions are used to identify the piezoelectric parameter and thermal coefficients of the materials. In addition, numerical results are obtained for the analysis of the loaded console by two different types of loading. In this study we show also that changing the linear thermal parameters of the material does not affect the distribution of the stress and the induction of the beam. However it affetcs the components of the deformation, electric field, the displacement and the electric potential of the console.
Thermo-electro-hydrodynamic convection under microgravity: a review
Energy Technology Data Exchange (ETDEWEB)
Mutabazi, Innocent; Yoshikawa, Harunori N; Fogaing, Mireille Tadie; Travnikov, Vadim; Crumeyrolle, Olivier [Laboratoire Ondes et Milieux Complexes, UMR 6294, CNRS-Université du Havre, CS 80450, F-76058 Le Havre Cedex (France); Futterer, Birgit; Egbers, Christoph, E-mail: Innocent.Mutabazi@univ-lehavre.fr [Department of Aerodynamics and Fluid Mechanics, Brandenburg University of Technology Cottbus-Senftenberg, Cottbus (Germany)
2016-12-15
Recent studies on thermo-electro-hydrodynamic (TEHD) convection are reviewed with focus on investigations motivated by the analogy with natural convection. TEHD convection originates in the action of the dielectrophoretic force generated by an alternating electric voltage applied to a dielectric fluid with a temperature gradient. This electrohydrodynamic force is analogous to Archimedean thermal buoyancy and can be regarded as a thermal buoyancy force in electric effective gravity. The review is concerned with TEHD convection in plane, cylindrical, and spherical capacitors under microgravity conditions, where the electric gravity can induce convection without any complexities arising from geometry or the buoyancy force due to the Earth’s gravity. We will highlight the convection in spherical geometry, comparing developed theories and numerical simulations with the GEOFLOW experiments performed on board the International Space Station (ISS). (paper)
Thermo-responsive hydrogels for intravitreal injection and biomolecule release
Drapala, Pawel
In this dissertation, we develop an injectable polymer system to enable localized and prolonged release of therapeutic biomolecules for improved treatment of Age-Related Macular Degeneration (AMD). Thermo-responsive hydrogels derived from N-isopropylacrylamide (NIPAAm) and cross-linked with poly(ethylene glycol) (PEG) poly(L-Lactic acid) (PLLA) copolymer were synthesized via free-radical polymerization. These materials were investigated for (a) phase change behavior, (b) in-vitro degradation, (c) capacity for controlled drug delivery, and (d) biocompatibility. The volume-phase transition temperature (VPTT) of the PNIPAAm- co-PEG-b-PLLA hydrogels was adjusted using hydrophilic and hydrophobic moieties so that it is ca. 33°C. These hydrogels did not initially show evidence of degradation at 37°C due to physical cross-links of collapsed PNIPAAm. Only after addition of glutathione chain transfer agents (CTA)s to the precursor did the collapsed hydrogels become fully soluble at 37°C. CTAs significantly affected the release kinetics of biomolecules; addition of 1.0 mg/mL glutathione to 3 mM cross-linker accelerated hydrogel degradation, resulting in 100% release in less than 2 days. This work also explored the effect of PEGylation in order to tether biomolecules to the polymer matrix. It was demonstrated that non-site-specific PEGylation can postpone the burst release of solutes (up to 10 days in hydrogels with 0.5 mg/mL glutathione). Cell viability assays showed that at least two 20-minute buffer extraction steps were needed to remove cytotoxic elements from the hydrogels. Clinically-used therapeutic biomolecules LucentisRTM and AvastinRTM were demonstrated to be both stable and bioactive after release form PNIPAAm-co-PEG-b-PLLA hydrogels. The thermo-responsive hydrogels presented here offer a promising platform for the localized delivery of proteins such as recombinant antibodies.
Thermo-hydrogenating treatments in Ti-6Al-4V
International Nuclear Information System (INIS)
Guitar, A; Domizzi, G; Luppo, M.I; Vigna, G
2006-01-01
The production of components of Ti alloys, specifically Ti-6Al-4V, involves some difficulties in obtaining the final desired microstructure, producing decrease in the material's mechanical properties. In the specific case of materials to be used for surgical implants an equiaxial fine grain microstructure of α phase a with an homogenously precipitated β phase is needed. The modification of certain microstructural features is not possible based on simple thermal treatments. Thermomechanical treatments are effective for transforming the lamellar α phase into equiaxial α, but these methods include major deformations in the (α + β) two-phase field. In order to avoid this stage, thermo-hydrogenating processes were used (THP). The THP involve a treatment of β solubilization before, during or after the hydrogenation, a possible isothermal treatment below the β hydrogenated transus temperature and the final vacuum dehydrogenation. The development of treatments using hydrogen as a temporary alloying element creates a new class of microstructures, which are finer than equiaxial structures and respond well to resistance to traction and fatigue. Since the THP do not include the working of the material to control the microstructure, they are more appropriate for use with shaped components close to the end, like those obtained by powder metallurgy or smelting. Different thermo-hydrogenating treatments in Ti-6Al-4V to modify the microstructure were studied. Final microstructures of α fine phase and β disperse phase were obtained using THP in samples with initial lamellar α phase separated by thin sheets of β phase. The characterization of the initial material and of the transformed material was carried out using optic and scanning electron microscopy (CW)
Inkjet-Printed Biofunctional Thermo-Plasmonic Interfaces for Patterned Neuromodulation.
Kang, Hongki; Lee, Gu-Haeng; Jung, Hyunjun; Lee, Jee Woong; Nam, Yoonkey
2018-02-27
Localized heat generation by the thermo-plasmonic effect of metal nanoparticles has great potential in biomedical engineering research. Precise patterning of the nanoparticles using inkjet printing can enable the application of the thermo-plasmonic effect in a well-controlled way (shape and intensity). However, a universally applicable inkjet printing process that allows good control in patterning and assembly of nanoparticles with good biocompatibility is missing. Here we developed inkjet-printing-based biofunctional thermo-plasmonic interfaces that can modulate biological activities. We found that inkjet printing of plasmonic nanoparticles on a polyelectrolyte layer-by-layer substrate coating enables high-quality, biocompatible thermo-plasmonic interfaces across various substrates (rigid/flexible, hydrophobic/hydrophilic) by induced contact line pinning and electrostatically assisted nanoparticle assembly. We experimentally confirmed that the generated heat from the inkjet-printed thermo-plasmonic patterns can be applied in micrometer resolution over a large area. Lastly, we demonstrated that the patterned thermo-plasmonic effect from the inkjet-printed gold nanorods can selectively modulate neuronal network activities. This inkjet printing process therefore can be a universal method for biofunctional thermo-plasmonic interfaces in various bioengineering applications.
Janković, Bojan; Marinović-Cincović, Milena; Janković, Marija
2017-09-01
Kinetics of degradation for Aronia melanocarpa fresh fruits in argon and air atmospheres were investigated. The investigation was based on probability distributions of apparent activation energy of counterparts (ε a ). Isoconversional analysis results indicated that the degradation process in an inert atmosphere was governed by decomposition reactions of esterified compounds. Also, based on same kinetics approach, it was assumed that in an air atmosphere, the primary compound in degradation pathways could be anthocyanins, which undergo rapid chemical reactions. A new model of reactivity demonstrated that, under inert atmospheres, expectation values for ε a occured at levels of statistical probability. These values corresponded to decomposition processes in which polyphenolic compounds might be involved. ε a values obeyed laws of binomial distribution. It was established that, for thermo-oxidative degradation, Poisson distribution represented a very successful approximation for ε a values where there was additional mechanistic complexity and the binomial distribution was no longer valid. Copyright © 2017 Elsevier Ltd. All rights reserved.
G.M.S.I. - A generalised multigroup system of calculations using the IBM 7030 (stretch) computer
International Nuclear Information System (INIS)
Gratton, C.P.; Smith, P.E.
1965-02-01
G.M.S. is a generalised system of reactor physics calculations written in the FORTRAN programming language for the IBM 7030 (STRETCH) computer. The programme will perform cell, supercell and overall reactor physics problems within the limits of one dimension using either slab or cylindrical geometry. The Winfrith DSN(1) programme has been incorporated as a subroutine and all flux distribution calculations are performed using Transport Theory. A facility for the study of fuel burn-up has been included with boundary conditions for either cell irradiations or the burn-up characteristics of the overall reactor. Particular attention has been paid to the simplicity of the input to the programme to allow a wide range of potential users and extensive edit facilities have been incorporated to allow the better understanding of the multigroup output. A graphical output option is available using the STROMBERG-CARLSON equipment and, by a series of commands, graphs of power peaking and burn-up rates are produced for the designer or, alternatively, neutron spectra and activations for comparison with experiment, The logic of the programme is written in terms of 40 neutron energy groups, the group boundaries of which may be selected by the user. Clearly the use of the system for fast reactor calculations requires that greater emphasis be given to the representation of neutron events at high energy than would be necessary for thermal reactor studies. It is envisaged therefore that eventually two library files will be made available to cover the extreme cases of fast and thermal systems and that the group boundaries of these will be selected to enable adequate treatment of intermediate spectrum reactors. At the present time the thermal reactor library file only has been prepared; the content of this library is discussed in Section 2. The group structure is comprised of twelve groups in the fast and intermediate energy region, ten groups in and around the plutonium 240 resonance at 1
International Nuclear Information System (INIS)
Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.
1990-09-01
Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations
Energy Technology Data Exchange (ETDEWEB)
Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-05-15
Parallelization of Monte Carlo simulation is widely adpoted. There are also several parallel algorithms developed for the SN transport theory using the parallel wave sweeping algorithm and for the CPM using parallel ray tracing. For practical purpose of reactor physics application, the thermal feedback and burnup effects on the multigroup cross section should be considered. In this respect, the domain decomposition method(DDM) is suitable for distributing the expensive cross section calculation work. Parallel transport code and diffusion code based on the Raviart-Thomas mixed finite element method was developed. However most of the developed methods rely on the heuristic convergence of flux and current at the domain interfaces. Convergence was not attained in some cases. Mechanical stress computation community has also work on the DDM to solve the stress-strain equation using the finite element methods. The most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multigroup diffusion problem in this study.
Kravchenko, O. V.; Mitroshin, I. V.; Gabdulkhakov, A. G.; Nikonov, S. V.; Garber, M. B.
2011-07-01
Lateral L12-stalk (P1-stalk in Archaea, P1/P2-stalk in eukaryotes) is an obligatory morphological element of large ribosomal subunits in all organisms studied. This stalk is composed of the complex of ribosomal proteins L10(P0) and L12(P1) and interacts with 23S rRNA through the protein L10(P0). L12(P1)-stalk is involved in the formation of GTPase center of the ribosome and plays an important role in the ribosome interaction with translation factors. High mobility of this stalk puts obstacles in determination of its structure within the intact ribosome. Crystals of a two-domain N-terminal fragment of ribosomal protein L10(P0) from the archaeon Methanococcus jannaschii in complex with a specific fragment of rRNA from the same organism have been obtained. The crystals diffract X-rays at 3.2 Å resolution.
Energy Technology Data Exchange (ETDEWEB)
Kravchenko, O. V.; Mitroshin, I. V.; Gabdulkhakov, A. G.; Nikonov, S. V.; Garber, M. B., E-mail: garber@vega.protres.ru [Institute of Protein Research RAS (Russian Federation)
2011-07-15
Lateral L12-stalk (P1-stalk in Archaea, P1/P2-stalk in eukaryotes) is an obligatory morphological element of large ribosomal subunits in all organisms studied. This stalk is composed of the complex of ribosomal proteins L10(P0) and L12(P1) and interacts with 23S rRNA through the protein L10(P0). L12(P1)-stalk is involved in the formation of GTPase center of the ribosome and plays an important role in the ribosome interaction with translation factors. High mobility of this stalk puts obstacles in determination of its structure within the intact ribosome. Crystals of a two-domain N-terminal fragment of ribosomal protein L10(P0) from the archaeon Methanococcus jannaschii in complex with a specific fragment of rRNA from the same organism have been obtained. The crystals diffract X-rays at 3.2 Angstrom-Sign resolution.
Thermo-electric oxidization of iron in lithium niobate crystals
International Nuclear Information System (INIS)
Falk, Matthias
2007-01-01
Lithium niobate crystals (LiNbO 3 ) are a promising material for nonlinear-optical applications like frequency conversion to generate visible light, e.g., in laser displays, but their achievable output power is greatly limited by the ''optical damage'', i.e., light-induced refractive-index changes caused by excitation of electrons from iron impurities and the subsequent retrapping in unilluminated areas of the crystal. The resulting space-charge fields modify the refractive indices due to the electro-optic effect. By this ''photorefractive effect'' the phase-matching condition, i.e., the avoidance of destructive interference between light generated at different crystal positions due to the dispersion of the fundamental wave and the converted wave, is disturbed critically above a certain light intensity threshold. The influence of annealing treatments conducted in the presence of an externally applied electric field (''thermo-electric oxidization'') on the valence state of iron impurities and thereby on the optical damage is investigated. It is observed that for highly iron-doped LiNbO 3 crystals this treatment leads to a nearly complete oxidization from Fe 2+ to Fe 3+ indicated by the disappearance of the absorption caused by Fe 2+ . During the treatment an absorption front forms that moves through the crystal. The absorption in the visible as well as the electrical conductivity are decreased by up to five orders of magnitude due to this novel treatment. The ratio of the Fe 2+ concentration to the total iron concentration - a measure for the strength of the oxidization - is in the order of 10 -6 for oxidized crystals whereas it is about 10 -1 for untreated samples. Birefringence changes are observed at the absorption front that are explained by the removal of hydrogen and lithium ions from the crystal that compensate for the charges of the also removed electrons from Fe 2+ . A microscopic shock-wave model is developed that explains the observed absorption front by
Data for effects of lanthanum complex on the thermo-oxidative aging of natural rubber
Directory of Open Access Journals (Sweden)
Wei Zheng
2015-12-01
Full Text Available Novel mixed antioxidants composed of antioxidant IPPD and lanthanum (La complex were added as a filler to form natural rubber (NR composites. By mechanical testing, Fourier transform infrared spectroscopy with attenuated total reflectance (FTIR-ATR and thermogravimetric analysis (TGA, a string of data, including the mechanical properties, the variation of internal groups and the thermal and thermo-oxidative decompositions of NR, was presented in this data article. The data accompanying its research article [1] studied the thermo-oxidative aging properties of NR in detail. The density function theoretical (DFT calculations were also used as an assistant to study the thermo-oxidative aging mechanism of NR. The data revealed that this new rare-earth antioxidant could indeed enhance the thermo-oxidative aging resistance of NR, which is associated with its different function mechanism from that of the pure antioxidant IPPD.
Data for effects of lanthanum complex on the thermo-oxidative aging of natural rubber.
Zheng, Wei; Liu, Li; Zhao, Xiuying; He, Jingwei; Wang, Ao; Chan, Tung W; Wu, Sizhu
2015-12-01
Novel mixed antioxidants composed of antioxidant IPPD and lanthanum (La) complex were added as a filler to form natural rubber (NR) composites. By mechanical testing, Fourier transform infrared spectroscopy with attenuated total reflectance (FTIR-ATR) and thermogravimetric analysis (TGA), a string of data, including the mechanical properties, the variation of internal groups and the thermal and thermo-oxidative decompositions of NR, was presented in this data article. The data accompanying its research article [1] studied the thermo-oxidative aging properties of NR in detail. The density function theoretical (DFT) calculations were also used as an assistant to study the thermo-oxidative aging mechanism of NR. The data revealed that this new rare-earth antioxidant could indeed enhance the thermo-oxidative aging resistance of NR, which is associated with its different function mechanism from that of the pure antioxidant IPPD.
Thermo-Plasmonics for Localized Graphitization and Welding of Polymeric Nanofibers
Directory of Open Access Journals (Sweden)
Ahnaf Usman Zillohu
2014-01-01
Full Text Available There is a growing interest in modulating the temperature under the illumination of light. As a heat source, metal nanoparticles (NPs have played an important role to pave the way for a new branch of plasmonics, i.e., thermo-plasmonics. While thermo-plasmonics have been well established in photo-thermal therapy, it has received comparatively less attention in materials science and chemistry. Here, we demonstrate the first proof of concept experiment of local chemistry and graphitization of metalized polymeric nanofibers through thermo-plasmonic effect. In particular, by tuning the plasmonic absorption of the nanohybrid through a change in the thickness of the deposited silver film on the fibers, the thermo-plasmonic effect can be adjusted in such a way that high enough temperature is generated enabling local welding and graphitization of the polymeric nanofibers.
CAS-ATLID (co-alignment sensor of ATLID instrument) thermo-structural design and performance
Moreno, Javier; Serrano, Javier; González, David; Rodríguez, Gemma; Manjón, Andrés.; Vásquez, Eloi; Carretero, Carlos; Martínez, Berta
2017-11-01
This paper describes the main thermo-mechanical design features and performances of the Co-Alignment Sensor (CAS) developed by LIDAX and CRISA under ESA program with AIRBUS Defence and Space as industry prime.
A thermo-degradable hydrogel with light-tunable degradation and drug release.
Hu, Jingjing; Chen, Yihua; Li, Yunqi; Zhou, Zhengjie; Cheng, Yiyun
2017-01-01
The development of thermo-degradable hydrogels is of great importance in drug delivery. However, it still remains a huge challenge to prepare thermo-degradable hydrogels with inherent degradation, reproducible, repeated and tunable dosing. Here, we reported a thermo-degradable hydrogel that is rapidly degraded above 44 °C by a facile chemistry. Besides thermo-degradability, the hydrogel also undergoes rapid photolysis with ultraviolet light. By embedding photothermal nanoparticles or upconversion nanoparticles into the gel, it can release the entrapped cargoes such as dyes, enzymes and anticancer drugs in an on-demand and dose-tunable fashion upon near-infrared light exposure. The smart hydrogel works well both in vitro and in vivo without involving sophisticated syntheses, and is well suited for clinical cancer therapy due to the high transparency and non-invasiveness features of near-infrared light. Copyright © 2016 Elsevier Ltd. All rights reserved.
Data for effects of lanthanum complex on the thermo-oxidative aging of natural rubber
Zheng, Wei; Liu, Li; Zhao, Xiuying; He, Jingwei; Wang, Ao; Chan, Tung W.; Wu, Sizhu
2015-01-01
Novel mixed antioxidants composed of antioxidant IPPD and lanthanum (La) complex were added as a filler to form natural rubber (NR) composites. By mechanical testing, Fourier transform infrared spectroscopy with attenuated total reflectance (FTIR-ATR) and thermogravimetric analysis (TGA), a string of data, including the mechanical properties, the variation of internal groups and the thermal and thermo-oxidative decompositions of NR, was presented in this data article. The data accompanying its research article [1] studied the thermo-oxidative aging properties of NR in detail. The density function theoretical (DFT) calculations were also used as an assistant to study the thermo-oxidative aging mechanism of NR. The data revealed that this new rare-earth antioxidant could indeed enhance the thermo-oxidative aging resistance of NR, which is associated with its different function mechanism from that of the pure antioxidant IPPD. PMID:26693513
Maruyama, Masashi; Shibuya, Keisuke
2017-08-22
Thermo-responsive adsorbents for immunoglobulin G (IgG) employing ε-polylysine (EPL) as a polymer backbone were developed. The introduction of mercaptoethylpyridine (MEP) as an IgG-binding ligand and hydrophobization of side chains afforded thermo-responsive IgG adsorbents, whose thermo-responsive IgG desorption ratio was up to 88% (EPL/MEP derivative 3m). The changes in surface densities of active MEP groups, which are caused by thermal conformational changes of the adsorbents, play key roles for IgG desorption. Although a trade-off of IgG adsorption capacity and IgG desorption ratio was observed, the present study offers a novel molecular design for thermo-responsive adsorbents with high synthetic accessibility and potentially low toxicity.
General thermo-elastic solution of radially heterogeneous, spherically isotropic rotating sphere
Energy Technology Data Exchange (ETDEWEB)
Bayat, Yahya; EkhteraeiToussi, THamid [Ferdowsi University of Mashhad, Mashhad (Iran, Islamic Republic of)
2015-06-15
A thick walled rotating spherical object made of transversely isotropic functionally graded materials (FGMs) with general types of thermo-mechanical boundary conditions is studied. The thermo-mechanical governing equations consisting of decoupled thermal and mechanical equations are represented. The centrifugal body forces of the rotation are considered in the modeling phase. The unsymmetrical thermo-mechanical boundary conditions and rotational body forces are expressed in terms of the Legendre series. The series method is also implemented in the solution of the resulting equations. The solutions are checked with the known literature and FEM based solutions of ABAQUS software. The effects of anisotropy and heterogeneity are studied through the case studies and the results are represented in different figures. The newly developed series form solution is applicable to the rotating FGM spherical transversely isotropic vessels having nonsymmetrical thermo-mechanical boundary condition.