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Sample records for cell calculation code

  1. BCG: a computer code for calculating neutron spectra and criticality in cells of fast reactors

    International Nuclear Information System (INIS)

    The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is discussed. The code solves the unidimensional neutron transport equation together with interface current relations at each energy point in an unionized energy grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstruced total microscopic cross sections of the atomic species in the cell. Results for a simplified fuel cell illustrate the high resolution and accuracy that can be obtained with the code. (author)

  2. Applying full multigroup cell characteristics from MCU code to finite difference calculations of neutron field in VVER core

    Energy Technology Data Exchange (ETDEWEB)

    Gorodkov, S.S.; Kalugin, M.A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Up to now core calculations with Monte Carlo provided only average cross-sections of mesh cells for further use either in finite difference calculations or as benchmark ones for approximate spectral algorithms. Now MCU code is capable to handle functions, which may be interpreted as average diffusion coefficients. Subsequently the results of finite difference calculations with cells characteristic sets obtained in such a way can be compared with Monte Carlo results as benchmarks, giving reliable information on quality of production code under consideration. As an example of such analysis, the results of mesh calculations with 1-, 2-, 4-, 8- and 12 neutron groups of some model VVER fuel assembly are presented in comparison with the exact Monte Carlo solution. As a second example, an analysis is presented of water gap approximate enlargement between fuel assemblies, allowing VVER core region be covered by regular mesh.

  3. Calculation of anisotropic few-group constants in asymptotic cells: the code ANICELL

    International Nuclear Information System (INIS)

    The theoretical background of the ANICELL computer program together with a user's manual is presented. ANICELL is a nuclear reactor neutron transport code which solves the traditional asymptotic and the so-called tilted flux transport problems in one-dimensional cylindrical geometry using linearly anisotropic scattering. The method of solution used is the first flight collision probability technique. Few-group constants including radial and axial diffusion coefficients for the cell are also prepared by the program. (author)

  4. PINSPEC. A Monte Carlo code for pin cell spectral calculations for educational applications

    International Nuclear Information System (INIS)

    Students in many reactor physics courses are exposed to canonical reactor physics concepts through theoretical problems simplified to allow for tractable analytical solutions. Such problems typically require tedious mathematical derivation which is often not the most effective approach to teaching basic reactor physics concepts. A new complementary methodology to introduce these concepts is made possible with PINSPEC, a pin cell Monte Carlo code for educational use. PINSPEC enables students to simulate pin cell models for various reactor types with a simple-to-use Python interface. PINSPEC uses point-wise cross section data and includes a module for Single-Level Breit-Wigner cross-section generation and Doppler broadening. The PINSPEC code supports a variety of tallies which students may use to compute resonance integrals, multi-group cross sections, and more for various materials and pin configurations. PINSPEC is undergoing review for open source release in the near future such that it will be a free and accessible tool for instructors developing reactor physics curricula with an applied and interactive approach to learning. (author)

  5. Calculations for a BWR Lattice with Adjacent Gadolinium Pins Using the Monte Carlo Cell Code Serpent v.1.1.7

    Directory of Open Access Journals (Sweden)

    Diego Ferraro

    2011-01-01

    Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.

  6. TEA: A Code Calculating Thermochemical Equilibrium Abundances

    Science.gov (United States)

    Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver

    2016-07-01

    We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. and Eriksson. It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature-pressure pairs. We tested the code against the method of Burrows & Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows & Sharp. Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is available under a reproducible-research, open-source license via https://github.com/dzesmin/TEA.

  7. TEA: A Code Calculating Thermochemical Equilibrium Abundances

    Science.gov (United States)

    Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver

    2016-07-01

    We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. and Eriksson. It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature–pressure pairs. We tested the code against the method of Burrows & Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows & Sharp. Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is available under a reproducible-research, open-source license via https://github.com/dzesmin/TEA.

  8. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)

  9. KENO-IV code benchmark calculation, (6)

    International Nuclear Information System (INIS)

    A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multigroup constants library MGCL. The present report describes the results of a benchmark test using criticality experiments about Plutonium fuel in various shape. In all, 33 cases of experiments have been calculated for Pu(NO3)4 aqueous solution, Pu metal or PuO2-polystyrene compact in various shape (sphere, cylinder, rectangular parallelepiped). The effective multiplication factors calculated for the 33 cases distribute widely between 0.955 and 1.045 due to wide range of system variables. (author)

  10. Data calculation program for RELAP 5 code

    Energy Technology Data Exchange (ETDEWEB)

    Silvestre, Larissa J.B.; Sabundjian, Gaiane, E-mail: larissajbs@usp.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    As the criteria and requirements for a nuclear power plant are extremely rigid, computer programs for simulation and safety analysis are required for certifying and licensing a plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors. A major difficulty in the simulation using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data leads to a very large number of mathematical operations for calculating the geometry of the components. Therefore, a mathematical friendly preprocessor was developed in order to perform these calculations and prepare RELAP5 input data. The Visual Basic for Application (VBA) combined with Microsoft EXCEL demonstrated to be an efficient tool to perform a number of tasks in the development of the program. Due to the absence of necessary information about some RELAP5 components, this work aims to make improvements to the Mathematic Preprocessor for RELAP5 code (PREREL5). For the new version of the preprocessor, new screens of some components that were not programmed in the original version were designed; moreover, screens of pre-existing components were redesigned to improve the program. In addition, an English version was provided for the new version of the PREREL5. The new design of PREREL5 contributes for saving time and minimizing mistakes made by users of the RELAP5 code. The final version of this preprocessor will be applied to Angra 2. (author)

  11. KENO-IV code benchmark calculation, (4)

    International Nuclear Information System (INIS)

    A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multi-group constants library MGCL. The present paper describes the results of a test using criticality experiments about slab-cylinder system of uranium nitrate solution. In all, 128 cases of experiments have been calculated for the slab-cylinder configuration with and without plexiglass reflector, having the various critical parameters such as the number of cylinders and height of the uranium nitrate solution. It is shown among several important results that the code and library gives a fairly good multiplication factor, that is, k sub(eff) -- 1.0 for heavily reflected cases, whereas k sub(eff) -- 0.91 for the unreflected ones. This suggests the necessity of more advanced treatment of the criticality calculation for the system where neutrons can easily leak out during slowing down process. (author)

  12. Electrical Conductivity Calculations from the Purgatorio Code

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, S B; Isaacs, W A; Sterne, P A; Wilson, B G; Sonnad, V; Young, D A

    2006-01-09

    The Purgatorio code [Wilson et al., JQSRT 99, 658-679 (2006)] is a new implementation of the Inferno model describing a spherically symmetric average atom embedded in a uniform plasma. Bound and continuum electrons are treated using a fully relativistic quantum mechanical description, giving the electron-thermal contribution to the equation of state (EOS). The free-electron density of states can also be used to calculate scattering cross sections for electron transport. Using the extended Ziman formulation, electrical conductivities are then obtained by convolving these transport cross sections with externally-imposed ion-ion structure factors.

  13. Methods and computer codes for nuclear systems calculations

    Indian Academy of Sciences (India)

    B P Kochurov; A P Knyazev; A Yu Kwaretzkheli

    2007-02-01

    Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space–time calculations. Computer code TRIFON solves space-energy problem in (, ) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space–time neutron processes simulation. Modification of TRIFON was developed for the simulation of space–time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.

  14. Calculation code MIXSET for Purex process

    International Nuclear Information System (INIS)

    MIXSET is a FORTRAN IV calculation code for Purex process that simulate the dynamic behavior of solvent extraction processes in mixer-settlers. Two options permit terminating dynamic phase by time or by achieving steady state. These options also permit continuing calculation successively using new inputs from a arbitrary phase. A third option permits artificial rapid close to steady state and a fourth option permits searching optimum input to satisfy both of specification and recovery rate of product. MIXSET handles maximum chemical system of eight components with or without mutual dependence of the distribution of the components. The chemical system in MIXSET includes chemical reactions and/or decaying reaction. Distribution data can be supplied by third-power polynominal equations or tables, and kinetic data by tables or given constants. The fluctuation of the interfacial level height in settler is converted into the flow rate changes of organic and aqueous stream to follow dynamic behavior of extraction process in detail. MIXSET can be applied to flowsheet study, start up and/or shut down procedure study and real time process management in countercurrent solvent extraction processes. (auth.)

  15. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

    A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)

  16. Analytical stress tensor and pressure calculations with the CRYSTAL code

    OpenAIRE

    Doll, Klaus

    2010-01-01

    Abstract The calculation of the stress tensor and related properties and its implementation in the CRYSTAL code are described. The stress tensor is obtained from the earlier implemented analytical gradients with respect to the cell parameters. Subsequently, the pressure and enthalpy is computed, and a test concerning the pressure driven phase transition in KI is used as an illustration. Finally, the possibility of applying external pressure is implemented. The ...

  17. The VADMAP code to calculate the SAF of photon

    International Nuclear Information System (INIS)

    A computer code VADMAP has been developed to calculate the Specific Absorbed Fraction, SAF, of photon. The development of the code is aimed at efficient and systematic preparation of the SAF data files for several different human phantoms in a suitable form as a direct input data file to DOSimetric DAta Calculation system, DOSDAC, which is being developed at Japan Atomic Energy Research Institute, JAERI. This document describes the methodology used in the code, the code structure, user's information including the way of implementing the code on FACOM/M-380, and the performance through calculation and preparation of the SAF data file. In order to show the performance of the code, a set of the SAF values for an adult human phantom was calculated and was organized to prepare the SAF file. Comparing the calculated SAF values with those tabulated in ORNL-5000, the quality of the code was examined. (author)

  18. MOx benchmark calculations by deterministic and Monte Carlo codes

    International Nuclear Information System (INIS)

    Highlights: ► MOx based depletion calculation. ► Methodology to create continuous energy pseudo cross section for lump of minor fission products. ► Mass inventory comparison between deterministic and Monte Carlo codes. ► Higher deviation was found for several isotopes. - Abstract: A depletion calculation benchmark devoted to MOx fuel is an ongoing objective of the OECD/NEA WPRS following the study of depletion calculation concerning UOx fuels. The objective of the proposed benchmark is to compare existing depletion calculations obtained with various codes and data libraries applied to fuel and back-end cycle configurations. In the present work the deterministic code NEWT/ORIGEN-S of the SCALE6 codes package and the Monte Carlo based code MONTEBURNS2.0 were used to calculate the masses of inventory isotopes. The methodology to apply the MONTEBURNS2.0 to this benchmark is also presented. Then the results from both code were compared.

  19. CONDOR: neutronic code for fuel elements calculation with rods

    International Nuclear Information System (INIS)

    CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)

  20. Xenon poisoning calculation code for miniature neutron source reactor (MNSR)

    International Nuclear Information System (INIS)

    In line with the actual requirements and based upon the specific characteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poisoning Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data

  1. Xenon poisoning calculation code for miniature neutron source reactor (MNSR)

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In line with the actual requirements and based upon the specific char acteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poison ing Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data.

  2. TEA: A Code for Calculating Thermochemical Equilibrium Abundances

    OpenAIRE

    Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver

    2015-01-01

    We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. (1958) and Eriksson (1971). It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature-pressure pairs. We tested the code against the method o...

  3. Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.

  4. Development and validation of a nodal code for core calculation

    International Nuclear Information System (INIS)

    The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency

  5. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

    The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, SN transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)

  6. TEA: A Code for Calculating Thermochemical Equilibrium Abundances

    CERN Document Server

    Blecic, Jasmina; Bowman, M Oliver

    2015-01-01

    We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. (1958) and Eriksson (1971). It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature-pressure pairs. We tested the code against the method of Burrows & Sharp (1999), the free thermochemical equilibrium code CEA (Chemical Equilibrium with Applications), and the example given by White et al. (1958). Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is ...

  7. MCOR - Monte Carlo depletion code for reference LWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

    2011-04-15

    Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally

  8. Calculation codes in radiation protection, radiation physics and dosimetry

    International Nuclear Information System (INIS)

    These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)

  9. Description of the CAREM Reactor Neutronic Calculation Codes

    International Nuclear Information System (INIS)

    In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included

  10. Calculation codes in radiation protection, radiation physics and dosimetry; Codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)

  11. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  12. CONSUL code package application for LMFR core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)

    2008-07-01

    CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)

  13. Progress on burnup calculation methods coupling Monte Carlo and depletion codes

    Energy Technology Data Exchange (ETDEWEB)

    Leszczynski, Francisco [Comision Nacional de Energia Atomica, San Carlos de Bariloche, RN (Argentina). Centro Atomico Bariloche]. E-mail: lesinki@cab.cnea.gob.ar

    2005-07-01

    Several methods of burnup calculations coupling Monte Carlo and depletion codes that were investigated and applied for the author last years are described. here. Some benchmark results and future possibilities are analyzed also. The methods are: depletion calculations at cell level with WIMS or other cell codes, and use of the resulting concentrations of fission products, poisons and actinides on Monte Carlo calculation for fixed burnup distributions obtained from diffusion codes; same as the first but using a method o coupling Monte Carlo (MCNP) and a depletion code (ORIGEN) at a cell level for obtaining the concentrations of nuclides, to be used on full reactor calculation with Monte Carlo code; and full calculation of the system with Monte Carlo and depletion codes, on several steps. All these methods were used for different problems for research reactors and some comparisons with experimental results of regular lattices were performed. On this work, a resume of all these works is presented and discussion of advantages and problems found are included. Also, a brief description of the methods adopted and MCQ system for coupling MCNP and ORIGEN codes is included. (author)

  14. Hot zero power reactor calculations using the Insilico code

    Science.gov (United States)

    Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.

    2016-06-01

    In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

  15. The MCEF code for nuclear evaporation and fission calculations

    Energy Technology Data Exchange (ETDEWEB)

    Deppman, A.; Pina, S.R. de; Likhachev, V.P.; Mesa, J. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica; Tavares, O.A.P.; Duarte, S.B.; Oliveira, E.C. de [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Arruda-Neto, J.D.T. [Universidade Santo Amaro (UNISA), SP (Brazil); Rodriguez, O. [Instituto Superior de Ciencias y Tecnologia Nucleares, La Habana (Cuba); Goncalves, M. [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil)

    2001-11-01

    We present an object oriented algorithm, written in the Java programming language, which performs a Monte Carlo calculation of the evaporation-fission process taking place inside an excited nucleus. We show that this nuclear physics problem is very suited for the object oriented programming by constructing two simple objects: one that handles all nuclear properties and another that takes care of the nuclear reaction. The MCEF code was used to calculate important results for nuclear reactions, and here we show examples of possible uses for this code. (author)

  16. The MCEF code for nuclear evaporation and fission calculations

    International Nuclear Information System (INIS)

    We present an object oriented algorithm, written in the Java programming language, which performs a Monte Carlo calculation of the evaporation-fission process taking place inside an excited nucleus. We show that this nuclear physics problem is very suited for the object oriented programming by constructing two simple objects: one that handles all nuclear properties and another that takes care of the nuclear reaction. The MCEF code was used to calculate important results for nuclear reactions, and here we show examples of possible uses for this code. (author)

  17. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  18. Decay heat calculation an international nuclear code comparison

    International Nuclear Information System (INIS)

    The results of an international code comparison on decay heat are presented and discussed. Participants from more than ten laboratories calculated, using the same input data, decay heat for thirteen cooling times between 1 and 1013 sec. Two irradiation cases were proposed: fission pulse and 3x107 seconds of irradiation of 235U fuel. The results are analysed and compared. This inter-comparison shows that, if the same input data are given, most of the codes give very similar results for the decay heat and consequently also for the fission product contribution

  19. Calculation code PULCO for Purex process in pulsed column

    International Nuclear Information System (INIS)

    The calculation code PULCO, which can simulate the Purex process using a pulsed column as an extractor, has been developed. The PULCO is based on the fundamental concept of mass transfer that the mass transfer within a pulsed column occurs through the interface of liquid drops and continuous phase fluid, and is the calculation code different from conventional ones, by which various phenomena such as the generation of liquid drops, their rising and falling, and the unification of liquid drops actually occurring in a pulsed column are exactly reflected and can be correctly simulated. In the PULCO, the actually measured values of the fundamental quantities representing the extraction behavior of liquid drops in a pulsed column are incorporated, such as the mass transfer coefficient of each component, the diameter and velocity of liquid drops in a pulsed column, the holdup of dispersed phase, and axial turbulent flow diffusion coefficient. The verification of the results calculated with the PULCO was carried out by installing a pulsed column of 50 mm inside diameter and 2 m length with 40 plate stage in a glove box for unirradiated uranium-plutonium mixed system. The results of the calculation and test were in good agreement, and the validity of the PULCO was confirmed. (Kako, I.)

  20. Comparison of computer code calculations with FEBA test data

    International Nuclear Information System (INIS)

    The FEBA forced feed reflood experiments included base line tests with unblocked geometry. The experiments consisted of separate effect tests on a full-length 5x5 rod bundle. Experimental cladding temperatures and heat transfer coefficients of FEBA test No. 216 are compared with the analytical data postcalculated utilizing the SSYST-3 computer code. The comparison indicates a satisfactory matching of the peak cladding temperatures, quench times and heat transfer coefficients for nearly all axial positions. This agreement was made possible by the use of an artificially adjusted value of the empirical code input parameter in the heat transfer for the dispersed flow regime. A limited comparison of test data and calculations using the RELAP4/MOD6 transient analysis code are also included. In this case the input data for the water entrainment fraction and the liquid weighting factor in the heat transfer for the dispersed flow regime were adjusted to match the experimental data. On the other hand, no fitting of the input parameters was made for the COBRA-TF calculations which are included in the data comparison. (orig.)

  1. Tokamak plasma power balance calculation code (TPC code) outline and operation manual

    International Nuclear Information System (INIS)

    This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (author)

  2. ANIGAM: a computer code for the automatic calculation of nuclear group data

    International Nuclear Information System (INIS)

    The computer code ANIGAM consists mainly of the well-known programmes GAM-I and ANISN as well as of a subroutine which reads the THERMOS cross section library and prepares it for ANISN. ANIGAM has been written for the automatic calculation of microscopic and macroscopic cross sections of light water reactor fuel assemblies. In a single computer run both were calculated, the cross sections representative for fuel assemblies in reactor core calculations and the cross sections of each cell type of a fuel assembly. The calculated data were delivered to EXTERMINATOR and CITATION for following diffusion or burn up calculations by an auxiliary programme. This report contains a detailed description of the computer codes and methods used in ANIGAM, a description of the subroutines, of the OVERLAY structure and an input and output description. (oririg.)

  3. Adjoint Monte Carlo techniques and codes for organ dose calculations

    International Nuclear Information System (INIS)

    Adjoint Monte Carlo simulations can be effectively used for the estimation of doses in small targets when the sources are extended in large volumes or surfaces. The main features of two computer codes for calculating doses at free points or in organs of an anthropomorphic phantom are described. In the first program (REBEL-3) natural gamma-emitting sources are contained in the walls of a dwelling room; in the second one (POKER-CAMP) the user can specify arbitrary gamma sources with different spatial distributions in the environment: in (or on the surface of) the ground and in the air. 3 figures

  4. Saphyr: a code system from reactor design to reference calculations

    International Nuclear Information System (INIS)

    In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels

  5. Saphyr: a code system from reactor design to reference calculations

    Energy Technology Data Exchange (ETDEWEB)

    Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)

    2003-07-01

    In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.

  6. Burnup calculations using serpent code in accelerator driven thorium reactors

    International Nuclear Information System (INIS)

    In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232Th and mixed 233U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)

  7. Burnup calculations using serpent code in accelerator driven thorium reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, M.E.; Agar, O. [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Physics Dept.; Yigit, M. [Aksaray Univ. (Turkey). Physics Dept.

    2013-07-15

    In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed {sup 232}Th and mixed {sup 233}U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)

  8. Miniature neutron source reactor burnup calculations using IRBURN code system

    International Nuclear Information System (INIS)

    Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.

  9. Vectorization of nuclear codes for atmospheric transport and exposure calculation of radioactive materials

    International Nuclear Information System (INIS)

    Three computer codes MATHEW, ADPIC of LLNL and GAMPUL of JAERI for prediction of wind field, concentration and external exposure rate of airborne radioactive materials are vectorized and the results are presented. Using the continuous equation of incompressible flow as a constraint, the MATHEW calculates the three dimensional wind field by a variational method. Using the particle-in -cell method, the ADPIC calculates the advection and diffusion of radioactive materials in three dimensional wind field and terrain, and gives the concentration of the materials in each cell of the domain. The GAMPUL calculates the external exposure rate assuming Gaussian plume type distribution of concentration. The vectorized code MATHEW attained 7.8 times speedup by a vector processor FACOM230-75 APU. The ADPIC and GAMPUL are estimated to attain 1.5 and 4 times speedup respectively on CRAY-1 type vector processor. (author)

  10. Fragmentation calculation by intranuclear-cascade-evaporation code

    Energy Technology Data Exchange (ETDEWEB)

    Shigyo, Nobuhiro; Iga, Kiminori; Ishibashi, Kenji [Kyushu Univ., Fukuoka (Japan)

    1997-03-01

    High Energy Transport Code (HETC) based on the intranuclear-cascade-evaporation model is modified for calculating the fragmentation cross section. For the intranuclear-cascade process, nucleon-nucleon cross sections are used for collision computation; effective in-medium-corrected cross sections are adopted instead of the original free-nucleon collision. The exciton model is adopted for improvement of backward nucleon-emission cross section for low-energy nucleon-incident events. The fragmentation reaction is incorporated into the original HETC as a subroutine set by the use of the systematics of the reaction. The modified HETC (HETC-3STEP/FRG) reproduces experimental fragment yields to a reasonable degree. (author)

  11. Comparison of computer codes for calculating dynamic loads in wind turbines

    Science.gov (United States)

    Spera, D. A.

    1978-01-01

    The development of computer codes for calculating dynamic loads in horizontal axis wind turbines was examined, and a brief overview of each code was given. The performance of individual codes was compared against two sets of test data measured on a 100 KW Mod-0 wind turbine. All codes are aeroelastic and include loads which are gravitational, inertial and aerodynamic in origin.

  12. 14 CFR 234.8 - Calculation of on-time performance codes.

    Science.gov (United States)

    2010-01-01

    ... (AVIATION PROCEEDINGS) ECONOMIC REGULATIONS AIRLINE SERVICE QUALITY PERFORMANCE REPORTS § 234.8 Calculation of on-time performance codes. (a) Each reporting carrier shall calculate an on-time performance code... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Calculation of on-time performance...

  13. NUFACE: An interface code for the calculation of nuclear responses

    International Nuclear Information System (INIS)

    The NUFACE interface code computes nuclear responses for use in the nuclear analysis of a given tokamak reactor design. The NUFACE code operates on the neutron and gamma fluxes provided by the one-dimensional neutral-particle transport code ONEDANT. Zonewise and zone-boundary responses are computed to obtain both zone-integrated values and maximum surface values. Information on each material mixture within a zone and on each element or isotope constituent of each material is computed. This feature allows for a detailed analysis of the reactor whereby one can easily identify the fractional contribution to the response of interest from each material and each element or isotope. 4 refs., 4 figs., 3 tabs

  14. NUFACE: An interface code for the calculation of nuclear responses

    Energy Technology Data Exchange (ETDEWEB)

    Henderson, D.L. (Oak Ridge National Lab., TN (USA)); Gomes, I.C. (Tennessee Univ., Knoxville, TN (USA))

    1990-01-01

    The NUFACE interface code computes nuclear responses for use in the nuclear analysis of a given tokamak reactor design. The NUFACE code operates on the neutron and gamma fluxes provided by the one-dimensional neutral-particle transport code ONEDANT. Zonewise and zone-boundary responses are computed to obtain both zone-integrated values and maximum surface values. Information on each material mixture within a zone and on each element or isotope constituent of each material is computed. This feature allows for a detailed analysis of the reactor whereby one can easily identify the fractional contribution to the response of interest from each material and each element or isotope. 4 refs., 4 figs., 3 tabs.

  15. Hacking the genetic code of mammalian cells.

    Science.gov (United States)

    Schwarzer, Dirk

    2009-07-01

    A genetic shuttle: The highlighted article, which was recently published by Schultz, Geierstanger and co-workers, describes a straightforward scheme for enlarging the genetic code of mammalian cells. An orthogonal tRNA/aminoacyl-tRNA synthetase pair specific for a new amino acid can be evolved in E. coli and subsequently transferred into mammalian cells. The feasibility of this approach was demonstrated by adding a photocaged lysine derivative to the genetic repertoire of a human cell line.

  16. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  17. MUXS: a code to generate multigroup cross sections for sputtering calculations

    International Nuclear Information System (INIS)

    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

  18. MUXS: a code to generate multigroup cross sections for sputtering calculations

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Development of the multistep compound process calculation code

    Energy Technology Data Exchange (ETDEWEB)

    Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan)

    1998-03-01

    A program `cmc` has been developed to calculate the multistep compound (MSC) process by Feshback-Kerman-Koonin. A radial overlap integral in the transition matrix element is calculated microscopically, and comparisons are made for neutron induced {sup 93}Nb reactions. Strengths of the two-body interaction V{sub 0} are estimated from the total MSC cross sections. (author)

  20. KARATE - a code for VVER-440 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Gado, J.; Hegedus, Cs.J.; Hegyi, Gy.; Kereszturi, A.; Makai, M.; Maraczi, Cs.; Telbisz, M.

    1994-12-31

    A modular calculation system has been elaborated at the KFKI Atomic Energy Research Institute for VVER-440 cores. The purpose of KARATE is the calculation of neutron physical and thermal-hydraulic processes in the core at normal, startup, and slow transient conditions. KARATE is under validation and verification (V&V) against mathematical, experimental, and operational data.

  1. SHETEMP: a computer code for calculation of fuel temperature behavior under reactivity initiated accidents

    International Nuclear Information System (INIS)

    A fast running computer code SHETEMP has been developed for analysis of reactivity initiated accidents under constant core cooling conditions such as coolant temperature and heat transfer coefficient on fuel rods. This code can predict core power and fuel temperature behaviours. A control rod movement can be taken into account in power control system. The objective of the code is to provide fast running capability with easy handling of the code required for audit and design calculations where a large number of calculations are performed for parameter surveys during short time period. The fast running capability of the code was realized by neglection of fluid flow calculation. The computer code SHETEMP was made up by extracting and conglomerating routines for reactor kinetics and heat conduction in the transient reactor thermal-hydraulic analysis code ALARM-P1, and by combining newly developed routines for reactor power control system. As ALARM-P1, SHETEMP solves point reactor kinetics equations by the modified Runge-Kutta method and one-dimensional transient heat conduction equations for slab and cylindrical geometries by the Crank-Nicholson methods. The model for reactor power control system takes into account effects of PID regulator and control rod drive mechanism. In order to check errors in programming of the code, calculated results by SHETEMP were compared with analytic solution. Based on the comparisons, the appropriateness of the programming was verified. Also, through a sample calculation for typical modelling, it was concluded that the code could satisfy the fast running capability required for audit and design calculations. This report will be described as a code manual of SHETEMP. It contains descriptions on a sample problem, code structure, input data specifications and usage of the code, in addition to analytical models and results of code verification calculations. (author)

  2. SAMDIST: A computer code for calculating statistical distributions for R-matrix resonance parameters

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.; Larson, N.M.

    1995-09-01

    The SAMDIST computer code has been developed to calculate distribution of resonance parameters of the Reich-Moore R-matrix type. The program assumes the parameters are in the format compatible with that of the multilevel R-matrix code SAMMY. SAMDIST calculates the energy-level spacing distribution, the resonance width distribution, and the long-range correlation of the energy levels. Results of these calculations are presented in both graphic and tabular forms.

  3. Determination of Solution Accuracy of Numerical Schemes as Part of Code and Calculation Verification

    Energy Technology Data Exchange (ETDEWEB)

    Blottner, F.G.; Lopez, A.R.

    1998-10-01

    This investigation is concerned with the accuracy of numerical schemes for solving partial differential equations used in science and engineering simulation codes. Richardson extrapolation methods for steady and unsteady problems with structured meshes are presented as part of the verification procedure to determine code and calculation accuracy. The local truncation error de- termination of a numerical difference scheme is shown to be a significant component of the veri- fication procedure as it determines the consistency of the numerical scheme, the order of the numerical scheme, and the restrictions on the mesh variation with a non-uniform mesh. Genera- tion of a series of co-located, refined meshes with the appropriate variation of mesh cell size is in- vestigated and is another important component of the verification procedure. The importance of mesh refinement studies is shown to be more significant than just a procedure to determine solu- tion accuracy. It is suggested that mesh refinement techniques can be developed to determine con- sistency of numerical schemes and to determine if governing equations are well posed. The present investigation provides further insight into the conditions and procedures required to effec- tively use Richardson extrapolation with mesh refinement studies to achieve confidence that sim- ulation codes are producing accurate numerical solutions.

  4. Radiative Transfer Code: Application to the calculation of PAR

    Indian Academy of Sciences (India)

    D Emmanuel; D Phillippe; C Malik

    2000-12-01

    The production of carbon in the ocean, the so-called primary production, depends on various physico- biological parameters: the biomass and nutrient amounts in oceans, the salinity and temperature of the water and the light available in the water column. We focus on the visible spectrum of the solar radiation defined as the Photosynthetically Active Radiation (PAR). We developed a model (Chami et al. 1997) to simulate the behavior of the solar beam in the atmosphere and the ocean. We first describe the theoretical basis of the code and the method we used to solve the radiative transfer equation (RTE): the successive orders of scattering (SO). The second part deals with a sensitivity study of the PAR just above and below the sea surface for various atmospheric conditions. In a cloudy sky, we computed a ratio between vector fluxes just above the sea surface and spherical fluxes just beneath the sea surface. When the optical thickness of the cloud increases this ratio remains constant and around 1.29. This parameter is convenient to convert vector flux at the sea surface as retrieved from satellite to PAR. Subsequently, we show how solar radiation as vector flux rather than PAR leads to an underestimate of the primary production up to 40% for extreme cases.

  5. Calculation of conversion coefficients Hp(3)/K air using the PENELOPE Monte Carlo code and comparison with MCNP calculation results

    International Nuclear Information System (INIS)

    The authors report calculations performed using the MNCP and PENELOPE codes to determine the Hp(3)/K air conversion coefficient which allows the Hp(3) dose equivalent to be determined from the measured value of the kerma in the air. They report the definition of the phantom, a 20 cm diameter and 20 cm high cylinder which is considered as representative of a head. Calculations are performed for an energy range corresponding to interventional radiology or cardiology (20 keV-110 keV). Results obtained with both codes are compared

  6. Quasiparticle GW calculations within the GPAW electronic structure code

    DEFF Research Database (Denmark)

    Hüser, Falco

    properties are to a large extent governed by the physics on the atomic scale, that means pure quantum mechanics. For many decades, Density Functional Theory has been the computational method of choice, since it provides a fairly easy and yet accurate way of determining electronic structures and related...... is considered, which can be regarded as the lowest level of the GW approximation. This thesis documents the implementation of the G0W0 approximation in GPAW. It serves two purposes: First, it can be read as a manual by anyone who is interested in doing GW calculations with GPAW. All features and requirements...

  7. Development of leak rate calculation model and code in piping

    International Nuclear Information System (INIS)

    Background: With the development of fracture mechanics, Leak-Before-Break (LBB) is widely used in nuclear power plant piping design. Purpose: In order to support the application of LBB, leak rate through crack need to be calculated. Methods: In this text, an analytical flow model is developed based on homogeneous non-equilibrium model, a computer program is also developed based on this analytical model. Results: Comparison between the results from the above program and test results shows that deviations of calc. results to test results are within ±50%. Conclusions: Conclusions can be got that this analytical model and computer program meet the requirement of engineering application. (authors)

  8. Validation of the Monteburns code for criticality calculation of TRIGA reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Jeraj, Robert [Jozef Stafan Institute, Ljubljana (Slovenia)

    2002-07-01

    Use of Monte Carlo methods in burnup calculations of nuclear fuel has become practical due to increased speed of computers. Monteburns is an automated computational tool that links the Monte Carlo code MCNP with the burnup and decay code ORIGEN2.1. This code system was used to simulate a criticality benchmark experiment with burned fuel on a TRIGA Mark II research reactor. Two core configurations were simulated and k{sub eff} values calculated. The comparison between the calculated and experimental values shows good agreement, which indicates that the MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulations of TRIGA reactors. (author)

  9. Comparison between SERPENT and MONTEBURNS codes applied to burnup calculations of a GFR-like configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chersola, Davide [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Lomonaco, Guglielmo, E-mail: guglielmo.lomonaco@unige.it [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Marotta, Riccardo [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Mazzini, Guido [Centrum výzkumu Řež (Research Centre Rez), Husinec-Rez, cp. 130, 25068 Rez (Czech Republic)

    2014-07-01

    Highlights: • MC codes are widely adopted to analyze nuclear facilities, including GEN-IV reactors. • Burnup calculations are an efficient tool to test neutronic Monte Carlo codes. • In this comparison the used codes show some differences but a good agreement exists. - Abstract: This paper presents the comparison between two Monte Carlo based burnup codes: SERPENT and MONTEBURNS. Monte Carlo codes are fully and worldwide adopted to perform analyses on nuclear facilities, also in the frame of Generation IV advanced reactors simulations. Thus, faster and most powerful calculation codes are needed with the aim to analyze complex geometries and specific neutronic behaviors. Burnup calculations are an efficient tool to test neutronic Monte Carlo codes: indeed these calculations couple transport and depletion procedures, so that neutronic reactor behavior can be simulated in its totality. Comparisons have been performed on a configuration representing the Allegro MOX 75 MW{sub th} reactor proposed by the European GoFastR (Gas-cooled Fast Reactor) Project in the frame of the 7th Euratom Framework Program. Although in burnup and criticality comparisons the codes used in simulations show different calculation times and some differences in amounts and in precision (in term of statistical errors), a reasonably good agreement between them exists.

  10. Code of hybrid calculation for the study of heat exchangers

    International Nuclear Information System (INIS)

    A series integration method has been chosen, machine time being similar to space, the computer storing the solutions passed to allow for an approximation at the finished differences of the timed derivatives. Space distributions of the functions of state (enthalpy and temperature) of the two fluids are calculated by analogical integration in the direction of the circulation of the fluids. This is made possible by the disconnection of the two fluids thanks to a method of prediction-correction for the equation of the wall. The great rapidity of the analogical integrations and the absence of iterations make the calculation in actual time possible (assuming that 200 points are taken for the storage of the solutions passed, the time required for a full calculation is 0.4 second). It is, therefore, possible to integrate this simulation in an actual (for example Nuclear Station) or simulated loop. The exchanger may or may not be with counter-current, with or without change of phase. Thermodynamic tables with two variables are introduced into the computer, interpolation is effected in analogical and this allows, inter alia, enthalpy, temperature, pressure and volumic mass to be perfectly in phase one with the other. The work bears most particularly on the accuracy and stability of simulation of the timed derivatives which are obtained by difference between an analogical function and the same function stored in the computer at the preceding step. Particular emphasis is made to bear on the operation of the model, a monitor program controlling the various phases of utilization in function of the type of exchanger, experiments to be made; a program of control of the results enables the operator to obtain the results either immediately in printed form, curve tracer or oscilloscope, put to scale, or in deferred a more elaborate processing of the results. The same program allows for the simulation of any exchanger (different fluid, variable geometry, etc.) or any number of

  11. Aspects of cell calculations in deterministic reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Varvayanni, M. [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Savva, P., E-mail: savvapan@ipta.demokritos.gr [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N. [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    {Tau}he capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available

  12. User's guide for vectorized code EQUIL for calculating equilibrium chemistry on Control Data STAR-100 computer

    Science.gov (United States)

    Kumar, A.; Graves, R. A., Jr.; Weilmuenster, K. J.

    1980-01-01

    A vectorized code, EQUIL, was developed for calculating the equilibrium chemistry of a reacting gas mixture on the Control Data STAR-100 computer. The code provides species mole fractions, mass fractions, and thermodynamic and transport properties of the mixture for given temperature, pressure, and elemental mass fractions. The code is set up for the electrons H, He, C, O, N system of elements. In all, 24 chemical species are included.

  13. Chromaticity calculations and code comparisons for x-ray lithography source XLS and SXLS rings

    International Nuclear Information System (INIS)

    This note presents the chromaticity calculations and code comparison results for the (x-ray lithography source) XLS (Chasman Green, XUV Cosy lattice) and (2 magnet 4T) SXLS lattices, with the standard beam optic codes, including programs SYNCH88.5, MAD6, PATRICIA88.4, PATPET88.2, DIMAD, BETA, and MARYLIE. This analysis is a part of our ongoing accelerator physics code studies. 4 figs., 10 tabs

  14. Chromaticity calculations and code comparisons for x-ray lithography source XLS and SXLS rings

    Energy Technology Data Exchange (ETDEWEB)

    Parsa, Z.

    1988-06-16

    This note presents the chromaticity calculations and code comparison results for the (x-ray lithography source) XLS (Chasman Green, XUV Cosy lattice) and (2 magnet 4T) SXLS lattices, with the standard beam optic codes, including programs SYNCH88.5, MAD6, PATRICIA88.4, PATPET88.2, DIMAD, BETA, and MARYLIE. This analysis is a part of our ongoing accelerator physics code studies. 4 figs., 10 tabs.

  15. TEMP: a computer code to calculate fuel pin temperatures during a transient

    International Nuclear Information System (INIS)

    The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method

  16. HADOC: a computer code for calculation of external and inhalation doses from acute radionuclide releases

    Energy Technology Data Exchange (ETDEWEB)

    Strenge, D.L.; Peloquin, R.A.

    1981-04-01

    The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested.

  17. Application of Monte Carlo code EGS4 to calculate gamma exposure buildup factors

    International Nuclear Information System (INIS)

    Exposure buildup factors up to 40 mean free paths ranging from 0.015 MeV to 15 MeV photon energy were calculated by using the Monte Carlo simulation code EGS4 for ordinary concrete. The calculation involves PHOTX cross section library, a point isotropic source, infinite uniform medium model and a particle splitting method and considers the Bremsstrahlung, fluorescent effect, correlative (Rayleigh) scatter. The results were compared with the relevant data. Results show that the data of the buildup factors calculated by the Monte Carlo code EGS4 was reliable. The Monte Carlo method can be used widely to calculate gamma-ray exposure buildup factors. (authors)

  18. HADOC: a computer code for calculation of external and inhalation doses from acute radionuclide releases

    International Nuclear Information System (INIS)

    The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested

  19. THIDA-2: an advanced code system for calculation of transmutation, activation, decay heat and dose rate

    International Nuclear Information System (INIS)

    In a D-T burning fusion reactor, the radioactivity induced by the 14 MeV neutrons causes many problems. It limits personnel access to the reactor during shutdown, generates decay heat and produces radwastes. A code system THIDA had been developed in 1978 to calculate the radioactivity and dose rate around a fusion device. The THIDA system consisted of the followings: one- and two-dimensional discrete ordinates radiation transport codes; induced activity calculation code; three libraries for transmutation and decay chain data, transmutation cross sections and delayed gamma-ray emission data. The present report gives a complete description of THIDA-2, a new advanced version of the THIDA system which has the following major improvements: 1. Capability to treat three-dimensional calculation models by the use of a Monte Carlo transport code. 2. Accurate decay heat calculation following the transport of delayed gamma rays. 3. Simplification of the data input process by the use of free format scheme and closer coupling between the radiation transport codes and the induced activity calculation code. 4. Self-descriptive output format and additional plotter output. 5. Capability to calculate problems requiring larger core memory by the use of variable dimension. (author)

  20. Boltzmann-Fokker-Planck calculations using standard discrete-ordinates codes

    International Nuclear Information System (INIS)

    The Boltzmann-Fokker-Planck (BFP) equation can be used to describe both neutral and charged-particle transport. Over the past several years, the author and several collaborators have developed methods for representing Fokker-Planck operators with standard multigroup-Legendre cross-section data. When these data are input to a standard S/sub n/ code such as ONETRAN, the code actually solves the Boltzmann-Fokker-Planck equation rather than the Boltzmann equation. This is achieved wihout any modification to the S/sub n/ codes. Because BFP calculations can be more demanding from a numerical viewpoint than standard neutronics calculations, we have found it useful to implement new quadrature methods ad convergence acceleration methods in the standard discrete-ordinates code, ONETRAN. We discuss our BFP cross-section representation techniques, our improved quadrature and acceleration techniques, and present results from BFP coupled electron-photon transport calculations performed with ONETRAN. 19 refs., 7 figs

  1. Comparison of code calculations with experiments on containment response during LOCA conditions

    International Nuclear Information System (INIS)

    A series of experiments were performed on a one-tenth scale model of PHWR containment, incorporating pressure suppression system. The pressure-temperature transients in the model containment observed during simulated LOCA (Loss of Coolant) blowdown conditions were compared against calculated results form computer code PACSR, for purposes of verification of the code. Comparison of results indicated that calculated values of peak pressure in various compartment were significantly higher than observed ones. This disagreement was attributed mainly to modelling for energy absorption from containment atmosphere to structural surfaces, this effect being particularly important in a scaled down model. Good agreement between calculation and experiment was obtained after heat transfer correlation for energy absorption on surfaces were modified in the code. The study demonstrates the conservatism of the results from the code. (author). 6 refs., 1 tab., 9 figs

  2. RADIATION DOSE CALCULATION FOR FUEL HANDLING FACILITY CLOSURE CELL EQUIPMENT

    International Nuclear Information System (INIS)

    This calculation evaluates the energy deposition rates in silicon, gamma and neutron flux spectra at various locations of interest throughout FHF closure cell. The physical configuration features a complex geometry, with particle flux attenuation of many orders of magnitude that cannot be modeled by computer codes that use deterministic methods. Therefore, in this calculation the Monte Carlo method was used to solve the photon and neutron transport. In contrast with the deterministic methods, Monte Carlo does not solve an explicit transport equation, but rather obtain answers by simulating individual particles, recording the aspects of interest of their average behavior, and estimates the statistical precision of the results

  3. Manual of Nucost 1.0 - code for calculation of nuclear power generation costs

    International Nuclear Information System (INIS)

    Nucost is a computer code developed at CDTN to perform cost calculation of electric power generated in PWR nuclear power plants, based on present worth cost method. The Nucost version 1.0 performs calculations of nuclear fuel cost cycle by cycle during the time life of the power plant. That calculation is performed with enough details permitting optimization and minimization. The code is also a tool to aid reload projects and economic operation of PWR reactors. This manual presents a description of Nucost version 1.0, instruction to enter data preparation and description of the Nucost output. (M.I.)

  4. Computer codes in nuclear safety, radiation transport and dosimetry; Les codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M

    2006-07-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.

  5. Development of the Joyo MK-II core bowing reactivity calculation code

    Energy Technology Data Exchange (ETDEWEB)

    Tabuchi, Shiro; Torimaru, Tadahiko; Yoshida, Akihiro; Aoyama, Takafumi [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-09-01

    The study on the passive safety test by using the Experimental Fast Reactor Joyo has been performed to demonstrate the inherent safety of fast breeder reactors. In this study, emphasis was placed on the improvement on the accuracy of the feedback reactivity analysis. As a bowing reactivity might play a significant roll in ATWS analysis because of its effectively short time constant and relatively large magnitude, an emphasis was placed upon the evaluation of the analysis precision of bowing reactivity. Taking into account of the refueling and irradiation history of the individual core component, the core bowing behavior in Joyo has been analyzed by using the MK-II core management code system MAGI, the interface code TETRAS which interpolate neutron flux and coolant temperature at the position of wrapper tube, and the core bowing calculation code BEACON. Calculation accuracy of above mentioned system was evaluated through the comparison of calculated and measured permanent distortion of subassemblies. In 1996, core bowing reactivity was calculated by AURORA code using the above calculated bowing behavior of individual core component as input. But because an approximate two dimensional material reactivity worth map was utilized in AURORA, it was made clear that some amount of error caused by extrapolation could not be neglected. Therefore calculation code ARCHCOM (Analysis of Reactivity Change due to Core Mechanics) which utilize three dimensional material reactivity worth map as input was developed for the Joyo MK-II core bowing reactivity calculation. This code reduces above mentioned extrapolation error that used to be occurred at isolated core component, such as control rod or irradiation rig and at the interface region between fuel and reflector which had sharp bowing reactivity worth gradient. (author)

  6. Methods, algorithms and computer codes for calculation of electron-impact excitation parameters

    CERN Document Server

    Bogdanovich, P; Stonys, D

    2015-01-01

    We describe the computer codes, developed at Vilnius University, for the calculation of electron-impact excitation cross sections, collision strengths, and excitation rates in the plane-wave Born approximation. These codes utilize the multireference atomic wavefunctions which are also adopted to calculate radiative transition parameters of complex many-electron ions. This leads to consistent data sets suitable in plasma modelling codes. Two versions of electron scattering codes are considered in the present work, both of them employing configuration interaction method for inclusion of correlation effects and Breit-Pauli approximation to account for relativistic effects. These versions differ only by one-electron radial orbitals, where the first one employs the non-relativistic numerical radial orbitals, while another version uses the quasirelativistic radial orbitals. The accuracy of produced results is assessed by comparing radiative transition and electron-impact excitation data for neutral hydrogen, helium...

  7. THEHYCO-3DT: Thermal hydrodynamic code for the 3 dimensional transient calculation of advanced LMFBR core

    Energy Technology Data Exchange (ETDEWEB)

    Vitruk, S.G.; Korsun, A.S. [Moscow Engineering Physics Institute (Russian Federation); Ushakov, P.A. [Institute of Physics and Power Engineering, Obninsk (R)] [and others

    1995-09-01

    The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors.

  8. A FORTRAN computer code for calculating flows in multiple-blade-element cascades

    Science.gov (United States)

    Mcfarland, E. R.

    1985-01-01

    A solution technique has been developed for solving the multiple-blade-element, surface-of-revolution, blade-to-blade flow problem in turbomachinery. The calculation solves approximate flow equations which include the effects of compressibility, radius change, blade-row rotation, and variable stream sheet thickness. An integral equation solution (i.e., panel method) is used to solve the equations. A description of the computer code and computer code input is given in this report.

  9. A computer code for calculations in the algebraic collective model of the atomic nucleus

    CERN Document Server

    Welsh, T A

    2016-01-01

    A Maple code is presented for algebraic collective model (ACM) calculations. The ACM is an algebraic version of the Bohr model of the atomic nucleus, in which all required matrix elements are derived by exploiting the model's SU(1,1) x SO(5) dynamical group. This, in particular, obviates the use of coefficients of fractional parentage. This paper reviews the mathematical formulation of the ACM, and serves as a manual for the code. The code makes use of expressions for matrix elements derived elsewhere and newly derived matrix elements of the operators [pi x q x pi]_0 and [pi x pi]_{LM}, where q_M are the model's quadrupole moments, and pi_N are corresponding conjugate momenta (-2>=M,N<=2). The code also provides ready access to SO(3)-reduced SO(5) Clebsch-Gordan coefficients through data files provided with the code.

  10. Citham a computer code for calculating fuel depletion-description, tests, modifications and evaluation

    International Nuclear Information System (INIS)

    The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author)

  11. An Efficient Group Key Management Using Code for Key Calculation for Simultaneous Join/Leave: CKCS

    Directory of Open Access Journals (Sweden)

    Melisa Hajyvahabzadeh

    2012-08-01

    Full Text Available This paper presents an efficient group key management protocol, CKCS (Code for Key Calculation in Simultaneous join/leave for simultaneous join/leave in secure multicast. This protocol is based on logical key hierarchy. In this protocol, when new members join the group simultaneously, server sends only thegroup key for those new members. Then, current members and new members calculate the necessary keys by node codes and one-way hash function. A node code is a random number which is assigned to each key to help users calculate the necessary keys. Again, at leave, the server just sends the new group key to remaining members. The results show that CKCS reduces computational and communication overhead, and also message size in simultaneous join/leave.

  12. Development of neutral transport lattice code DENT-2D and benchmark calculation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. S.; Kim, H. Y.; Ji, S. K. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    We developed new transport lattice code called DENT-2D (Deterministic Neutral Particle Transport Code in 2-D imensional Space)primarily to generate few- group constants for the reactor physics analysis diffusion codes. This code is designed to be coupled with KAERI reactor analysis nodal code, MASTER [1] ,to complete the design system package. CASMO-3 and HELIOS have been used in generating the few- group constant for MASTER. Currently DENT-2D includes only neutron particle transport calculation in 2-dimensional Cartesian geometry. The characteristics method is adopted for the spatial discretization, which is advantageous for the treatment of the complicated geometry structure and the highly anisotropic scattering. The subgroup method is used for the resonance treatment. B1 approximation has been used to obtain the criticality spectrum considering the leakage effect in the real core situation. The exponential matrix method has been used for the depletion calculation. The results of benchmark calculations show that the prediction capability of DENT-2D is comparable to the other lattice codes such as HELIOS and CASMO-3.

  13. Calculation code evaluating the confinement of a nuclear facility in case of fires

    Energy Technology Data Exchange (ETDEWEB)

    Laborde, J.C.; Prevost, C.; Vendel, J. [and others

    1995-02-01

    Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation.

  14. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    International Nuclear Information System (INIS)

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  15. The FLUFF code for calculating finned surface heat transfer -description and user's guide

    International Nuclear Information System (INIS)

    FLUFF is a computer code for calculating heat transfer from finned surfaces by convection and radiation. It can also represent heat transfer by radiation to a partially emitting and absorbing medium within the fin cavity. The FLUFF code is useful not only for studying the behaviour of finned surfaces but also for deriving heat fluxes which can be applied as boundary conditions to other heat transfer codes. In this way models of bodies with finned surfaces may be greatly simplified since the fins need not be explicitly represented. (author)

  16. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors

    International Nuclear Information System (INIS)

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d4 code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  17. VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, Pavel; Khrennikov, Nikolay [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) Malaya Krasnoselskaya ul., 2/8, bld. 5, 107140 Moscow (Russian Federation)

    2008-07-01

    Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)

  18. Development of continuous-energy eigenvalue sensitivity coefficient calculation methods in the shift Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Perfetti, C.; Martin, W. [Univ. of Michigan, Dept. of Nuclear Engineering and Radiological Sciences, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109-2104 (United States); Rearden, B.; Williams, M. [Oak Ridge National Laboratory, Reactor and Nuclear Systems Div., Bldg. 5700, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)

    2012-07-01

    Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the Shift Monte Carlo code within the SCALE code package. The methods were used for two small-scale test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods. (authors)

  19. Calculation of Quad-Cities Central Bundle Documented by the U.S. in FY98 Using Russian Computer Codes

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-06-19

    The report presents calculation results of isotopic composition of irradiated fuel performed for the Quad Cities-1 reactor bundle with UO{sub 2} and MOX fuel. The MCU-REA code was used for calculations. The code is developed in Kurchatov Institute, Russia. The MCU-REA results are compared with the experimental data and HELIOS code results.

  20. Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

    International Nuclear Information System (INIS)

    Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

  1. Benchmarking of the WIMSD/CITATION deterministic code system for the neutronic calculations of TRIGA Mark-III research reactors

    International Nuclear Information System (INIS)

    Highlights: • Benchmark study performed for the neutronic calculations of TRIGA research reactors. • WIMSD-5B/CITATION is the utilized code system along with the WIMSD-IAEA-69 library. • The studied condensed spectra are five and seven energy groups spectra. • Analyzed: lattice parameters, reactivities, CR worth, flux and power distribution. • The lattice and neutronic parameters showed the accuracy of both condensed spectra. - Abstract: The objective of this paper is to assess the suitability and accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA Mark-III research reactors using the WIMSD/CITATION code system in proposed condensed energy spectra of five and seven energy groups with one and three thermal groups respectively. The utilized cell transport calculations code and core diffusion calculations code are the WIMSD-5B and the CITVAP v3.1 codes respectively, along with the WIMSD-IAEA-69 nuclear data library. Firstly, the assessment goes through analyzing the integral parameters – keff, ρ238, δ235, δ238, and C* – of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark-III Thai research reactor at the condensed energy spectra. The results include the excess reactivities of the cores and the worth of selected control rods which were compared with reference Monte Carlo results and experimental values. The results show good agreement with the references at both energy spectra and the better accuracy are attainable in the five energy groups spectrum. The results also include neutron flux distributions which are evaluated for future comparisons with other calculational techniques even they are comparable to reactors and fuels of the same type. The

  2. Erosion corrosion in power plant piping systems - Calculation code for predicting wall thinning

    International Nuclear Information System (INIS)

    Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the operator of the plant are taken into consideration. Together with a continuously updated erosion corrosion data base the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to minimize material loss due to erosion corrosion, reduce non-destructive testing and curtail monitoring programs for piping systems, recommend life-extending measures. (author). 12 refs, 17 figs

  3. WOLF: a computer code package for the calculation of ion beam trajectories

    Energy Technology Data Exchange (ETDEWEB)

    Vogel, D.L.

    1985-10-01

    The WOLF code solves POISSON'S equation within a user-defined problem boundary of arbitrary shape. The code is compatible with ANSI FORTRAN and uses a two-dimensional Cartesian coordinate geometry represented on a triangular lattice. The vacuum electric fields and equipotential lines are calculated for the input problem. The use may then introduce a series of emitters from which particles of different charge-to-mass ratios and initial energies can originate. These non-relativistic particles will then be traced by WOLF through the user-defined region. Effects of ion and electron space charge are included in the calculation. A subprogram PISA forms part of this code and enables optimization of various aspects of the problem. The WOLF package also allows detailed graphics analysis of the computed results to be performed.

  4. SYN3D: a single-channel, spatial flux synthesis code for diffusion theory calculations

    Energy Technology Data Exchange (ETDEWEB)

    Adams, C. H.

    1976-07-01

    This report is a user's manual for SYN3D, a computer code which uses single-channel, spatial flux synthesis to calculate approximate solutions to two- and three-dimensional, finite-difference, multigroup neutron diffusion theory equations. SYN3D is designed to run in conjunction with any one of several one- and two-dimensional, finite-difference codes (required to generate the synthesis expansion functions) currently being used in the fast reactor community. The report describes the theory and equations, the use of the code, and the implementation on the IBM 370/195 and CDC 7600 of the version of SYN3D available through the Argonne Code Center.

  5. The solution of the LEU and MOX WWER-1000 calculation benchmark with the CARATE - multicell code

    International Nuclear Information System (INIS)

    Preparations for disposition of weapons grade plutonium in WWER-1000 reactors are in progress. Benchmark: Defined by the Kurchatov Institute (S. Bychkov, M. Kalugin, A. Lazarenko) to assess the applicability of computer codes for weapons grade MOX assembly calculations. Framework: 'Task force on reactor-based plutonium disposition' of OECD Nuclear Energy Agency. (Authors)

  6. CPS: a continuous-point-source computer code for plume dispersion and deposition calculations

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, K.R.; Crawford, T.V.; Lawson, L.A.

    1976-05-21

    The continuous-point-source computer code calculates concentrations and surface deposition of radioactive and chemical pollutants at distances from 0.1 to 100 km, assuming a Gaussian plume. The basic input is atmospheric stability category and wind speed, but a number of refinements are also included.

  7. BALTORO a general purpose code for coupling discrete ordinates and Monte-Carlo radiation transport calculations

    International Nuclear Information System (INIS)

    The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)

  8. TMRBAR: a code to calculate plasma parameters for tandem-mirror reactors operating in the MARS mode

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, R.B.

    1983-08-30

    The purpose of this report is to document the plasma power balance model currently used by LLNL to calculate steady state operating points for tandem mirror reactors. The code developed from this model, TMRBAR, has been used to predict the performance and define supplementary heating requirements for drivers used in the Mirror Advanced Reactor Study (MARS) and for the Fusion Power Demonstration (FPD) study. The equations solved included particle and energy balance for central cell and end cell species, quasineutrality at several cardinal points in the end cell region, as well as calculations of volumes, densities and average energies based on given constraints of beta profiles and fusion power output. Alpha particle ash is treated self-consistently, but no other impurity species is treated.

  9. IPEN/MB-01 heavy reflector benchmark calculations using Serpent code

    International Nuclear Information System (INIS)

    A series of critical experiments with water-moderated square-pitched lattices with low-enriched uranium fuel rods was conducted at the IPEN/MB-01 research reactor facility, in 2005. Later, this data become some benchmarks. In one of these experiments the west face of the reactor core was covered with a set of thin SS-304 plates to simulate a heavy reflector as used in the EPR reactor (LEU-COMP-HERM-043). The plates are 3 mm thick and their width and axial length were large enough to cover one whole side of the active core of the reactor. The critical configurations were found as a function of the number of plates. Fuel rods containing UO2 with uranium enriched to 4.3% 235U were arranged in specific geometric configurations to be as close as possible to the critical state. In this work, these benchmark configurations with heavy reflectors were modeled using the Serpent Monte Carlo Code. Serpent uses a universe-based geometry model, which allows the description of practically any three-dimensional fuel or reactor configuration. Neutron transport is based on a combination of surface-to-surface ray-tracing and the Woodcock delta-tracking method. Woodcock method is many times faster than ray-tracing, so compared to MCNP code, Serpent code can bring huge gains in processing time of reactor calculations and reaction rate calculations. The results of these calculations were compared with experimental data and calculations with codes MCNP5 and SCALE6 (KENO-VI) using ENDF/B-VII.0 as cross-section input data. The codes performances are compared in terms of CPU calculation time and agreement with experimental data. Additional y, sensitivity on keff of Serpent woodcock threshold parameter was analyzed. (author)

  10. 2D Resistive Magnetohydrodynamics Calculations with an Arbitrary Lagrange Eulerian Code

    Science.gov (United States)

    Rousculp, C. L.; Gianakon, T. A.; Lipnikov, K. N.; Nelson, E. M.

    2015-11-01

    Single fluid resistive MHD is useful for modeling Z-pinch configurations in cylindrical geometry. One such example is thin walled liners for shock physics or HEDP experiments driven by capacitor banks such as the LANL's PHELIX or Sandia-Z. MHD is also useful for modeling high-explosive-driven flux compression generators (FCGs) and their high-current switches. The resistive MHD in our arbitrary Lagrange Eulerian (ALE) code operates in one and two dimensions in both Cartesian and cylindrical geometry. It is implemented as a time-step split operator, which consists of, ideal MHD connected to the explicit hydro momentum and energy equations and a second order mimetic discretization solver for implicit solution of the magnetic diffusion equation. In a staggered grid scheme, a single-component of cell-centered magnetic flux is conserved in the Lagrangian frame exactly, while magnetic forces are accumulated at the nodes. Total energy is conserved to round off. Total flux is conserved under the ALE relaxation and remap. The diffusion solver consistently computes Ohmic heating. Both Neumann and Dirichlet boundary conditions are available with coupling to external circuit models. Example calculations will be shown.

  11. Blind Calculation of RD-14M Small Break LOCA Tests by CATHENA Code

    International Nuclear Information System (INIS)

    KAERI participated with the computer code CATHENA, which is used to analyze Pressurized Heavy Water Reactors (PHWRs), in an IAEA International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate thermal-hydraulic computer code against qualified data for Small Break Loss of Coolant Accident (SBLOCA) scenario generated on RD-14M Test Facility. Two specific SBLOCA tests selected for this ICSP titled 'Comparison of HWR Code Predictions with SBLOCA Experimental Data', are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow down. This report presents the blind calculation results for these tests conducted by CATHENA code before the test data are distributed to participants. For B9006 test, CATHENA code simulated all the phases of the transient such as blowdown, high-pressure ECI, secondary pressure ramp, refill, switch from high pressure ECI to low pressure ECI, exponential pump ramp, and natural circulation. For B9802 test, CATHENA calculation was intended to predict temperature rise of the FES sheath due to channel boiling, and power supply trip on high FES sheath temperature (600 .deg. C) process protection trip

  12. Linear calculations of edge current driven kink modes with BOUT++ code

    Science.gov (United States)

    Li, G. Q.; Xu, X. Q.; Snyder, P. B.; Turnbull, A. D.; Xia, T. Y.; Ma, C. H.; Xi, P. W.

    2014-10-01

    This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density.

  13. Linear calculations of edge current driven kink modes with BOUT++ code

    Energy Technology Data Exchange (ETDEWEB)

    Li, G. Q., E-mail: ligq@ipp.ac.cn; Xia, T. Y. [Institute of Plasma Physics, CAS, Hefei, Anhui 230031 (China); Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Xu, X. Q. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Snyder, P. B.; Turnbull, A. D. [General Atomics, San Diego, California 92186 (United States); Ma, C. H.; Xi, P. W. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); FSC, School of Physics, Peking University, Beijing 100871 (China)

    2014-10-15

    This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density.

  14. Development of an effective delayed neutron fraction calculation code, BETA-K

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taek Kyum; Song, Hoon; Kim, Young Il; Kim, Young In; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-08-01

    BETA-K, an effective delayed neutron fraction calculation code consistent with Nodal Expansion Method (NEM), has been developed. By using relevant output files of DIF3D code, it can calculate the effective delayed neutron fraction({beta}{sub eff}), neutron lifetime(l{sub eff}), fission spectrum ({chi}-bar) and fission yield data({nu}) for each fissionable isotope, composition of fuels and over the whole core. BETA-K code has been validated by comparing the calculated values to the measured ones of effective delayed neutron fraction in two critical experiments, BFS73-1 and BFS55-1. BFS73-1 is a metal uranium core and BFS55-1 is a metal plutonium core. The C/E values, 1.007 and 0.992 for BFS73-1 and BFS55-1 respectively, agreed well with the experimental values within the experiment errors. BETA-K code predicts 0.00709 and 0.356 {mu}sec as the effective delayed neutron fraction and neutron life time for the uranium metallic fueled equilibrium core of 150MWe KALIMER. (author). 9 refs., 6 figs., 12 tabs.

  15. Linear calculations of edge current driven kink modes with BOUT++ code

    International Nuclear Information System (INIS)

    This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density

  16. The spectral code Apollo2: from lattice to 2D core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I. [CEA Saclay (DEN/DANS/SERMA), 91 - Gif-sur-Yvette (France); Santamarina, A. [CEA Cadarache (CEA/DEN/DER/SPRC), 13 - Saint Paul lez Durance (France)

    2005-07-01

    Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations.

  17. MITRA an advanced code to calculate radionuclide release from nuclear fuels under general irradiation conditions

    International Nuclear Information System (INIS)

    The paper presents the computer code Mitra (Multicomponent isotope transport) which has been constructed to calculate the release of radioactive fission products from nuclear fuels under non-stationary conditions. The code is based on a new integration method fo the mass transport equation in the presence of precipitation, re-solution and radioactive decay. The starting equations and the assumed physical models are briefly described in the main part of the report. A very detailed description of the formulae used and of the Mitra subprograms are presented in extended appendices

  18. A computer code for calculations in the algebraic collective model of the atomic nucleus

    Science.gov (United States)

    Welsh, T. A.; Rowe, D. J.

    2016-03-01

    A Maple code is presented for algebraic collective model (ACM) calculations. The ACM is an algebraic version of the Bohr model of the atomic nucleus, in which all required matrix elements are derived by exploiting the model's SU(1 , 1) × SO(5) dynamical group. This paper reviews the mathematical formulation of the ACM, and serves as a manual for the code. The code enables a wide range of model Hamiltonians to be analysed. This range includes essentially all Hamiltonians that are rational functions of the model's quadrupole moments qˆM and are at most quadratic in the corresponding conjugate momenta πˆN (- 2 ≤ M , N ≤ 2). The code makes use of expressions for matrix elements derived elsewhere and newly derived matrix elements of the operators [ π ˆ ⊗ q ˆ ⊗ π ˆ ] 0 and [ π ˆ ⊗ π ˆ ] LM. The code is made efficient by use of an analytical expression for the needed SO(5)-reduced matrix elements, and use of SO(5) ⊃ SO(3) Clebsch-Gordan coefficients obtained from precomputed data files provided with the code.

  19. Calculation of Gamma-ray Responses for HPGe Detectors with TRIPOLI-4 Monte Carlo Code

    Science.gov (United States)

    Lee, Yi-Kang; Garg, Ruchi

    2014-06-01

    The gamma-ray response calculation of HPGe (High Purity Germanium) detector is one of the most important topics of the Monte Carlo transport codes for nuclear instrumentation applications. In this study the new options of TRIPOLI-4 Monte Carlo transport code for gamma-ray spectrometry were investigated. Recent improvements include the gamma-rays modeling of the electron-position annihilation, the low energy electron transport modeling, and the low energy characteristic X-ray production. The impact of these improvements on the detector efficiency of the gamma-ray spectrometry calculations was verified. Four models of HPGe detectors and sample sources were studied. The germanium crystal, the dead layer of the crystal, the central hole, the beryllium window, and the metal housing are the essential parts in detector modeling. A point source, a disc source, and a cylindrical extended source containing a liquid radioactive solution were used to study the TRIPOLI-4 calculations for the gamma-ray energy deposition and the gamma-ray self-shielding. The calculations of full-energy-peak and total detector efficiencies for different sample-detector geometries were performed. Using TRIPOLI-4 code, different gamma-ray energies were applied in order to establish the efficiency curves of the HPGe gamma-ray detectors.

  20. Emergency Doses (ED) - Revision 3: A calculator code for environmental dose computations

    International Nuclear Information System (INIS)

    The calculator program ED (Emergency Doses) was developed from several HP-41CV calculator programs documented in the report Seven Health Physics Calculator Programs for the HP-41CV, RHO-HS-ST-5P (Rittman 1984). The program was developed to enable estimates of offsite impacts more rapidly and reliably than was possible with the software available for emergency response at that time. The ED - Revision 3, documented in this report, revises the inhalation dose model to match that of ICRP 30, and adds the simple estimates for air concentration downwind from a chemical release. In addition, the method for calculating the Pasquill dispersion parameters was revised to match the GENII code within the limitations of a hand-held calculator (e.g., plume rise and building wake effects are not included). The summary report generator for printed output, which had been present in the code from the original version, was eliminated in Revision 3 to make room for the dispersion model, the chemical release portion, and the methods of looping back to an input menu until there is no further no change. This program runs on the Hewlett-Packard programmable calculators known as the HP-41CV and the HP-41CX. The documentation for ED - Revision 3 includes a guide for users, sample problems, detailed verification tests and results, model descriptions, code description (with program listing), and independent peer review. This software is intended to be used by individuals with some training in the use of air transport models. There are some user inputs that require intelligent application of the model to the actual conditions of the accident. The results calculated using ED - Revision 3 are only correct to the extent allowed by the mathematical models. 9 refs., 36 tabs

  1. Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh

    International Nuclear Information System (INIS)

    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)

  2. Wall-touching kink mode calculations with the M3D code

    International Nuclear Information System (INIS)

    This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicated by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall

  3. GOBLIN computer code. Comparison between calculations and TLTA small break test

    International Nuclear Information System (INIS)

    GOBLIN calcuations have been performed for two simulation tests of the boiling water reactor (BWR) small break loss-of-coolant accidents (LOCAs) which were conducted in the two loop test apparatus (TLTA). The first test investigated the small break with nondegraded emergency core coolant (ECC) systems and the second test studied the same small break but with degraded ECC systems in which the high pressure core spray (HPCS) was assumed unavailable. Very good agreement between test data and calculations is achieved. The second test is the most challenging from code comparison point of view and the code prediction of the complicated mass distribution pattern which changes with time is very satisfactory. In the first test and to some extent late in the second test multidimensional subchannel effects are evident in the core bundle region. These are not and cannot be reproduced by the code since the bundle model of GOBLIN is strictly one-dimensional. (Author)

  4. Wall-touching kink mode calculations with the M3D code

    Energy Technology Data Exchange (ETDEWEB)

    Breslau, J. A., E-mail: jbreslau@pppl.gov; Bhattacharjee, A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08542 (United States)

    2015-06-15

    This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicated by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.

  5. CHARADE: A characteristic code for calculating rate-dependent shock-wave response

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.N.; Tonks, D.L.

    1991-01-01

    In this report we apply spatially one-dimensional methods and simple shock-tracking techniques to the solution of rate-dependent material response under flat-plate-impact conditions. This method of solution eliminates potential confusion of material dissipation with artificial dissipative effects inherent in finite-difference codes, and thus lends itself to accurate calculation of elastic-plastic deformation, shock-to-detonation transition in solid explosives, and shock-induced structural phase transformation. Equations are presented for rate-dependent thermoelastic-plastic deformation for (100) planar shock-wave propagation in materials of cubic symmetry (or higher). Specific numerical calculations are presented for polycrystalline copper using the mechanical threshold stress model of Follansbee and Kocks with transition to dislocation drag. A listing of the CHARADE (for characteristic rate dependence) code and sample input deck are given. 26 refs., 11 figs.

  6. Comprehensive nuclear model calculations: Introduction to the theory and use of the GNASH code

    International Nuclear Information System (INIS)

    A user's manual describing the theory and operation of the GNASH nuclear reaction computer code is presented. This work is based on a series of lectures describing the statistical Hauser-Feshbach plus preequilibrium version of the code with full angular momentum conservation. This version is expected to be most applicable for incident particle energies between 1 key and 50 MeV. General features of the code, the nuclear models that are utilized, input parameters needed to perform calculations, and the output quantities from typical problems are described in detail. The computational structure of the code and the subroutines and functions that are called are summarized as well. Two detailed examples are considered: 14-MeV neutrons incident on 93Nb and 12-MeV neutrons incident on 238U. The former example illustrates a typical calculation aimed at determining neutron, proton, and alpha emission spectra from 14-MeV reactions, and the latter example demonstrates use of the fission model in GNASH

  7. Calculation of the Novovoronezh Recriticality Experiment with the KARATE-440 code system

    Energy Technology Data Exchange (ETDEWEB)

    Hegyi, György, E-mail: ghegyi@aeki.kfki.hu [MTA KFKI Atomic Energy Research Institute, Budapest (Hungary)

    2011-07-01

    In this paper the results of KARATE-440 calculations on Novovoronezh NPP Recriticality Experiment are presented, the corresponding parameters are analyzed. The simulation of the processes and the comparison of the results with the measurements are of particular interest as these efforts make our code to be validated in a higher level. The KARATE-440 code system has been developed and applied for VVER-440 core analysis during near twenty years, as a close collaboration among the developers and the specialists at the 4 Hungarian nuclear power units. KARATE is now a mature, demonstrated, complete and integrated system of computer codes and procedures that provide full and independent VVER core analysis capabilities. Even if only some well defined states of the experiment were simulated, satisfactory agreement was found between measured and calculated data. The results present evidence that the KARATE- 440 code package can adequately model the reactor states in a wide range of performance parameters and the special core type referred in the experiment so it is acceptable for neutronic analysis of all the VVER-440 NPP's. (author)

  8. A Calculation of the radioactivity induced in PWR cluster control rods with the origin and casmo codes

    International Nuclear Information System (INIS)

    The radioactivity induced in PWR cluster control rods during reactor operation has been calculated using the computer programme ORIGEN. Neutron fluxes and spectrum conditions as well as the strongly shielded cross sections for the absorber materials Ag, In and Cd have been obtained by running the cell and assembly code CASMO for a couple of typical cases. The results show that Ag-110m, Fe-55 and Co-60 give the largest activity contributions in the interval 1-10 years after the end of irradiation, and Ni-63 and Cd-113m in a longer time perspective. (author)

  9. The FLUKA Monte Carlo code coupled with the local effect model for biological calculations in carbon ion therapy

    CERN Document Server

    Mairani, A; Kraemer, M; Sommerer, F; Parodi, K; Scholz, M; Cerutti, F; Ferrari, A; Fasso, A

    2010-01-01

    Clinical Monte Carlo (MC) calculations for carbon ion therapy have to provide absorbed and RBE-weighted dose. The latter is defined as the product of the dose and the relative biological effectiveness (RBE). At the GSI Helmholtzzentrum fur Schwerionenforschung as well as at the Heidelberg Ion Therapy Center (HIT), the RBE values are calculated according to the local effect model (LEM). In this paper, we describe the approach followed for coupling the FLUKA MC code with the LEM and its application to dose and RBE-weighted dose calculations for a superimposition of two opposed C-12 ion fields as applied in therapeutic irradiations. The obtained results are compared with the available experimental data of CHO (Chinese hamster ovary) cell survival and the outcomes of the GSI analytical treatment planning code TRiP98. Some discrepancies have been observed between the analytical and MC calculations of absorbed physical dose profiles, which can be explained by the differences between the laterally integrated depth-d...

  10. MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA

    International Nuclear Information System (INIS)

    MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations. A CD-ROM is attached as an appendix. (J.P.N.)

  11. Calculation of effective delayed neutron fraction with modified library of Monte Carlo code

    International Nuclear Information System (INIS)

    Highlights: ► We propose a new Monte Carlo method to calculate the effective delayed neutron fraction by changing the library. ► We study the stability of our method. When the particles and cycles are sufficiently great, the stability is very good. ► The final result is determined to make the deviation least. ► We verify our method on several benchmarks, and the results are very good. - Abstract: A new Monte Carlo method is proposed to calculate the effective delayed neutron fraction βeff. Based on perturbation theory, βeff is calculated with modified library of Monte Carlo code. To verify the proposed method, calculations are performed on several benchmarks. The error of the method is analyzed and the way to reduce error is proposed. The results are in good agreement with the reference data

  12. BETHSY 6.2TC test calculation with TRACE and RELAP5 computer code

    International Nuclear Information System (INIS)

    The TRACE code is still under development and it will have all capabilities of RELAP5. The purpose of the present study was therefore to assess the accuracy of the TRACE calculation of BETHSY 6.2TC test, which is 15.24 cm equivalent diameter horizontal cold leg break. For calculations the TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used. The overall results obtained with TRACE were similar to the results obtained by RELAP5/MOD3.3. The results show that the discrepancies were reasonable. (author)

  13. First vapor explosion calculations performed with MC3D thermal-hydraulic code

    Energy Technology Data Exchange (ETDEWEB)

    Brayer, C.; Berthoud, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires

    1998-01-01

    This paper presents the first calculations performed with the `explosion` module of the multiphase computer code MC3D, which is devoted to the fine fragmentation and explosion phase of a fuel coolant interaction. A complete description of the physical laws included in this module is given. The fragmentation models, taking into account two fragmentation mechanisms, a thermal one and an hydrodynamic one, are also developed here. Results to some calculations to test the numerical behavior of MC3D and to test the explosion models in 1D or 2D are also presented. (author)

  14. Development of nuclear decay data library JDDL, and nuclear generation and decay calculation code COMRAD

    International Nuclear Information System (INIS)

    For safety evaluation of nuclear fuel facilities, a nuclear decay data library named JDDL and a computer code COMRAD have been developed to calculate isotopic composition of each nuclide, radiation source intensity, energy spectrum of γ-ray and neutron, and decay heat of spent fuel. JDDL has been produced mainly from the evaluated nuclear data file ENSDF to use new nuclear data. To supplement the data file for short life nuclides, the JNDC data set were also used which had been evaluated by Japan Nuclear Data Committee. Using these data, calculations became possible from short period to long period after irradiation. (author)

  15. Accuracy evaluation of pin exposure calculations in current LWR core design codes

    International Nuclear Information System (INIS)

    The deregulated utility environment and better utilization of fuel assemblies in nuclear power plants has allowed designers to burn fuel assemblies to maximum allowable exposures. Any uncertainties, associated with the technical approach and numerical methods used to perform pin exposure calculations may cause either peak power exposure to exceed the Nuclear Regulatory Commission (NRC) exposure limit or lead to excessive conservatism and thus inefficient fuel utilization. In this work, a Monte Carlo based coupled depletion code (MCNP5/ORIGEN-S) is utilized to provide reference solutions in order to assess the accuracy of pin power and pin exposure reconstruction methods in the current commercial and licensed three-dimensional (3D) nodal Light Water Reactor (LWR) core design codes. The developed at Pennsylvania State University (PSU) MCNP5/ORIGEN-S coupled depletion code system was validated using measured data from the PSU TRIGA research reactor critical experiments. A number of test cases (modeling benchmarks) representative of LWRs were developed starting from the least complex model towards more complicated and more realistic models. The accuracy evaluation of the pin reconstruction methods was performed by using the CASMO-4 and SIMULATE-3 codes as the representative of current commercial LWR core design systems. Two-dimensional (2D) transport calculations with the TRITON module from the SCALE5 package were employed to produce the spectrum averaged cross-section libraries as a function of burnup for ORIGEN-S calculations. The burnup dependent cross-section libraries are specifically generated for each lattice configuration type. For the MCNP5 calculations continuous cross-section libraries for different isotopes at hot operating temperatures are generated and subsequently utilized. Realistic lattice configurations of the GE13 BWR fuel assemblies (unrodded and rodded) depleted under operating conditions were studied in this research because of their heterogeneous

  16. A New, Efficient Stellar Evolution Code for Calculating Complete Evolutionary Tracks

    CERN Document Server

    Kovetz, Attay; Prialnik, Dina

    2008-01-01

    We present a new stellar evolution code and a set of results, demonstrating its capability at calculating full evolutionary tracks for a wide range of masses and metallicities. The code is fast and efficient, and is capable of following through all evolutionary phases, without interruption or human intervention. It is meant to be used also in the context of modeling the evolution of dense stellar systems, for performing live calculations for both normal star models and merger-products. The code is based on a fully implicit, adaptive-grid numerical scheme that solves simultaneously for structure, mesh and chemical composition. Full details are given for the treatment of convection, equation of state, opacity, nuclear reactions and mass loss. Results of evolutionary calculations are shown for a solar model that matches the characteristics of the present sun to an accuracy of better than 1%; a $1 \\Msun$ model for a wide range of metallicities; a series of models of stellar populations I and II, for the mass rang...

  17. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  18. Presentation of the D.A.R.C. code. Which future for a university made calculation code?; Presentation du code D.A.R.C. Quel avenir pour un code de calcul universitaire?

    Energy Technology Data Exchange (ETDEWEB)

    Mavroulakis, A.; Trombe, A. [INSA - Genie Civl, Laboratoire d`Etudes Thermiques et Mecaniques, 31 - Toulouse (France)

    1996-12-31

    This paper presents the main processes which allow to determine and to take into account in terms of form factors, a scene seen from an emitter and projected onto a receiver. The elements that compose the emitter have a triangular shape while no subdivision is made on the receiver. The analytical method used for the calculation of the form factors of one element in front of a polygonal receiver is briefly presented. Two cell configurations are presented, the second one having not convex facets with no prerequisite subdivision. The sums of form factors from one given emitter are less than 0.01 away from the unit value. For each configuration, the influence of obstacles is encoded as change rates of individual form factors. Finally, in order to illustrate the interest of these form factor calculations, an example of computerized simulation applied to a complex cavity is presented. (J.S.) 6 refs.

  19. HOTB: High precision parallel code for calculation of four-particle harmonic oscillator transformation brackets

    Science.gov (United States)

    Stepšys, A.; Mickevicius, S.; Germanas, D.; Kalinauskas, R. K.

    2014-11-01

    This new version of the HOTB program for calculation of the three and four particle harmonic oscillator transformation brackets provides some enhancements and corrections to the earlier version (Germanas et al., 2010) [1]. In particular, new version allows calculations of harmonic oscillator transformation brackets be performed in parallel using MPI parallel communication standard. Moreover, higher precision of intermediate calculations using GNU Quadruple Precision and arbitrary precision library FMLib [2] is done. A package of Fortran code is presented. Calculation time of large matrices can be significantly reduced using effective parallel code. Use of Higher Precision methods in intermediate calculations increases the stability of algorithms and extends the validity of used algorithms for larger input values. Catalogue identifier: AEFQ_v4_0 Program summary URL: http://cpc.cs.qub.ac.uk/summaries/AEFQ_v4_0.html Program obtainable from: CPC Program Library, Queen’s University of Belfast, N. Ireland Licensing provisions: GNU General Public License, version 3 Number of lines in programs, including test data, etc.: 1711 Number of bytes in distributed programs, including test data, etc.: 11667 Distribution format: tar.gz Program language used: FORTRAN 90 with MPI extensions for parallelism Computer: Any computer with FORTRAN 90 compiler Operating system: Windows, Linux, FreeBSD, True64 Unix Has the code been vectorized of parallelized?: Yes, parallelism using MPI extensions. Number of CPUs used: up to 999 RAM(per CPU core): Depending on allocated binomial and trinomial matrices and use of precision; at least 500 MB Catalogue identifier of previous version: AEFQ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 181, Issue 2, (2010) 420-425 Does the new version supersede the previous version? Yes Nature of problem: Calculation of matrices of three-particle harmonic oscillator brackets (3HOB) and four-particle harmonic oscillator brackets (4HOB) in a more

  20. Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code

    International Nuclear Information System (INIS)

    A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were transported up to approximately 1 m before the region for benchmark calculation. Finally, the energy spectra of neutrons behind the very thick shield were calculated down to the thermal energy with good statistics, and typically agree well within a factor of two with the experimental data over a broad energy range. The 12C(n,2n)11C reaction rates behind the bulk shield were also calculated, which agree with the experimental data typically within 60%. These results are quite impressive in calculation accuracy for deep-penetration problem. In this report, the calculation conditions, geometry and the variance reduction techniques used in the deep-penetration calculation with the MARS14 code are clarified, and several subroutines of MARS14 which were used in our calculation are also given in the appendix. The numerical data of the calculated neutron energy spectra, reaction rates, dose rates and their C/E (Calculation/Experiment) values are also summarized. The

  1. A Comparative Study of the Code Calculations for Local Flow Blockages in the KALIMER-150 Core

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Won Pyo; Ha, Ki Suk; Lee, Yong Bum

    2009-09-15

    A sub-channel blockage may be caused by ingression of damaged fuel debris or foreign obstacles into a core fuel subassembly for a liquid metal reactor(LMR) due to its geometrical compactness of the core design. Local coolant temperature could rise during the incident and it might eventually lead to the degradation of the fuel rods. An analysis computer code is obviously needed not only to assure the safe design of the core, but also to design an effective monitoring system to prevent it from propagating to a serious consequence. The code, therefore, must be capable of representing the thermal-hydraulic phenomena anticipated during the incident reasonably enough to be used to evaluate fuel rod intactness. Most of the technically leading countries for LMR have developed and are using the codes for sub-channel blockage analyses, Korea couldn't afford the resources sponsoring the develop of such a code past years. It was realized later that such an analysis code would ultimately be a prerequisite in the future licensing process of KALIMER as its conceptual design was being elaborated. Since those advanced countries had been reluctant to transfer such the codes, MATRA-LMR/FB had to be developed independently in Korea. It is a revised version of the existing MATRA-LMR code which was aimed for the core sub-channel analysis of LMRs. Some of its models have been improved so appropriately to be able to analyze the sub-channel blockages. Nevertheless, an experiment relevant to the sub-channel blockages had never been conducted for investigating the phenomenon in Korea. Further more, very few experimental data are available on published papers or reports world wide. Under this circumstance, a study has been made as an effort to evaluate the prediction capability of the MATRA-LMR/FB code by comparing the calculation results of the SABRE code, which had already been applied to EFR design. In result, a discrepancy has been observed in a case, but an overall agreement has

  2. Program POD. A computer code to calculate cross sections for neutron-induced nuclear reactions

    International Nuclear Information System (INIS)

    A computer code, POD, was developed for neutron-induced nuclear data evaluations. This program is based on four theoretical models, (1) the optical model to calculate shape-elastic scattering and reaction cross sections, (2) the distorted wave Born approximation to calculate neutron inelastic scattering cross sections, (3) the preequilibrium model, and (4) the multi-step statistical model. With this program, cross sections can be calculated for reactions (n, γ), (n, n'), (n, p), (n, α), (n, d), (n, t), (n, 3He), (n, 2n), (n, np), (n, nα), (n, nd), and (n, 3n) in the neutron energy range above the resonance region to 20 MeV. The computational methods and input parameters are explained in this report, with sample inputs and outputs. (author)

  3. Development of capsule design support subprograms for 3-dimensional temperature calculation using FEM Code NISA

    Energy Technology Data Exchange (ETDEWEB)

    Tobita, Masahiro; Matsui, Yoshinori [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Prediction of irradiation temperature is one of the important issues in the design of the capsule for irradiation test. Many kinds of capsules with complex structure have been designed for recent irradiation requests, and three-dimensional (3D) temperature calculation becomes inevitable for the evaluation of irradiation temperature. For such 3D calculation, however, many works are usually needed for input data preparation, and a lot of time and resources are necessary for parametric studies in the design. To improve such situation, JAERI introduced 3D-FEM (finite element method) code NISA (Numerically Integrated elements for System Analysis) and developed several subprograms, which enabled to support input preparation works in the capsule design. The 3D temperature calculation of the capsule are able to carried out in much easier way by the help of the subprograms, and specific features in the irradiation tests such as non-uniform gamma heating in the capsule, becomes to be considered. (author)

  4. Calculations to an IAHR-benchmark test using the CFD-code CFX-4

    Energy Technology Data Exchange (ETDEWEB)

    Krepper, E.

    1998-10-01

    The calculation concerns a test, which was defined as a benchmark for 3-D codes by the working group of advanced nuclear reactor types of IAHR (International Association of Hydraulic Research). The test is well documented and detailed measuring results are available. The test aims at the investigation of phenomena, which are important for heat removal at natural circulation conditions in a nuclear reactor. The task for the calculation was the modelling of the forced flow field of a single phase incompressible fluid with consideration of heat transfer and influence of gravity. These phenomena are typical also for other industrial processes. The importance of correct modelling of these phenomena also for other applications is a motivation for performing these calculations. (orig.)

  5. Recent R and D around the Monte-Carlo code Tripoli-4 for criticality calculation

    Energy Technology Data Exchange (ETDEWEB)

    Hugot, F.X.; Lee, Y.K.; Malvagi, F. [CEA - DEN/DANS/DM2S/SERMA/LTSD, Saclay (France)

    2008-07-01

    TRIPOLI-4 [1] is the fourth generation of the TRIPOLI family of Monte Carlo codes developed from the 60's by CEA. It simulates the 3D transport of neutrons, photons, electrons and positrons as well as coupled neutron-photon propagation and electron-photons cascade showers. The code addresses radiation protection and shielding problems, as well as criticality and reactor physics problems through both critical and subcritical neutronics calculations. It uses full pointwise as well as multigroup cross-sections. The code has been validated through several hundred benchmarks as well as measurement campaigns. It is used as a reference tool by CEA as well as its industrial and institutional partners, and in the NURESIM [2] European project. Section 2 reviews its main features, with emphasis on the latest developments. Section 3 presents some recent R and D for criticality calculations. Fission matrix, Eigen-values and eigenvectors computations will be exposed. Corrections on the standard deviation estimator in the case of correlations between generation steps will be detailed. Section 4 presents some preliminary results obtained by the new mesh tally feature. The last section presents the interest of using XML format output files. (authors)

  6. CEQCSY: a new code for chemical equilibrium calculation in multiphased systems

    International Nuclear Information System (INIS)

    As part of the CEC Chemval/mirage project, a method is presented for calculating the thermodynamic equilibrium state of a multiphase system, by minimizing its Gibbs free energy constrained by mass balances. Compared to the other algorithms available in the literature, the method has three main characteristics: - the sets of equations corresponding to the conditions of homogeneous and heterogeneous equilibria are simultaneously solved, - a mathematical criterion for bringing a new multicomponent phase in the system is rigorously demontrated. - It enables a detailed representation of the multisite solid solutions with constraints called site closure relation. The code CEQCSY (Chemical Equilibrium in Complex SYstem) uses this formalism, and works with the thermodynamic data base from the EQ3/6 code. This choice makes easier several compared tests with EQ6: quartz dissolution in water, water-atmospheric air equilibrium, theoretical re-equilibrium of seawater, hydrothermal alteration of granite including solid solutions. The test results demonstrate the high efficiency and velocity of the code CEQCSY, when working on equilibrium state of multiphase systems. This high velocity was the aim of this work, in order to couple with thermic, hydrodynamic or mechanic codes

  7. One and half dimensional particle in cell Euterpe code description

    International Nuclear Information System (INIS)

    A 1D 1/2 electromagnetic particle-in-cell named EUTERPE is described. Firstly, the main features of this code are reported. Secondly, stability tests are presented. As a conclusion, the present-day applications of this code in the electromagnetic field-plasma interaction is given

  8. Development of burnup calculation function in reactor Monte Carlo code RMC

    International Nuclear Information System (INIS)

    This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua University of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency. (authors)

  9. Numerical modeling of laser tunneling ionization in explicit particle-in-cell codes

    International Nuclear Information System (INIS)

    Methods for the calculation of laser tunneling ionization in explicit particle-in-cell codes used for modeling laser–plasma interactions are compared and validated against theoretical predictions. Improved accuracy is obtained by using the direct current form for the ionization rate. Multi level ionization in a single time step and energy conservation have been considered during the ionization process. The effects of grid resolution and number of macro-particles per cell are examined. Implementation of the ionization algorithm in two different particle-in-cell codes is compared for the case of ionization-based electron injection in a laser–plasma accelerator

  10. DCHAIN-SP 2001: High energy particle induced radioactivity calculation code

    Energy Technology Data Exchange (ETDEWEB)

    Kai, Tetsuya; Maekawa, Fujio; Kasugai, Yoshimi; Takada, Hiroshi; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kosako, Kazuaki [Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan)

    2001-03-01

    For the purpose of contribution to safety design calculations for induced radioactivities in the JAERI/KEK high-intensity proton accelerator project facilities, the DCHAIN-SP which calculates the high energy particle induced radioactivity has been updated to DCHAIN-SP 2001. The following three items were improved: (1) Fission yield data are included to apply the code to experimental facility design for nuclear transmutation of long-lived radioactive waste where fissionable materials are treated. (2) Activation cross section data below 20 MeV are revised. In particular, attentions are paid to cross section data of materials which have close relation to the facilities, i.e., mercury, lead and bismuth, and to tritium production cross sections which are important in terms of safety of the facilities. (3) User-interface for input/output data is sophisticated to perform calculations more efficiently than that in the previous version. Information needed for use of the code is attached in Appendices; the DCHAIN-SP 2001 manual, the procedures of installation and execution of DCHAIN-SP, and sample problems. (author)

  11. Comparison of two numerical modelling codes for hydraulic and transport calculations in the near-field

    Energy Technology Data Exchange (ETDEWEB)

    Kalin, J., E-mail: jan.kalin@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Petkovsek, B., E-mail: borut.petkovsek@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Montarnal, Ph., E-mail: philippe.montarnal@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Genty, A., E-mail: alain.genty@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Deville, E., E-mail: estelle.deville@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Krivic, J., E-mail: jure.krivic@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia); Ratej, J., E-mail: joze.ratej@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.

  12. Preliminary Development of Thermal Power Calculation Code H-Power for a Supercritical Water Reactor

    Directory of Open Access Journals (Sweden)

    Fan Zhang

    2014-01-01

    Full Text Available SCWR (Supercritical Water Reactor is one of the promising Generation IV nuclear systems, which has higher thermal power efficiency than current pressurized water reactor. It is necessary to perform the thermal equilibrium and thermal power calculation for the conceptual design and further monitoring and calibration of the SCWR. One visual software named H-Power was developed to calculate thermal power and its uncertainty of SCWR, in which the advanced IAPWS-IF97 industrial formulation was used to calculate the thermodynamic properties of water and steam. The ISO-5167-4: 2003 standard was incorporated in the code as the basis of orifice plate to compute the flow rate. New heat balance model and uncertainty estimate have also been included in the code. In order to validate H-Power, an assessment was carried out by using data published by US and Qinshan Phase II. The results showed that H-Power was able to estimate the thermal power of SCWR.

  13. ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®

    Science.gov (United States)

    Damian, F.; Brun, E.

    2014-06-01

    ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

  14. DIST: a computer code system for calculation of distribution ratios of solutes in the purex system

    Energy Technology Data Exchange (ETDEWEB)

    Tachimori, Shoichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-05-01

    Purex is a solvent extraction process for reprocessing the spent nuclear fuel using tri n-butylphosphate (TBP). A computer code system DIST has been developed to calculate distribution ratios for the major solutes in the Purex process. The DIST system is composed of database storing experimental distribution data of U(IV), U(VI), Pu(III), Pu(IV), Pu(VI), Np(IV), Np(VI), HNO{sub 3} and HNO{sub 2}: DISTEX and of Zr(IV), Tc(VII): DISTEXFP and calculation programs to calculate distribution ratios of U(IV), U(VI), Pu(III), Pu(IV), Pu(VI), Np(IV), Np(VI), HNO{sub 3} and HNO{sub 2}(DIST1), and Zr(IV), Tc(VII)(DITS2). The DIST1 and DIST2 determine, by the best-fit procedures, the most appropriate values of many parameters put on empirical equations by using the DISTEX data which fulfill the assigned conditions and are applied to calculate distribution ratios of the respective solutes. Approximately 5,000 data were stored in the DISTEX and DISTEXFP. In the present report, the following items are described, 1) specific features of DIST1 and DIST2 codes and the examples of calculation 2) explanation of databases, DISTEX, DISTEXFP and a program DISTIN, which manages the data in the DISTEX and DISTEXFP by functions as input, search, correction and delete. and at the annex, 3) programs of DIST1, DIST2, and figure-drawing programs DIST1G and DIST2G 4) user manual for DISTIN. 5) source programs of DIST1 and DIST2. 6) the experimental data stored in the DISTEX and DISTEXFP. (author). 122 refs.

  15. Coding and traceability for cells, tissues and organs for transplantation.

    Science.gov (United States)

    Strong, D Michael; Shinozaki, Naoshi

    2010-11-01

    Modern transplantation of cells, tissues and organs has been practiced within the last century achieving both life saving and enhancing results. Associated risks have been recognized including infectious disease transmission, malignancy, immune mediated disease and graft failure. This has resulted in establishment of government regulation, professional standard setting and establishment of vigilance and surveillance systems for early detection and prevention and to improve patient safety. The increased transportation of grafts across national boundaries has made traceability difficult and sometimes impossible. Experience during the first Gulf War with mis-identification of blood units coming from multiple countries without standardized coding and labeling has led international organizations to develop standardized nomenclature and coding for blood. Following this example, cell therapy and tissue transplant practitioners have also moved to standardization of coding systems. Establishment of an international coding system has progressed rapidly and implementation for blood has demonstrated multiple advantages. WHO has held two global consultations on human cells and tissues for transplantation, which recognized the global circulation of cells and tissues and growing commercialization and the need for means of coding to identify tissues and cells used in transplantation, are essential for full traceability. There is currently a wide diversity in the identification and coding of tissue and cell products. For tissues, with a few exceptions, product terminology has not been standardized even at the national level. Progress has been made in blood and cell therapies with a slow and steady trend towards implementation of the international code ISBT 128. Across all fields, there are now 3,700 licensed facilities in 66 countries. Efforts are necessary to encourage the introduction of a standardized international coding system for donation identification numbers, such as ISBT

  16. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  17. FOOD: an interactive code to calculate internal radiation doses from contaminated food products

    International Nuclear Information System (INIS)

    An interactive code, FOOD, has been written in BASIC for the UNIVAC 1108 to facilitate calculation of internal radiation doses to man from radionuclides in food products. In the dose model, vegetation may be contaminated by either air or irrigation water containing radionuclides. The model considers two mechanisms for radionuclide contamination of vegetation: direct deposition on leaves and uptake from soil through the root system. The user may select up to 14 food categories with corresponding consumption rates, growing periods and either irrigation rates or atmospheric deposition rates. These foods include various kinds of produce, grains and animal products. At present, doses may be calculated for the skin, total body and five internal organs from 190 radionuclides. Dose summaries can be displayed at the local terminal. Further details on percent contribution to dose by nuclide and by food type are available from an auxiliary high-speed printer. This output also includes estimated radionuclide concentrations in soil, plants and animal products

  18. OPT13B and OPTIM4 - computer codes for optical model calculations

    International Nuclear Information System (INIS)

    OPT13B is a computer code in FORTRAN for optical model calculations with automatic search. A summary of different formulae used for computation is given. Numerical methods are discussed. The 'search' technique followed to obtain the set of optical model parameters which produce best fit to experimental data in a least-square sense is also discussed. Different subroutines of the program are briefly described. Input-output specifications are given in detail. A modified version of OPT13B specifications are given in detail. A modified version of OPT13B is OPTIM4. It can be used for optical model calculations where the form factors of different parts of the optical potential are known point by point. A brief description of the modifications is given. (author)

  19. Opacity calculation for target physics using the ABAKO/RAPCAL code

    Science.gov (United States)

    Mínguez, E.; Florido, R.; Rodríguez, R.; Gil, J. M.; Rubiano, J. G.; Mendoza, M. A.; Suárez, D.; Martel, P.

    2010-01-01

    Radiative properties of hot dense plasmas remain a subject of current interest since they play an important role in inertial confinement fusion (ICF) research, as well as in studies on stellar physics. In particular, the understanding of ICF plasmas requires emissivities and opacities for both hydro-simulations and diagnostics. Nevertheless, the accurate calculation of these properties is still an open question and continuous efforts are being made to develop new models and numerical codes that can facilitate the evaluation of such properties. In this work the set of atomic models ABAKO/RAPCAL is presented, as well as a series of results for carbon and aluminum to show its capability for modeling the population kinetics of plasmas in both LTE and NLTE regimes. Also, the spectroscopic diagnostics of a laser-produced aluminum plasma using ABAKO/RAPCAL is discussed. Additionally, as an interesting application of these codes, fitting analytical formulas for Rosseland and Planck mean opacities for carbon plasmas are reported. These formulas are useful as input data in hydrodynamic simulation of targets where the computation task is so hard that in line computation with sophisticated opacity codes is prohibitive.

  20. SMARTIES: User-friendly codes for fast and accurate calculations of light scattering by spheroids

    Science.gov (United States)

    Somerville, W. R. C.; Auguié, B.; Le Ru, E. C.

    2016-05-01

    We provide a detailed user guide for SMARTIES, a suite of MATLAB codes for the calculation of the optical properties of oblate and prolate spheroidal particles, with comparable capabilities and ease-of-use as Mie theory for spheres. SMARTIES is a MATLAB implementation of an improved T-matrix algorithm for the theoretical modelling of electromagnetic scattering by particles of spheroidal shape. The theory behind the improvements in numerical accuracy and convergence is briefly summarized, with reference to the original publications. Instructions of use, and a detailed description of the code structure, its range of applicability, as well as guidelines for further developments by advanced users are discussed in separate sections of this user guide. The code may be useful to researchers seeking a fast, accurate and reliable tool to simulate the near-field and far-field optical properties of elongated particles, but will also appeal to other developers of light-scattering software seeking a reliable benchmark for non-spherical particles with a challenging aspect ratio and/or refractive index contrast.

  1. Calculation of the CAREM reactor with the HUEMUL-PUMA-THERMIT chain of codes

    International Nuclear Information System (INIS)

    The purpose of the work was the evaluation of the the CAREM 25 reactor core, using a chain of codes (HUEMUL-PUMA-THERMIT) different to the one used in the original design (CONDOR-CITVAP-THERMIT). First, we performed a partial validation of the our codes in lattices similar to CAREM and reproduced a benchmark for simulation of gadolinium burnup. The results were considered satisfactory for this stage of the project. Then, we calculated the core along the normal operating equilibrium cycle and in hot and cold shut-down conditions. The main outcome of our evaluation confirms the general behaviour of the reference calculations except in one important point referring to the cold shut down. In this condition, the failure of one single rod of bank number 13 of the shut down system, leaves the core in a supercritical state at the beginning of the cycle and this anomaly persists during almost a third of the overall cycle. A new design of the core is proposed with minor modifications of the reference one, without introducing new types of fuel elements and keeping the same fuel management scheme. This new core fulfills all the design requirements. (author)

  2. Decay heat experiment and validation of calculation code systems for fusion reactor

    International Nuclear Information System (INIS)

    Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)

  3. Decay heat experiment and validation of calculation code systems for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki

    1999-10-01

    Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)

  4. A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

    International Nuclear Information System (INIS)

    As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

  5. A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

    Energy Technology Data Exchange (ETDEWEB)

    Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)

    2014-08-15

    As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

  6. Calculation of equilibria at elevated temperatures using the MINTEQ geochemical code

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.W.

    1988-12-01

    Coefficients and equations for calculating mineral hydrolysis constants, solubility products and formation constants for 60 minerals and 57 aqueous species in the 13 component thermodynamic system K/sub 2/O-Na/sub 2/O-CaO-MgO-FeO-Al/sub 2/O/sub 3/-SiO/sub 2/-CO/sub 2/-H/sub 2/O-HF-HCl-H/sub 2/S-H/sub 2/SO/sub 4/ are presented in a format suitable for inclusion in the MINTEQ computer code. The temperature functions presented for minerals are based on the MINTEQ data base at 25/degree/C and the integration of analytical heat capacity power functions. This approach ensures that the temperature functions join smoothly with the low-temperature data base. A new subroutine, DEBYE, was added to MINTEQ that is used to calculate the theoretical Debye-Hueckel parameters A and B as a function of temperature. In addition, this subroutine also calculates a universal value of the extended Debye-Hueckel parameter, b/sub i/, as a function of temperature. The coefficients and equations provide the capability to use MINTEQ to more accurately calculate water/rock equilibrium for temperatures of up to 250/degree/C, and in dilute, low-sulfate, near neutral groundwaters to 300/degree/C. 52 refs., 1 fig., 6 tabs.

  7. Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code

    CERN Document Server

    Nunomiya, T; Nakamura, T; Nakao, N

    2002-01-01

    A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were trans...

  8. Three-dimensional whole core transport calculation method and performance of the DeCART code

    International Nuclear Information System (INIS)

    The three-dimensional (3D) transport calculation method implemented in a whole core neutron transport code DeCART is presented and its performance is examined in terms of solution accuracy and execution speed. The 3D flux calculation in DeCART is based on a transverse-integration method in which the radial and axial dependencies are handled separately. The radial dependence is resolved by the elaborated two-dimensional method of characteristics (MOC) whereas the axial dependence is dealt with the simple one-dimensional diffusion model. The global balance of the 3D flux distribution is incorporated by the coarse mesh finite difference (CMFD) formulation. It is shown that the CMFD formulation enables the approximate three-dimensional transport calculation through the transverse-integration, and furthermore it is very effective in achieving rapid convergence. The accuracy of the approximate 3D whole-core transport calculation method is proved by analyzing rodded variations of the C5G7 MOX heterogeneous core benchmark problem for which Monte Carlo solutions are generated as the reference

  9. DEEP code to calculate dose equivalents in human phantom for external photon exposure by Monte Carlo method

    International Nuclear Information System (INIS)

    The present report describes a computer code DEEP which calculates the organ dose equivalents and the effective dose equivalent for external photon exposure by the Monte Carlo method. MORSE-CG, Monte Carlo radiation transport code, is incorporated into the DEEP code to simulate photon transport phenomena in and around a human body. The code treats an anthropomorphic phantom represented by mathematical formulae and user has a choice for the phantom sex: male, female and unisex. The phantom can wear personal dosimeters on it and user can specify their location and dimension. This document includes instruction and sample problem for the code as well as the general description of dose calculation, human phantom and computer code. (author)

  10. SFR whole core burnup calculations with TRIPOLI-4 Monte Carlo code

    International Nuclear Information System (INIS)

    Under the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD/NEA, an international collaboration benchmark was recently established on the neutronic analysis of four Sodium-cooled Fast Reactor (SFR) concepts of the Generation- IV nuclear energy systems. As the whole core Monte Carlo depletion calculation is one of the essential challenges of current reactor physics studies, the continuous-energy TRIPOLI-4 Monte Carlo transport code was firstly used in this study to perform whole core 3D neutronic calculations for these four SFR cores. Two medium size (1000 MWt) and two large size (3600 MWt) SFR of GEN-IV systems were analyzed. The medium size SFR concepts are from the Advanced Burner Reactors (ABR). The large size SFR concepts are from the self-breeding reactors. The TRIPOLI-4 depletion calculations were made with MOX and metallic U-Pu-Zr fuels for the ABR cores and with MOX and Carbide (U,Pu)C fuels for the self-breeding cores. The whole core reactor physics parameters calculations were performed for the BOEC and EOEC (Beginning and End of Equilibrium Cycle) conditions. This paper summarizes the TRIPOLI-4 calculation results of Keff, βeff, sodium void worth, Doppler constant, control rod worth, and core power distributions for the BOEC and EOEC conditions. The one-cycle depletion calculation results of the core inventory of U and TRU (Pu, Am, Cm, and Np) are also analyzed, after 328.5 days depletion irradiation for the ABR cores, 410 days for the large MOX core, and 500 days for the large carbide core. (author)

  11. Transmutation of alloys in MFE facilities as calculated by REAC (a computer code system for activation and transmutation)

    International Nuclear Information System (INIS)

    A computer code system for fast calculation of activation and transmutation has been developed. The system consists of a driver code, cross-section libraries, flux libraries, a material library, and a decay library. The code is used to predict transmutations in a Ti-modified 316 stainless steel, a commercial ferritic alloy (HT9), and a V-15%Cr-5%Ti alloy in various magnetic fusion energy (MFE) test facilities and conceptual reactors

  12. FREEZE PROFILE AND HEAT BALANCE CALCULATION OF THE 160kA DRAINED CELL

    Institute of Scientific and Technical Information of China (English)

    X.P.Li; J.Li; Y.Q.Lai; H.Q.Zhao; Y.X.Liu

    2004-01-01

    A 2D full cell thermo-electric model of 160kA drained cell was set up using finite element code to calculate its freeze profile,then the drained cell model was modified according to the freeze profile computed and its heat balance was calculated.Compared with that of a 160kA conventional Hall-Heroult cell(H-H cell),though the melts volume of the drained cell reduced greatly,the whole heat loss from it didn't drop down apparently,and an analysis was presented in the paper.On the other hand,the anodecathode distance(ACD)of a drained cell was much less than that of a H-H cell,so the voltage drop on it and heat produced decreased too,steps should be taken to keep a workable heat balance on a drained cell.

  13. Dose and shielding calculation of galactic cosmic ray using FLUKA Mont Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, Hamide B. [Physics Department, University of Qom, Qom (Iran); Raisali, Golamreza; Babazade, Alireza [Radiation Applications Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Tehran (Iran); Feghhi, Amirhosein [Physics and Nuclear Engineering Department, Amirkabir University, Tehran (Iran)

    2009-07-01

    Astronauts' exposure to space radiation is a limiting factor for long-term missions. Therefore shielding is a critical issue in space mission success. In this work the FLUKA Monte Carlo code has been coupled with simple models of the spacecraft and equivalent phantom to calculate skin averaged doses due to exposure to Galactic Cosmic Rays (GCR) beyond various thicknesses of aluminium and polyethylene shields. Simulations have been performed for the most abundant elements including H, He, C and Fe ions. The spectra of these ions have been taken from Badhwar-O'Neill's model, and LET distribution of the ions and electrons calculated using SRIM and ESTAR computer programs, respectively. It has been observed that GCR absorbed dose behind the shields remained approximately constant with increasing shield thicknesses, but dose equivalent shows a slight decrease. It is also found that although polyethylene is a more effective GCR shield than aluminum as indicated in the results of similar investigations, but the practical thicknesses of polyethylene are still insufficient to shield high energy GCR ions encountered in long-term space missions.

  14. Calculational results using a survey type code system for the analysis of the Almaraz Unit 2 PWR benchmark

    International Nuclear Information System (INIS)

    The calculations performed for the Almaraz Unit 2 PWR using the code packages of the Atomic Energy Corporation of South Africa Ltd. are summarized. These calculations were done as part of the IAEA Coordinated Research Programme on In-Core Fuel Management Code Package Validation for LWRs. A brief description of the one-dimensional cross section generation package as well as of the Level II (scoping type) global core calculational package which was used is given. Detailed results are presented in several appendices. 29 figs., 20 tabs., 10 refs

  15. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  16. Development of a multi-group SN transport calculation code with unstructured tetrahedral meshes

    International Nuclear Information System (INIS)

    This paper reviews the computational methods used in the MUST (Multi-group Unstructured geometry SN Transport) code for solving the multi-group Sn transport equation in general geometries and describes the status of development of MUST. MUST solves the multi-group transport equation with unstructured tetrahedral meshes for modeling complicated geometrical problems. For tetrahedral mesh generation, input generation, and output visualization, we developed a management program where the mesh generation is based on Gmsh and TetGen that are open softwares. The geometrical modeling is done with the commercial CAD softwares such as CATIA. MUST uses the discontinuous finite element method (DFEM) and two-sub cell balance methods with linear discontinuous expansion (LDEM-SCB) to spatially discretize the transport equation. We applied MUST to three neutron and gamma coupled test problems for testing MUST. (author)

  17. Adaption, validation and application of advanced codes with 3-dimensional neutron kinetics for accident analysis calculations - STC with Bulgaria

    International Nuclear Information System (INIS)

    In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.)

  18. Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes

    International Nuclear Information System (INIS)

    The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR-R1 reactor are calculated. The idea is to obtain the systematic error for k∞ for this methodology comparing the calculated and the experimental results

  19. Parametric studies of radiolytic oxidation of iodide solutions with and without paint: comparison with code calculations

    Energy Technology Data Exchange (ETDEWEB)

    Poletiko, C.; Hueber, C. [Inst. de Protection et de Surete Nucleaire, C.E. Cadarache, St. Paul-lez-Durance (France); Fabre, B. [CISI, C.E. Cadarache, St. Paul-lez-Durance (France)

    1996-12-01

    In case of severe nuclear accident, radioactive material may be released into the environment. Among the fission products involved, are the very volatile iodine isotopes. However, the chemical forms are not well known due to the presence of different species in the containment with which iodine may rapidly react to form aerosols, molecular iodine, hydroiodic acid and iodo-organics. Tentative explanations of different mechanisms were performed through benchscale tests. A series of tests has been performed at AEA Harwell (GB) to study parameters such as pH, dose rate, concentration, gas flow rate, temperature in relation to molecular iodine production, under dynamic conditions. Another set of tests has been performed in AECL Whiteshell (CA) to study the behaviour of painted coupons, standing in gas phase or liquid phase or both, with iodine compounds under radiation. The purpose of our paper is to synthesize the data and compare the results to the IODE code calculation. Some parameters of the code were studied to fit the experimental result the best. A law, concerning the reverse reaction of iodide radiolytic oxidation, has been proposed versus: pH, concentrations and gas flow-rate. This law does not apply for dose rate variations. For the study of painted coupons, it has been pointed out that molecular iodine tends to be adsorbed or chemically absorbed on the surface in gas phase, but the mechanism should be more sophisticated in the aqueous phase. The iodo-organics present in liquid phase tend to be partly or totally destroyed by oxidation under radiation (depending upon the dose delivered). These points are discussed. (author) 18 figs., 3 tabs., 15 refs.

  20. A computer code to calculate the fast induced signals by electron swarms in gases

    Energy Technology Data Exchange (ETDEWEB)

    Tobias, Carmen C.B. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Mangiarotti, Alessio [Universidade de Coimbra (Portugal). Dept. de Fisica. Lab. de Instrumentacao e Fisica Experimental de Particulas

    2010-07-01

    Full text: The study of electron transport parameters (i.e. drift velocity, diffusion coefficients and first Townsend coefficient) in gases is very important in several areas of applied nuclear science. For example, they are a relevant input to the design of particle detector employing micro-structures (MSGC's, micromegas, GEM's) and RPC's (resistive plate chambers). Moreover, if the data are accurate and complete enough, they can be used to derive a set of electron impact cross-sections with their energy dependence, that are a key ingredient in micro-dosimetry calculations. Despite the fundamental need of such data and the long age of the field, the gases of possible interest are so many and the effort of obtaining good quality data so time demanding, that an important contribution can still be made. As an example, electrons drift velocity at moderate field strengths (up to 50 Td) in pure Isobutane (a tissue equivalent gas) has been measured only recently by the IPEN-LIP collaboration using a dedicated setup. The transport parameters are derived from the recorded electric pulse induced by a swarm started with a pulsed laser shining on the cathode. To aid the data analysis, a special code has been developed to calculate the induced pulse by solving the electrons continuity equation including growth, drift and diffusion. A realistic profile of the initial laser beam is taken into account as well as the boundary conditions at the cathode and anode. The approach is either semi-analytic, based on the expression derived by P. H. Purdie and J. Fletcher, or fully numerical, using a finite difference scheme improved over the one introduced by J. de Urquijo et al. The agreement between the two will be demonstrated under typical conditions for the mentioned experimental setup. A brief discussion on the stability of the finite difference scheme will be given. The new finite difference scheme allows a detailed investigation of the importance of back diffusion to

  1. Calculation of Sodium Fire Test-I (Run-E6) using sodium combustion analysis code ASSCOPS version 2.0

    Energy Technology Data Exchange (ETDEWEB)

    Nakagiri, Toshio; Ohno, Shuji; Miyake, Osamu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-11-01

    The calculation of Sodium Fire Test-I (Run-E6) was performed using the ASSCOPS (Analysis of Simultaneous Sodium Combustions in Pool and Spray) code version 2.0 in order to determine the parameters used in the code for the calculations of sodium combustion behavior of small or medium scale sodium leak, and to validate the applicability of the code. The parameters used in the code were determined and the validation of the code was confirmed because calculated temperatures, calculated oxygen concentration and other calculated values almost agreed with the test results. (author)

  2. Comparative calculations of the WWER fuel rod thermophysical characteristics employing the TOPRA-s and the TRANSURANUS computer codes

    International Nuclear Information System (INIS)

    A short description of the TOPRA-s computer code is presented. The code is developed to calculate the thermophysical cross-section characteristics of the WWER fuel rods: fuel temperature distributions and fuel-to-cladding gap conductance. The TOPRA-s input does not require the fuel rod irradiation pre-history (time dependent distributions of linear power, fast neutron flux and coolant temperature along the rod). The required input consists of the considered cross-section data (coolant temperature, burnup, linear power) and the overall fuel rod data (burnup and linear power). TOPRA-s is included into the KASKAD code package. Some results of the TOPRA-s code validation using the SOFIT-1 and IFA-503.1 experimental data, are shown. A short description of the TRANSURANUS code for thermal and mechanical predictions of the LWR fuel rod behavior at various irradiation conditions and its version for WWER reactors, are presented. (Authors)

  3. OSCAR, a code for the calculation of the yield of radioisotopes produced by charged-particle induced nuclear reactions

    International Nuclear Information System (INIS)

    A computer code OSCAR, operated on a main frame computer was developed for the calculation of the yield of radioisotopes produced by charged-particle induced nuclear reactions. The excitation functions required for calculating the yield were evaluated by means of an empirical rule which we developed on the basis of a systematics derived from a number of experimental data reported in the literature. The rule is valid for light ion (Z ≤ 2)-induced reactions followed by neutron emission processes. Other excitation functions are also obtainable from the data file in OSCAR. In addition, the code possesses functions useful for the calculation of the stopping power and range. The energy loss and the distribution of recoil products in stacked targets are also provided as options. The formalism, structure, and direction for the usage of the code are described together with the explanation of the functions of some routines. (author)

  4. A FORTRAN code for the calculation of probe volume geometry changes in a laser anemometry system caused by window refraction

    Science.gov (United States)

    Owen, Albert K.

    1987-01-01

    A computer code was written which utilizes ray tracing techniques to predict the changes in position and geometry of a laser Doppler velocimeter probe volume resulting from refraction effects. The code predicts the position change, changes in beam crossing angle, and the amount of uncrossing that occur when the beams traverse a region with a changed index of refraction, such as a glass window. The code calculates the changes for flat plate, cylinder, general axisymmetric and general surface windows and is currently operational on a VAX 8600 computer system.

  5. FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

    1981-01-01

    FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

  6. Neptun: an interactive code for calculating doses to man due to radionuclides in acquatic food chains

    International Nuclear Information System (INIS)

    A flexible and interactive code, NEPTUN, has been written in FORTRAN IV for the PDP-10 computer to assess the impact on man of radionuclides in aquatic food chains. NEPTUN is based on an equilibrium model of the linear-chain type, and calculates aquatic food concentrations and doses to man. A decay term is included for the holdup time of the various food types. A total of seven food types can be selected, which include drinking water, freshwater and salt-water plants, inverebrates and fish. Thirty different diets can be implemented and five different dose factor files can be chosen. These include dose conversion factors for infants and adults based on ICRP 2 and ICRP 26 methodologies. All dose factors involve a dose commitment of 50 years, or equivalently, 50 years of chronic exposure. To date, only stochastic ICRP 26 dose caluclations have been implemented. The basic concentration factor file contains data for 211 different radionuclides; the dose factor files are less comprehensive. However, all files can be readily expanded. The output includes tables of concentrations and doses for individual radionuclides, as well as summaries for groups of radionuclides. Existing aquatic food chain models and the sources of currently-used generic concentration factors are briefly reviewed, and dose factors based on ICRP 2 and ICRP 26 methodologies are contrasted. (auth)

  7. Wall touching kink mode calculations with the M3D code

    Science.gov (United States)

    Breslau, J. A.

    2014-10-01

    In recent years there have been a number of results published concerning the transient vessel currents and forces occurring during a tokamak VDE, as predicted by simulations with the nonlinear MHD code M3D. The nature of the simulations is such that these currents and forces occur at the boundary of the computational domain, making the proper choice of boundary conditions critical to the reliability of the results. The M3D boundary condition includes the prescription that the normal component of the velocity vanish at the wall. It has been argued that this prescription invalidates the calculations because it would seem to rule out the possibility of advection of plasma surface currents into the wall. This claim has been tested by applying M3D to an idealized case - a kink-unstable plasma column - in order to abstract the essential physics from the complications involved in the attempt to model real devices. While comparison of the results is complicated by effects arising from the higher dimensionality and complexity of M3D, we have verified that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the ``Hiro'' currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.

  8. Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code

    Energy Technology Data Exchange (ETDEWEB)

    Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)

    2013-09-15

    Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)

  9. SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR

    International Nuclear Information System (INIS)

    SPLOSH is a time dependent, one dimensional, finite difference (in time and space) coupled neutron kinetics and thermal hydraulics code for studying pressurised faults and control transients in water reactor systems. An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the reactor core. A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface. The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly. Circuit features modelled include pumps, spray cooling and the SGHWR steam drum. Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified. (UK)

  10. Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code

    International Nuclear Information System (INIS)

    A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (keff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of keff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of keff from MCNP5 were obtained when the control fuel rods were fully inserted, the keff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370±0.00054. When the control fuel rods were fully withdrawn the value of keff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)

  11. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations

    International Nuclear Information System (INIS)

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor keff (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  12. TRANS4: a computer code calculation of solid fuel penetration of a concrete barrier

    International Nuclear Information System (INIS)

    The computer code, TRANS4, models the melting and penetration of a solid barrier by a solid disc of fuel following a core disruptive accident. This computer code has been used to model fuel debris penetration of basalt, limestone concrete, basaltic concrete, and magnetite concrete. Sensitivity studies were performed to assess the importance of various properties on the rate of penetration. Comparisons were made with results from the GROWS II code

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-15

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

  14. Evaluation of the methodology for dose calculation in microdosimetry with electrons sources using the MCNP5 Code

    International Nuclear Information System (INIS)

    This study made a comparison between some of the major transport codes that employ the Monte Carlo stochastic approach in dosimetric calculations in nuclear medicine. We analyzed in detail the various physical and numerical models used by MCNP5 code in relation with codes like EGS and Penelope. The identification of its potential and limitations for solving microdosimetry problems were highlighted. The condensed history methodology used by MCNP resulted in lower values for energy deposition calculation. This showed a known feature of the condensed stories: its underestimates both the number of collisions along the trajectory of the electron and the number of secondary particles created. The use of transport codes like MCNP and Penelope for micrometer scales received special attention in this work. Class I and class II codes were studied and their main resources were exploited in order to transport electrons, which have particular importance in dosimetry. It is expected that the evaluation of available methodologies mentioned here contribute to a better understanding of the behavior of these codes, especially for this class of problems, common in microdosimetry. (author)

  15. The Plasma Simulation Code: A modern particle-in-cell code with patch-based load-balancing

    Science.gov (United States)

    Germaschewski, Kai; Fox, William; Abbott, Stephen; Ahmadi, Narges; Maynard, Kristofor; Wang, Liang; Ruhl, Hartmut; Bhattacharjee, Amitava

    2016-08-01

    This work describes the Plasma Simulation Code (PSC), an explicit, electromagnetic particle-in-cell code with support for different order particle shape functions. We review the basic components of the particle-in-cell method as well as the computational architecture of the PSC code that allows support for modular algorithms and data structure in the code. We then describe and analyze in detail a distinguishing feature of PSC: patch-based load balancing using space-filling curves which is shown to lead to major efficiency gains over unbalanced methods and a previously used simpler balancing method.

  16. Proceedings of 5. French speaking scientific days on calculation codes for radioprotection, radio-physics and dosimetry

    International Nuclear Information System (INIS)

    These scientific days were organised by the 'technical protection' Section of the French Society of Radiation Protection (SFRP) in cooperation with the French society of medical physicists (SFPM), the Swiss Romandie association of radioprotection (ARRAD) and the associated laboratories of radio-physics and dosimetry (LARD). The objective of these days was to review the existing calculation codes used in radiation transport, source estimation and dose management, and to identify some future prospects. This document brings together the available presentations (slides) together with their corresponding abstracts (in French) and dealing with: 1 - Presentation of the conference days (L. De Carlan); 2 - Simulating radionuclide transfers in the environment: what calculation codes and for what? (C. Mourlon); 3 - Contribution of Monte-Carlo calculation to the theoretical foundation analysis of calibration procedures and dosemeters design for radioprotection photon dosimetry (J.M. Bordy); 4 - Use of calculation codes in R and D for the development of a new passive dosemeter for photons and beta radiations (B. Moreno); 5 - Development of a new virtual sources model for the Monte-Carlo prediction of EPID (Electronic Portal Imaging Device) images and implementation in PENELOPE (I. Chabert); 6 - Prediction of high-resolution EPID images for in-vivo dosimetry (D. Patin); 7 - 4D thorax modeling by artificial neural networks (P.E. Leni); 8 - Presentation of the calculation utilities of the book 'Calculation of ionizing radiations generated doses' (Vivier, Lopez, EDP Sciences 2012) (A. Vivier); 9 - RayXpertC: a 3D modeling and Monte-Carlo dose rate calculation software (C. Dossat); 10 - TRIPOLI-4R Version 9 S Monte-Carlo code for radioprotection (F. Damian); 11 - Realistic radioprotection training with the digital school workshop (E. Courageot); 12 - Use of BEAMNRC code for dental prostheses influence evaluation in ENT cancers treatment by external radiotherapy (C. De Conto); 13

  17. DCHAIN 2: a computer code for calculation of transmutation of nuclides

    International Nuclear Information System (INIS)

    DCHAIN2 is a one-point depletion code which solves the coupled equation of radioactive growth and decay for a large number of nuclides by the Bateman method. A library of nuclear data for 1170 fission products has been prepared for providing input data to this code. The Bateman method surpasses the matrix exponential method in computational accuracies and in saving computer storage for the code. However, most existing computer codes based on the Bateman method have shown serious drawbacks in treating cyclic chains and more than a few specific types of decay chains. The present code has surmounted the above drawbacks by improving the code FP-S, and has the following characteristics: (1) The code can treat any type of transmutation through decays or neutron induced reactions. Multiple decays and reactions are allowed for a nuclide. (2) Unknown decay energy in the nuclear data library can be estimated. (3) The code constructs the decay scheme of each nuclide in the code and breaks it up into linear chains. Nuclide names, decay types and branching ratios of mother nuclides are necessary as the input data for each nuclide. Order of nuclides in the library is arbitrary because each nuclide is destinguished by its nuclide name. (4) The code can treat cyclic chains by an approximation. A library of the nuclear data has been prepared for 1170 fission products, including the data for half-lives, decay schemes, neutron absorption cross sections, fission yields, and disintegration energies. While DCHAIN2 is used to compute the compositions, radioactivity and decay heat of fission products, the gamma-ray spectrum of fission products can be computed also by a separate code FPGAM using the composition obtained from DCHAIN2. (J.P.N.)

  18. LWR-WIMS, a computer code for light water reactor lattice calculations

    International Nuclear Information System (INIS)

    LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)

  19. Electron and ion cyclotron heating calculations in the tandem-mirror modeling code MERTH

    International Nuclear Information System (INIS)

    To better understand and predict tandem-mirror experiments, we are building a comprehensive Mirror Equilibrium Radial Transport and Heating (MERTH) code. In this paper we first describe our method for developing the code. Then we report our plans for the installation of physics packages for electron- and ion-cyclotron heating of the plasma

  20. Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

    Directory of Open Access Journals (Sweden)

    Skrzypek Maciej

    2015-09-01

    Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

  1. Development of a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport

    CERN Document Server

    Jia, Xun; Sempau, Josep; Choi, Dongju; Majumdar, Amitava; Jiang, Steve B

    2009-01-01

    Monte Carlo simulation is the most accurate method for absorbed dose calculations in radiotherapy. Its efficiency still requires improvement for routine clinical applications, especially for online adaptive radiotherapy. In this paper, we report our recent development on a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport. We have implemented the Dose Planning Method (DPM) Monte Carlo dose calculation package (Sempau et al, Phys. Med. Biol., 45(2000)2263-2291) on GPU architecture under CUDA platform. The implementation has been tested with respect to the original sequential DPM code on CPU in two cases. Our results demonstrate the adequate accuracy of the GPU implementation for both electron and photon beams in radiotherapy energy range. A speed up factor of 4.5 and 5.5 times have been observed for electron and photon testing cases, respectively, using an NVIDIA Tesla C1060 GPU card against a 2.27GHz Intel Xeon CPU processor .

  2. The Neutronic And Power Distribution Calculations For Triga 2 MW Reactor Using WIMS-D/4 And Citation Codes

    International Nuclear Information System (INIS)

    . The neutronic calculation has been carried out for TRIGA 2 MW reactor. These included criticality flux and power distributions. Computer code Citation which solves 7-groups, 3-dimensional hexagonal geometry has been used. The multi groups-cross-section is generated by the WIMS-D/4 code.This 7-group-39x39x38-mesh-points problem takes about 90 minutes on the Pentium-133 MHz PC. The calculation of the initial core of TRIGA 2 MW reactor shows that the excess reactivity of the core is 7,8% and the thermal fluxes in the irradiation positions are between 1.0-2.9*1013n cm-2s-1. The results are about 10% deviate from those calculated by General Atomics. In the initial core, the highest power is produced in the C-9 position. The fuel element in this position produces 30.7 k W thermal power

  3. Load-balancing techniques for a parallel electromagnetic particle-in-cell code

    Energy Technology Data Exchange (ETDEWEB)

    PLIMPTON,STEVEN J.; SEIDEL,DAVID B.; PASIK,MICHAEL F.; COATS,REBECCA S.

    2000-01-01

    QUICKSILVER is a 3-d electromagnetic particle-in-cell simulation code developed and used at Sandia to model relativistic charged particle transport. It models the time-response of electromagnetic fields and low-density-plasmas in a self-consistent manner: the fields push the plasma particles and the plasma current modifies the fields. Through an LDRD project a new parallel version of QUICKSILVER was created to enable large-scale plasma simulations to be run on massively-parallel distributed-memory supercomputers with thousands of processors, such as the Intel Tflops and DEC CPlant machines at Sandia. The new parallel code implements nearly all the features of the original serial QUICKSILVER and can be run on any platform which supports the message-passing interface (MPI) standard as well as on single-processor workstations. This report describes basic strategies useful for parallelizing and load-balancing particle-in-cell codes, outlines the parallel algorithms used in this implementation, and provides a summary of the modifications made to QUICKSILVER. It also highlights a series of benchmark simulations which have been run with the new code that illustrate its performance and parallel efficiency. These calculations have up to a billion grid cells and particles and were run on thousands of processors. This report also serves as a user manual for people wishing to run parallel QUICKSILVER.

  4. The MARS15-based FermiCORD code system for calculation of the accelerator-induced residual dose

    CERN Document Server

    Grebe, A; Lu, T; Mokhov, N; Pronskikh, V

    2016-01-01

    The FermiCORD code system, a set of codes based on MARS15 that calculates the accelerator-induced residual doses at experimental facilities of arbitrary configurations, has been developed. FermiCORD is written in C++ as an add-on to Fortran-based MARS15. The FermiCORD algorithm consists of two stages: 1) simulation of residual doses on contact with the surfaces surrounding the studied location and of radionuclide inventories in the structures surrounding those locations using MARS15, and 2) simulation of the emission of the nuclear decay gamma-quanta by the residuals in the activated structures and scoring the prompt doses of these gamma-quanta at arbitrary distances from those structures. The FermiCORD code system has been benchmarked against similar algorithms based on other code systems and showed a good agreement. The code system has been applied for calculation of the residual dose of the target station for the Mu2e experiment and the results have been compared to approximate dosimetric approaches.

  5. A development of NRESPG Monte Carlo code for the calculation of neutron response function for gas counters

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, N. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Kudo, K. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Toyokawa, H. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Torii, T. [Japan Power Reactor and Nuclear Fuel Development Corporation, Tsuruga Office, Fukui 919-12 (Japan); Hashimoto, M. [Japan Power Reactor and Nuclear Fuel Development Corporation, O-arai Engineering Center, Ibaraki 311-13 (Japan); Sugita, T. [Science System Laboratory, Ibaraki 309-17 (Japan); Dietze, G. [Physikalisch-Technische Bundesanstalt, 38023 Braunschweig (Germany); Yang, X. [China Institute of Atomic Energy (China)

    1999-02-11

    A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for {sup 3}He, H{sub 2}, or BF{sub 3} gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of {sup 3}He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the {sup 3}He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF{sub 3} counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations.

  6. Evaluation of PENFAST - A fast Monte Carlo code for dose calculations in photon and electron radiotherapy treatment planning

    Energy Technology Data Exchange (ETDEWEB)

    Habib, B.; Poumarede, B.; Tola, F.; Barthe, J. [CEA, LIST, Dept Technol Capteur et Signal, F-91191 Gif Sur Yvette, (France)

    2010-07-01

    The aim of the present study is to demonstrate the potential of accelerated dose calculations, using the fast Monte Carlo (MC) code referred to as PENFAST, rather than the conventional MC code PENELOPE, without losing accuracy in the computed dose. For this purpose, experimental measurements of dose distributions in homogeneous and inhomogeneous phantoms were compared with simulated results using both PENELOPE and PENFAST. The simulations and experiments were performed using a Saturne 43 linac operated at 12 MV (photons), and at 18 MeV (electrons). Pre-calculated phase space files (PSFs) were used as input data to both the PENELOPE and PENFAST dose simulations. Since depth-dose and dose profile comparisons between simulations and measurements in water were found to be in good agreement (within {+-} 1% to 1 mm), the PSF calculation is considered to have been validated. In addition, measured dose distributions were compared to simulated results in a set of clinically relevant, inhomogeneous phantoms, consisting of lung and bone heterogeneities in a water tank. In general, the PENFAST results agree to within a 1% to 1 mm difference with those produced by PENELOPE, and to within a 2% to 2 mm difference with measured values. Our study thus provides a pre-clinical validation of the PENFAST code. It also demonstrates that PENFAST provides accurate results for both photon and electron beams, equivalent to those obtained with PENELOPE. CPU time comparisons between both MC codes show that PENFAST is generally about 9-21 times faster than PENELOPE. (authors)

  7. A development of NRESPG Monte Carlo code for the calculation of neutron response function for gas counters

    International Nuclear Information System (INIS)

    A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for 3He, H2, or BF3 gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of 3He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the 3He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF3 counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations

  8. PWR neutron ex-vessel detection calculations using three-dimensional codes; Calculs de detection neutronique externe dans un rep

    Energy Technology Data Exchange (ETDEWEB)

    Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)

    1997-10-01

    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.

  9. Comparison of MACCS users calculations for the international comparison exercise on probabilistic accident consequence assessment code, October 1989--June 1993

    Energy Technology Data Exchange (ETDEWEB)

    Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)

    1994-04-01

    Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions.

  10. ExoPlex: A code for calculating interior structure and mineralogy and mass-radius relationships for exoplanets

    Science.gov (United States)

    Desch, Steven; Lorenzo, Alejandro; Ko, Byeongkwan

    2016-06-01

    We present a computer code we have written for general release that calculates the interior structure and mass-radius relationships of solid exoplanets up to a few Earth masses. The basic algorithm is that of Seager et al. (2007), Zeng & Sasselov (2013) and Dorn et al. (2015): the code integrates the 1-D (spherical) equation of hydrostatic equilibrium to find pressure in shells of various depths assuming a gravitational acceleration, uses the bulk modulus of the materials as inputs to an equation of state to convert pressures into density and volume in each shell, recomputes the shell thicknesses and gravitational acceleration, and iterates the solution to convergence. Unlike most existing codes, we do not impose a particular mineralogy in each shell. Instead we adopt the approach of Dorn et al. (2015), in which we impose a stoichiometry in each shell; for rocky shells and the metal core the code calls the PerpleX code (Connolly et al. 2005) to compute the mineralogy and material properties appropriate to that shell’s stoichiometry, pressure and temperature. Unique attributes of the code are as follows. The mineralogy is complete in the Fe-Mg-Si-O system, including species like FeSi and FeO in the core. We also include FeS (VII) in the core. We have also included an approximate phase diagram for water ice to account for an icy mantle. We also include the effects of adiabatic temperature profiles and a temperature jump at the core-mantle boundary. Finally, we have created a user-friendly interface allowing the code to be downloaded and used as a teaching tool. Results of the code and a demonstration of its use will be presented at the meeting.

  11. Investigation of methods used in calculations of solar cell parameters

    OpenAIRE

    Shvets, E. Ya.; Khrypko, S. L.; Zubko, E. I.

    2009-01-01

    Analytical expressions have been obtained for extracting the electrical parameters and characteristics of solar cells, including series and shunt resistances, and the saturation current. The method of Lagrange multipliers was used for computing the shape factor of the current–voltage characteristic (CVC) of solar cell. The calculation results demonstrated a satisfactory agreement with experimental data.

  12. TRANCS, a computer code for calculating fission product release from high temperature gas-cooled reactor fuel, (2)

    International Nuclear Information System (INIS)

    This report describes the calculation procedure of the TRANCS code, which deals with fission product transport in fuel rod of high temperature gas-cooled reactor (HTGR). The fundamental equation modeled in the code is a cylindrical one-dimensional diffusion equation with generation and decay terms, and the non-stationary solution of the equation is obtained numerically by a finite difference method. The generation terms consist of the diffusional release from coated fuel particles, recoil release from outer-most coating layer of the fuel particle and generation due to contaminating uranium in the graphite matrix of the fuel compact. The decay term deals with neutron capture as well as beta decay. Factors affecting the computation error has been examined, and further extention of the code has been discussed in the fields of radial transport of fission products from graphite sleeve into coolant helium gas and axial transport in the fuel rod. (author)

  13. Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    1998-10-01

    Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)

  14. RO-75: a FORTRAN code for calculation and design optimization of reverse osmosis seawater desalination plants

    Energy Technology Data Exchange (ETDEWEB)

    Glueckstern, P.; Reed, S.A.; Wilson, J.V.

    1976-11-01

    The reverse osmosis process has been used extensively for the conversion of brackish waters to potable water. The process is now nearing commercialization as a means for the conversion of seawater. The computer program (RO-75) is a Fortran code for the optimizatin of the design and economics of seawater reverse osmosis plants. The examples described are based on currently available, commercial membrane modules and prevailing prices. However, the code is very flexible and can be used to optimize plants utilizing future technological improvements and different economic parameters.

  15. New developments of the CARTE thermochemical code: Calculation of detonation properties of high explosives

    Science.gov (United States)

    Dubois, Vincent; Desbiens, Nicolas; Auroux, Eric

    2010-07-01

    We present the improvements of the CARTE thermochemical code which provides thermodynamic properties and chemical compositions of CHON systems over a large range of temperature and pressure with a very small computational cost. The detonation products are split in one or two fluid phase (s), treated with the MCRSR equation of state (EOS), and one condensed phase of carbon, modeled with a multiphase EOS which evolves with the chemical composition of the explosives. We have developed a new optimization procedure to obtain an accurate multicomponents EOS. We show here that the results of CARTE code are in good agreement with the specific data of molecular systems and measured detonation properties for several explosives.

  16. Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes

    International Nuclear Information System (INIS)

    The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)

  17. OECD/CSNI ISP Nr. 43 rapid boron dilution transient tests for code verification post test calculation with CFX-4

    International Nuclear Information System (INIS)

    The need of the experimental support for validation of the computational tools to be applied to analyze the mixing of diluted slugs has been recognized in various countries. The test series for the International Standard Problem ISP-43 provides a platform for experiences to be applied to the simulation of a well-defined test series. Test A and B of the UM2x4 loop test facility were calculated with the CFD Code CFX-4.3. The results show qualitatively good agreement with the experimental data for both tests. The structure of the flow field and the form of the propagating temperature perturbation front are well modeled by the CFD code. However, deviations occur at local positions. Comparative calculations with and without taking into account buoyancy have shown, that buoyancy effects are noticeable, but the mixing is mainly momentum controlled. (orig.)

  18. Statistical uncertainty analysis applied to the DRAGONv4 code lattice calculations and based on JENDL-4 covariance data

    International Nuclear Information System (INIS)

    In this paper, multi-group microscopic cross-section uncertainty is propagated through the DRAGON (Version 4) lattice code, in order to perform uncertainty analysis on k∞ and 2-group homogenized macroscopic cross-sections predictions. A statistical methodology is employed for such purposes, where cross-sections of certain isotopes of various elements belonging to the 172 groups DRAGLIB library format, are considered as normal random variables. This library is based on JENDL-4 data, because JENDL-4 contains the largest amount of isotopic covariance matrixes among the different major nuclear data libraries. The aim is to propagate multi-group nuclide uncertainty by running the DRAGONv4 code 500 times, and to assess the output uncertainty of a test case corresponding to a 17 x 17 PWR fuel assembly segment without poison. The chosen sampling strategy for the current study is Latin Hypercube Sampling (LHS). The quasi-random LHS allows a much better coverage of the input uncertainties than simple random sampling (SRS) because it densely stratifies across the range of each input probability distribution. Output uncertainty assessment is based on the tolerance limits concept, where the sample formed by the code calculations infers to cover 95% of the output population with at least a 95% of confidence. This analysis is the first attempt to propagate parameter uncertainties of modern multi-group libraries, which are used to feed advanced lattice codes that perform state of the art resonant self-shielding calculations such as DRAGONv4. (authors)

  19. WETAIR: A computer code for calculating thermodynamic and transport properties of air-water mixtures

    Science.gov (United States)

    Fessler, T. E.

    1979-01-01

    A computer program subroutine, WETAIR, was developed to calculate the thermodynamic and transport properties of air water mixtures. It determines the thermodynamic state from assigned values of temperature and density, pressure and density, temperature and pressure, pressure and entropy, or pressure and enthalpy. The WETAIR calculates the properties of dry air and water (steam) by interpolating to obtain values from property tables. Then it uses simple mixing laws to calculate the properties of air water mixtures. Properties of mixtures with water contents below 40 percent (by mass) can be calculated at temperatures from 273.2 to 1497 K and pressures to 450 MN/sq m. Dry air properties can be calculated at temperatures as low as 150 K. Water properties can be calculated at temperatures to 1747 K and pressures to 100 MN/sq m. The WETAIR is available in both SFTRAN and FORTRAN.

  20. A computer code for beam optics calculation--third order approximation

    Institute of Scientific and Technical Information of China (English)

    L(U) Jianqin; LI Jinhai

    2006-01-01

    To calculate the beam transport in the ion optical systems accurately, a beam dynamics computer program of third order approximation is developed. Many conventional optical elements are incorporated in the program. Particle distributions of uniform type or Gaussian type in the ( x, y, z ) 3D ellipses can be selected by the users. The optimization procedures are provided to make the calculations reasonable and fast. The calculated results can be graphically displayed on the computer monitor.

  1. The Wims-Traca code for the calculation of fuel elements. User's manual and input data

    International Nuclear Information System (INIS)

    The set of modifications and new options developped for the Wims-D code is explained. The input data of the new version Wims-Traca are described. The printed output of results is also explained. The contents and the source of the nuclear data in the basic library is exposed. (author)

  2. Calculation of proton and neutron emission spectra from proton reactions with {sup 90}Zr and {sup 208}Pb to 160 MeV with the GNASH code

    Energy Technology Data Exchange (ETDEWEB)

    Young, P.G. [Los Alamos National Lab., NM (United States); Chadwick, M.B. [Lawrence Livermore National Lab., CA (United States)

    1994-06-01

    A number of modifications have been made to the reaction theory code GNASH in order the accuracy of calculations at incident particle energies up to 200 MeV. Direct reaction a level density models appropriate for higher energy calculations are now used in the code, and most importantly, improved preequilibrium models have been incorporated into the code system. The code has been used to calculate proton-induced reactions on {sup 90}Zr and {sup 208}Pb for the International Code and Model Intercomparison for Intermediate Energy Reactions organized by the NEA. Calculations were performed with GNASH at incident proton energies of 25, 45, 80, and 160 mev using both the exciton model and Feshbach-Kerman-Koonin theory for the preequilibrium component. The models and procedures used in the GNASH calculations with the exciton model are described here. The results are compared to experimental data and to results from the GNASH calculations with Feshbach-Kerman-Koonin preequilibrium theory.

  3. Determination of the fuel element burn-up for mixed TRIGA core by measurement and calculation with new TRIGLAV code

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))

    1999-12-15

    Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)

  4. Off-design computer code for calculating the aerodynamic performance of axial-flow fans and compressors

    Science.gov (United States)

    Schmidt, James F.

    1995-01-01

    An off-design axial-flow compressor code is presented and is available from COSMIC for predicting the aerodynamic performance maps of fans and compressors. Steady axisymmetric flow is assumed and the aerodynamic solution reduces to solving the two-dimensional flow field in the meridional plane. A streamline curvature method is used for calculating this flow-field outside the blade rows. This code allows for bleed flows and the first five stators can be reset for each rotational speed, capabilities which are necessary for large multistage compressors. The accuracy of the off-design performance predictions depend upon the validity of the flow loss and deviation correlation models. These empirical correlations for the flow loss and deviation are used to model the real flow effects and the off-design code will compute through small reverse flow regions. The input to this off-design code is fully described and a user's example case for a two-stage fan is included with complete input and output data sets. Also, a comparison of the off-design code predictions with experimental data is included which generally shows good agreement.

  5. Domain Decomposition strategy for pin-wise full-core Monte Carlo depletion calculation with the reactor Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)

    2016-06-15

    Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.

  6. Benchmark test of TRIPOLI-4 code through simple model calculation and analysis of fusion neutronics experiments at JAEA/FNS

    Energy Technology Data Exchange (ETDEWEB)

    Ohta, Masayuki, E-mail: ohta.masayuki@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)

    2013-10-15

    In order to examine a basic performance of the TRIPOLI code, two types of analyses were carried out with TRIPOLI-4.4 and MCNP5-1.40; one is a simple model calculation and the other is an analysis of iron fusion neutronics experiments with DT neutrons at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA). In the simple model calculation, we adopted a sphere of 0.5 m in radius with a 20 MeV neutron source in the center and calculated leakage neutron spectra from the sphere. We also analyzed in situ and Time-of-Flight (TOF) experiments for iron at JAEA/FNS. For the in situ experiment, neutron spectra and reaction rates for dosimetry reactions were calculated for several points inside the assembly. For the TOF experiment, angular neutron leakage spectra from the assembly were calculated. Results with TRIPOLI were comparable to those with MCNP in most calculations, but a difference between TRIPOLI and MCNP calculation results, probably caused by inadequate treatment of inelastic scattering data in TRIPOLI, appears in some calculations.

  7. BOA, Beam Optics Analyzer A Particle-In-Cell Code

    Energy Technology Data Exchange (ETDEWEB)

    Thuc Bui

    2007-12-06

    The program was tasked with implementing time dependent analysis of charges particles into an existing finite element code with adaptive meshing, called Beam Optics Analyzer (BOA). BOA was initially funded by a DOE Phase II program to use the finite element method with adaptive meshing to track particles in unstructured meshes. It uses modern programming techniques, state-of-the-art data structures, so that new methods, features and capabilities are easily added and maintained. This Phase II program was funded to implement plasma simulations in BOA and extend its capabilities to model thermal electrons, secondary emissions, self magnetic field and implement a more comprehensive post-processing and feature-rich GUI. The program was successful in implementing thermal electrons, secondary emissions, and self magnetic field calculations. The BOA GUI was also upgraded significantly, and CCR is receiving interest from the microwave tube and semiconductor equipment industry for the code. Implementation of PIC analysis was partially successful. Computational resource requirements for modeling more than 2000 particles begin to exceed the capability of most readily available computers. Modern plasma analysis typically requires modeling of approximately 2 million particles or more. The problem is that tracking many particles in an unstructured mesh that is adapting becomes inefficient. In particular memory requirements become excessive. This probably makes particle tracking in unstructured meshes currently unfeasible with commonly available computer resources. Consequently, Calabazas Creek Research, Inc. is exploring hybrid codes where the electromagnetic fields are solved on the unstructured, adaptive mesh while particles are tracked on a fixed mesh. Efficient interpolation routines should be able to transfer information between nodes of the two meshes. If successfully developed, this could provide high accuracy and reasonable computational efficiency.

  8. Development of a computer code for neutronic calculations of a hexagonal lattice of nuclear reactor using the flux expansion nodal method

    Directory of Open Access Journals (Sweden)

    Mohammadnia Meysam

    2013-01-01

    Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

  9. Verification calculations as per CFD FLOWVISION code for sodium-cooled reactor plants

    International Nuclear Information System (INIS)

    The paper studies the experience in application of CFD FlowVision software for analytical validation of sodium-cooled fast reactor structure components and the results of performed verification, namely: – development and implementation of new model of turbulent heat transfer in liquid sodium (LMS) in FlowVision software and model verification based on thermohydraulic characteristics studied by experiment at TEFLU test facility; – simulation of flowing and mixing of coolant with different temperatures in the upper mixing chamber of fast neutron reactor through the example of BN-600 (comparison with the results obtained at the operating reactor). Based on the analysis of the results obtained, the efficiency of CFD codes application for the considered problems is shown, and the proposals for CFD codes verification development as applied to the advanced sodium-cooled fast reactor designs are stated. (author)

  10. Calculation of the Phenix end-of-life test “control rod withdrawal” with the ERANOS code

    International Nuclear Information System (INIS)

    The Institute of Radioprotection and Nuclear Safety (IRSN) being established as technical support organization for French public authorities is in charge of safety assessment of both operating and under construction reactors and nuclear facilities. It provides safety studies of advanced and innovative projects like fast sodium cooled reactors as well. In this context, one of the IRSN objectives is to evaluate comprehensively the accuracy of numerical tools and their performance on studies of safety relay items. Reactor physics studies step in the safety assessment support from different points of view, among which the design of core and its protection system. They are essential in the cores behavior analysis in normal, perturbed and accidental conditions in order to assess the integrity of the first barrier and the exclusion of prompt criticality and re-criticality risks. The codes capability to compute in an accurate manner the fission power distribution in the core during the whole reactor lifetime could indicate the codes' accuracy for many so-called spatial dependent values calculations. The IAEA Coordinated Research Project on the Phenix end-of-life test “Control Rod Withdrawal” has been a good opportunity to check the capability of calculation tools by comparison with the measured radial power distributions on fast reactor. IRSN participated to this benchmark with the ERANOS code package developed by CEA for fast reactors studies. The challenge for this code package was that in the considered core configurations the neutron fields were notably deformed. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement has been found with available measures considering the approximations done in the modeling. The work underlines the importance of precise knowledge of the details of burn-up distribution as it could impact the calculations of the power distribution. (author)

  11. Inverse Load Calculation of Wind Turbine Support Structures - A Numerical Verification Using the Comprehensive Simulation Code FAST: Preprint (Revised)

    Energy Technology Data Exchange (ETDEWEB)

    Pahn, T.; Jonkman, J.; Rolges, R.; Robertson, A.

    2012-11-01

    Physically measuring the dynamic responses of wind turbine support structures enables the calculation of the applied loads using an inverse procedure. In this process, inverse means deriving the inputs/forces from the outputs/responses. This paper presents results of a numerical verification of such an inverse load calculation. For this verification, the comprehensive simulation code FAST is used. FAST accounts for the coupled dynamics of wind inflow, aerodynamics, elasticity and turbine controls. Simulations are run using a 5-MW onshore wind turbine model with a tubular tower. Both the applied loads due to the instantaneous wind field and the resulting system responses are known from the simulations. Using the system responses as inputs to the inverse calculation, the applied loads are calculated, which in this case are the rotor thrust forces. These forces are compared to the rotor thrust forces known from the FAST simulations. The results of these comparisons are presented to assess the accuracy of the inverse calculation. To study the influences of turbine controls, load cases in normal operation between cut-in and rated wind speed, near rated wind speed and between rated and cut-out wind speed are chosen. The presented study shows that the inverse load calculation is capable of computing very good estimates of the rotor thrust. The accuracy of the inverse calculation does not depend on the control activity of the wind turbine.

  12. Fast neutron reaction data calculations with the computer code STAPRE-H

    International Nuclear Information System (INIS)

    Description of the specific features of the version STAPRE-H are given. Illustration of the model options and parameter influence on the calculated results is done to trace the accurate reproducing of large body of correlated data. (authors)

  13. Beam Dynamics in an Electron Lens with the Warp Particle-in-cell Code

    CERN Document Server

    Stancari, Giulio; Redaelli, Stefano

    2014-01-01

    Electron lenses are a mature technique for beam manipulation in colliders and storage rings. In an electron lens, a pulsed, magnetically confined electron beam with a given current-density profile interacts with the circulating beam to obtain the desired effect. Electron lenses were used in the Fermilab Tevatron collider for beam-beam compensation, for abort-gap clearing, and for halo scraping. They will be used in RHIC at BNL for head-on beam-beam compensation, and their application to the Large Hadron Collider for halo control is under development. At Fermilab, electron lenses will be implemented as lattice elements for nonlinear integrable optics. The design of electron lenses requires tools to calculate the kicks and wakefields experienced by the circulating beam. We use the Warp particle-in-cell code to study generation, transport, and evolution of the electron beam. For the first time, a fully 3-dimensional code is used for this purpose.

  14. GOLEM: a versatile computer code for reactor neutronic calculation advances in qualification of the different modules

    International Nuclear Information System (INIS)

    The last 12 years studies about the CABRI, SCARABEE and PHEBUS projects are summarized. It describes the object and the genesis of the cores, the evolution of the core concept and the associated neutronic problems. The calculational scheme used is presented, together with its qualification. The formalism, and the qualification of the different modules of GOLEM are presented. COXYS: module of physical analysis in order to determine the best energetic and spatial mesh for the case of interest. GOLU.B: input data management module. VAREC: calculation module of perturbations due to materials enables to compute perturbed flux and reactivity variation. VARYX: calculation module of geometric perturbations. TRACASYN: module of 3D power shape calculation. Finally TRACASTORE: module of management and graphic exploitation of results. Then, one gives utilization directions for these different modules. Qualification results show that GOLEM is able to analyse the fine physics of many various cases, to calculate by perturbation effects greater than 5000 pcm, to rebuild perturbed flux with margins near 3% for difficult situations, like reactor voiding or spectral or spectral variation in a PWR. Furthermore, 3D hot spots are calculated within margins of a magnitude comparable to experimental ones

  15. The Plasma Simulation Code: A modern particle-in-cell code with load-balancing and GPU support

    CERN Document Server

    Germaschewski, Kai; Ahmadi, Narges; Wang, Liang; Abbott, Stephen; Ruhl, Hartmut; Bhattacharjee, Amitava

    2013-01-01

    Recent increases in supercomputing power, driven by the multi-core revolution and accelerators such as the IBM Cell processor, graphics processing units (GPUs) and Intel's Many Integrated Core (MIC) technology have enabled kinetic simulations of plasmas at unprecedented resolutions, but changing HPC architectures also come with challenges for writing efficient numerical codes. This paper describes the Plasma Simulation Code (PSC), an explicit, electromagnetic particle-in-cell code with support for different order particle shape functions. We focus on two distinguishing feature of the code: patch-based load balancing using space-filling curves, and support for Nvidia GPUs, which achieves substantial speed-up of up to more than 6x on the Cray XK7 architecture compared to a CPU-only implementation.

  16. Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

    Energy Technology Data Exchange (ETDEWEB)

    Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

    1978-12-01

    The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment.

  17. Computer calculation of neutron cross sections with Hauser-Feshbach code STAPRE incorporating the hybrid pre-compound emission model

    International Nuclear Information System (INIS)

    Computer codes incorporating advanced nuclear models (optical, statistical and pre-equilibrium decay nuclear reaction models) were used to calculate neutron cross sections needed for fusion reactor technology. The elastic and inelastic scattering (n,2n), (n,p), (n,n'p), (n,d) and (n,γ) cross sections for stable molybdenum isotopes Mosup(92,94,95,96,97,98,100) and incident neutron energy from about 100 keV or a threshold to 20 MeV were calculated using the consistent set of input parameters. The hydrogen production cross section which determined the radiation damage in structural materials of fusion reactors can be simply deduced from the presented results. The more elaborated microscopic models of nuclear level density are required for high accuracy calculations

  18. STATIC{sub T}EMP: a useful computer code for calculating static formation temperatures in geothermal wells

    Energy Technology Data Exchange (ETDEWEB)

    Santoyo, E. [Universidad Nacional Autonoma de Mexico, Centro de Investigacion en Energia, Temixco (Mexico); Garcia, A.; Santoyo, S. [Unidad Geotermia, Inst. de Investigaciones Electricas, Temixco (Mexico); Espinosa, G. [Universidad Autonoma Metropolitana, Co. Vicentina (Mexico); Hernandez, I. [ITESM, Centro de Sistemas de Manufactura, Monterrey (Mexico)

    2000-07-01

    The development and application of the computer code STATIC{sub T}EMP, a useful tool for calculating static formation temperatures from actual bottomhole temperature data logged in geothermal wells is described. STATIC{sub T}EMP is based on five analytical methods which are the most frequently used in the geothermal industry. Conductive and convective heat flow models (radial, spherical/radial and cylindrical/radial) were selected. The computer code is a useful tool that can be reliably used in situ to determine static formation temperatures before or during the completion stages of geothermal wells (drilling and cementing). Shut-in time and bottomhole temperature measurements logged during well completion activities are required as input data. Output results can include up to seven computations of the static formation temperature by each wellbore temperature data set analysed. STATIC{sub T}EMP was written in Fortran-77 Microsoft language for MS-DOS environment using structured programming techniques. It runs on most IBM compatible personal computers. The source code and its computational architecture as well as the input and output files are described in detail. Validation and application examples on the use of this computer code with wellbore temperature data (obtained from specialised literature) and with actual bottomhole temperature data (taken from completion operations of some geothermal wells) are also presented. (Author)

  19. TRANCS, a computer code for calculating fission product release from high temperature gas-cooled reactor fuel, (1)

    International Nuclear Information System (INIS)

    The computer program, TRANCS, has been developed for evaluating the fractional release of long-lived fission products from coated fuel particles. This code numerically gives the non-stationary solution of the diffusion equation with birth and decay terms. The birth term deals with the fissile material in the fuel kernel, the contamination in the coating layers and the fission-recoil transfer from the kernel into the buffer layer; and the decay term deals with effective decay not only due to beta decay but also due to neutron capture, if appropriate input data are given. The code calculates the concentration profile, the release to birth rates (R/B), and the release and residual fractions in the coated fuel particle. Results obtained numerically have been in good agreement with the corresponding analytical solutions after the Booth model. Thus, the validity of the present code was confirmed, and further undate of the code has been discussed for extention of its computation scopes and models. (author)

  20. Second order gyrokinetic theory for particle-in-cell codes

    Science.gov (United States)

    Tronko, Natalia; Bottino, Alberto; Sonnendrücker, Eric

    2016-08-01

    The main idea of the gyrokinetic dynamical reduction consists in a systematical removal of the fast scale motion (the gyromotion) from the dynamics of the plasma, resulting in a considerable simplification and a significant gain of computational time. The gyrokinetic Maxwell-Vlasov equations are nowadays implemented in for modeling (both laboratory and astrophysical) strongly magnetized plasmas. Different versions of the reduced set of equations exist, depending on the construction of the gyrokinetic reduction procedure and the approximations performed in the derivation. The purpose of this article is to explicitly show the connection between the general second order gyrokinetic Maxwell-Vlasov system issued from the modern gyrokinetic theory and the model currently implemented in the global electromagnetic Particle-in-Cell code ORB5. Necessary information about the modern gyrokinetic formalism is given together with the consistent derivation of the gyrokinetic Maxwell-Vlasov equations from first principles. The variational formulation of the dynamics is used to obtain the corresponding energy conservation law, which in turn is used for the verification of energy conservation diagnostics currently implemented in ORB5. This work fits within the context of the code verification project VeriGyro currently run at IPP Max-Planck Institut in collaboration with others European institutions.

  1. State of the art of aerolastic codes for wind turbine calculations

    Energy Technology Data Exchange (ETDEWEB)

    Maribo Pedersen, B. [ed.

    1996-09-01

    The technological development of modern wind turbines has been dependent on the parallel development of the computational skills of the designers. The combination of the calculation of the flow field around the wind turbine rotor - both far field and near field - and the calculation of the response of the wind turbine structure to the resulting, non-stationary air loads, also known as aero-elastic calculations have now reached a reasonable degree of maturity. At this expert meeting two main points may be clarified. To what level of accuracy can we now determine the behaviour of the different elements of a wind turbine, i.e. how well are we able to compute deflections, fluctuating loads and power output. Which are the main outstanding areas upon which our next research efforts should be focused. (EG)

  2. WASP: A flexible FORTRAN 4 computer code for calculating water and steam properties

    Science.gov (United States)

    Hendricks, R. C.; Peller, I. C.; Baron, A. K.

    1973-01-01

    A FORTRAN 4 subprogram, WASP, was developed to calculate the thermodynamic and transport properties of water and steam. The temperature range is from the triple point to 1750 K, and the pressure range is from 0.1 to 100 MN/m2 (1 to 1000 bars) for the thermodynamic properties and to 50 MN/m2 (500 bars) for thermal conductivity and to 80 MN/m2 (800 bars) for viscosity. WASP accepts any two of pressure, temperature, and density as input conditions. In addition, pressure and either entropy or enthalpy are also allowable input variables. This flexibility is especially useful in cycle analysis. The properties available in any combination as output include temperature, density, pressure, entropy, enthalpy, specific heats, sonic velocity, viscosity, thermal conductivity, surface tension, and the Laplace constant. The subroutine structure is modular so that the user can choose only those subroutines necessary to his calculations. Metastable calculations can also be made by using WASP.

  3. Reactor pressure vessel strength calculations - comparing the AD/TRD and the ASME code. Pt. 1

    International Nuclear Information System (INIS)

    The dimensioning criteria applied in the various technical rules are illustrated by the example of a reactor pressure vessel nozzle, especially with a view to the characteristic data of the materials used. Using a detailed finite element analysis of the main coolant nozzle permits an evaluation of the different calculation methods. The second part of the report discusses safety problems, e.g. fatigue analysis, the necessity of carrying out 3D-elastoplastic FE calculations, or the assessment of transient loads on the reactor pressure vessel by means of a fracture-mechanical analysis. (orig./HP)

  4. The three-dimensional PWR transient code ANTI; rod ejection test calculation

    International Nuclear Information System (INIS)

    ANTI is a computer program being developed for three-dimensional coupled neutronics and thermal-hydraulics description of a PWR core under transient conditions. In this report a test example calculated by the program is described. The test example is a simulation of a control rod ejection from a very small reactor core (to save somputing time). In order to show the influence of cross flow between adjacent fuel elements the same calculation was performed both with the cross flow option and with closed hydraulic channels. (author)

  5. Second order Gyrokinetic theory for Particle-In-Cell codes

    CERN Document Server

    Tronko, Natalia; Sonnendruecker, Eric

    2016-01-01

    The main idea of Gyrokinetic dynamical reduction consists in systematical removing of fastest scale of motion (the gyro motion) from plasma's dynamics, resulting in a considerable model simplification and gain of computing time. Gyrokinetic Maxwell-Vlasov system is broadly implemented in nowadays numerical experiments for modeling strongly magnetized plasma (both laboratory and astrophysical). Different versions of reduced set of equations exist depending on the construction of the Gyrokinetic reduction procedure and approximations assumed while their derivation. The purpose of this paper is to explicitly show the connection between the general second order gyrokinetic Maxwell-Vlasov system issued from the Modern Gyrokinetic theory derivation and the model currently implemented in global electromagnetic Particle in Cell code ORB5. Strictly necessary information about the Modern Gyrokinetic formalism is given together with the consistent derivation of the gyrokinetic Maxwell-Vlasov equations from the first pri...

  6. A New Radiation Hydrodynamics Code and Application to the Calculation of Type Ia Supernovae Light Curves and Continuum Spectra

    CERN Document Server

    Zhang, X; Zhang, Xiao-he; Sutherland, Peter

    1993-01-01

    A new, fully dynamic and self-consistent radiation hydrodynamics code, suitable for the calculation of supernovae light curves and continuum spectra, is described. It is a multigroup (frequency-dependent) code and includes all important $O(v/c)$ effects. It is applied to the model W7 of Nomoto, Thielemann, \\& Yokoi (1984) for supernovae of type Ia. Radioactive energy deposition is incorporated through use of tables based upon Monte Carlo results. Effects of line opacity (both static or line blanketing and expansion or line blocking) are neglected, although these may prove to be important. At maximum light, models based upon different treatments of the opacity lead to values for $M_{B,max}$ in the range of -19.0 to -19.4. This range falls between the values for observed supernova claimed by Leibundgut \\& Tammann (1990) and by Pierce, Ressler, \\& Shure (1992).

  7. Brief evaluation of the radiological hazards after a nuclear accident - description and mode of operation of this calculation code Orion

    International Nuclear Information System (INIS)

    The ORION code is designed to determine very quickly the immediate consequences (such as plume passage time, instantaneous maximum hazards irradiation, inhalation, deposit) due to an accident spreading out radioactive or chemical pollution into the atmosphere, from a source point, a stack release, (with heightening calculation) outspread sources (transport accident such as, for instance, road fire or car crash) or from a cylindrical cloud defined by different vertical sources (for instance pyrotechnical accident, missile firing...). The diffusion code DOURY type (french official methods) is written in FORTRAN. Data are entered in a conversational mode with auto-checking. Results are output to tables an isorisks curves drawn at map scales. At the Bruyeres-le-Chatel Radiation Protection Unit, a team is on permanent duty, can carry out results in a few minutes and transmit the evaluation by TELEFAX anywhere on the National territory

  8. Decay heat measurement on fusion reactor materials and validation of calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

  9. Calculation of electron and isotopes dose point kernels with FLUKA Monte Carlo code for dosimetry in nuclear medicine therapy

    CERN Document Server

    Mairani, A; Valente, M; Battistoni, G; Botta, F; Pedroli, G; Ferrari, A; Cremonesi, M; Di Dia, A; Ferrari, M; Fasso, A

    2011-01-01

    Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy ((89)Sr, (90)Y, (131)I, (153)Sm, (177)Lu, (186)Re, and (188)Re). Point isotropic...

  10. Diffusion coefficients for LMFBR cells calculated with MOC and Monte Carlo methods

    Energy Technology Data Exchange (ETDEWEB)

    Rooijen, W.F.G. van, E-mail: rooijen@u-fukui.ac.j [Research Institute of Nuclear Energy, University of Fukui, Bunkyo 3-9-1, Fukui-shi, Fukui-ken 910-8507 (Japan); Chiba, G., E-mail: chiba.go@jaea.go.j [Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)

    2011-01-15

    The present work discusses the calculation of the diffusion coefficient of a lattice of hexagonal cells, with both 'sodium present' and 'sodium absent' conditions. Calculations are performed in the framework of lattice theory (also known as fundamental mode approximation). Unlike the classical approaches, our heterogeneous leakage model allows the calculation of diffusion coefficients under all conditions, even if planar voids are present in the lattice. Equations resulting from this model are solved using the method of characteristics (MOC). Independent confirmation of the MOC result comes from Monte Carlo calculations, in which the diffusion coefficient is obtained without any of the assumptions of lattice theory. It is shown by comparison to the Monte Carlo results that the MOC solution yields correct values of the diffusion coefficient under all conditions, even in cases where the classic calculation of the diffusion coefficient fails. This work is a first step in the development of a robust method to calculate the diffusion coefficient of lattice cells. Adoption into production codes will require more development and validation of the method.

  11. Prostate dose calculations for permanent implants using the MCNPX code and the Voxels phantom MAX

    Energy Technology Data Exchange (ETDEWEB)

    Reis Junior, Juraci Passos dos; Silva, Ademir Xavier da, E-mail: jjunior@con.ufrj.b, E-mail: Ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Facure, Alessandro N.S., E-mail: facure@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2010-07-01

    This paper presents the modeling of 80, 88 and 100 of {sup 125}I seeds, punctual and volumetric inserted into the phantom spherical volume representing the prostate and prostate phantom voxels MAX. Starting values of minimum and maximum activity, 0.27 mCi and 0.38 mCi, respectively, were simulated in the Monte Carlo code MCNPX in order to determine whether the final dose, according to the integration of the equation of decay at time t = 0 to t = {infinity} corresponds to the default value set by the AAPM 64 which is 144 Gy. The results showed that consider sources results in doses exceeding the percentage discrepancy of the default value of 200%, while volumetric consider sources result in doses close to 144 Gy. (author)

  12. Grid cells generate an analog error-correcting code for singularly precise neural computation.

    Science.gov (United States)

    Sreenivasan, Sameet; Fiete, Ila

    2011-09-11

    Entorhinal grid cells in mammals fire as a function of animal location, with spatially periodic response patterns. This nonlocal periodic representation of location, a local variable, is unlike other neural codes. There is no theoretical explanation for why such a code should exist. We examined how accurately the grid code with noisy neurons allows an ideal observer to estimate location and found this code to be a previously unknown type of population code with unprecedented robustness to noise. In particular, the representational accuracy attained by grid cells over the coding range was in a qualitatively different class from what is possible with observed sensory and motor population codes. We found that a simple neural network can effectively correct the grid code. To the best of our knowledge, these results are the first demonstration that the brain contains, and may exploit, powerful error-correcting codes for analog variables.

  13. PyVCI: A flexible open-source code for calculating accurate molecular infrared spectra

    Science.gov (United States)

    Sibaev, Marat; Crittenden, Deborah L.

    2016-06-01

    The PyVCI program package is a general purpose open-source code for simulating accurate molecular spectra, based upon force field expansions of the potential energy surface in normal mode coordinates. It includes harmonic normal coordinate analysis and vibrational configuration interaction (VCI) algorithms, implemented primarily in Python for accessibility but with time-consuming routines written in C. Coriolis coupling terms may be optionally included in the vibrational Hamiltonian. Non-negligible VCI matrix elements are stored in sparse matrix format to alleviate the diagonalization problem. CPU and memory requirements may be further controlled by algorithmic choices and/or numerical screening procedures, and recommended values are established by benchmarking using a test set of 44 molecules for which accurate analytical potential energy surfaces are available. Force fields in normal mode coordinates are obtained from the PyPES library of high quality analytical potential energy surfaces (to 6th order) or by numerical differentiation of analytic second derivatives generated using the GAMESS quantum chemical program package (to 4th order).

  14. Comparison of energy deposition calculations by the LAHET Code System with experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Beard, C.A.; Lisowski, P.W.; Russell, G.J.; Waters, L.S.

    1993-08-01

    A comparison was performed between the energy deposition predicted by the LAHET Code System (LCS) with experimental values determined by Belyakov-Bodin et al. for 800, 1000, and 1200 MeV protons on targets composed of lead, bismuth, beryllium, carbon, and aluminum. The lead and bismuth showed agreement within approximately 10% at locations throughout the targets, and the agreement of the total energy deposited over the axial length of the targets ranged from 1% to 25%. For the lead and bismuth cases, the LCS predictions were always greater than the experimental results. For the lighter materials, the agreement at locations throughout the target only agreed within approximately 20%. No definable trend could be determined for the lighter materials since some LCS predictions were greater than the experimental results, some were less than the experimental results, and some showed very good agreement. The total energy deposited over the axial length of the targets was not compared for the lighter materials since it was not explicitly given with the experimental data.

  15. An interactive computer code for calculation of gas-phase chemical equilibrium (EQLBRM)

    Science.gov (United States)

    Pratt, B. S.; Pratt, D. T.

    1984-01-01

    A user friendly, menu driven, interactive computer program known as EQLBRM which calculates the adiabatic equilibrium temperature and product composition resulting from the combustion of hydrocarbon fuels with air, at specified constant pressure and enthalpy is discussed. The program is developed primarily as an instructional tool to be run on small computers to allow the user to economically and efficiency explore the effects of varying fuel type, air/fuel ratio, inlet air and/or fuel temperature, and operating pressure on the performance of continuous combustion devices such as gas turbine combustors, Stirling engine burners, and power generation furnaces.

  16. Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project

    International Nuclear Information System (INIS)

    In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)

  17. Benchmark calculations of a radiation heat transfer for a CANDU fuel channel analysis using the CFD code

    International Nuclear Information System (INIS)

    To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to a whole fuel channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with those solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer. (author)

  18. A parallel code to calculate rate-state seismicity evolution induced by time dependent, heterogeneous Coulomb stress changes

    Science.gov (United States)

    Cattania, C.; Khalid, F.

    2016-09-01

    The estimation of space and time-dependent earthquake probabilities, including aftershock sequences, has received increased attention in recent years, and Operational Earthquake Forecasting systems are currently being implemented in various countries. Physics based earthquake forecasting models compute time dependent earthquake rates based on Coulomb stress changes, coupled with seismicity evolution laws derived from rate-state friction. While early implementations of such models typically performed poorly compared to statistical models, recent studies indicate that significant performance improvements can be achieved by considering the spatial heterogeneity of the stress field and secondary sources of stress. However, the major drawback of these methods is a rapid increase in computational costs. Here we present a code to calculate seismicity induced by time dependent stress changes. An important feature of the code is the possibility to include aleatoric uncertainties due to the existence of multiple receiver faults and to the finite grid size, as well as epistemic uncertainties due to the choice of input slip model. To compensate for the growth in computational requirements, we have parallelized the code for shared memory systems (using OpenMP) and distributed memory systems (using MPI). Performance tests indicate that these parallelization strategies lead to a significant speedup for problems with different degrees of complexity, ranging from those which can be solved on standard multicore desktop computers, to those requiring a small cluster, to a large simulation that can be run using up to 1500 cores.

  19. Development of a computer code for shielding calculation in X-ray facilities

    International Nuclear Information System (INIS)

    The construction of an effective barrier against the interaction of ionizing radiation present in X-ray rooms requires consideration of many variables. The methodology used for specifying the thickness of primary and secondary shielding of an traditional X-ray room considers the following factors: factor of use, occupational factor, distance between the source and the wall, workload, Kerma in the air and distance between the patient and the receptor. With these data it was possible the development of a computer program in order to identify and use variables in functions obtained through graphics regressions offered by NCRP Report-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) for the calculation of shielding of the room walls as well as the wall of the darkroom and adjacent areas. With the built methodology, a program validation is done through comparing results with a base case provided by that report. The thickness of the obtained values comprise various materials such as steel, wood and concrete. After validation is made an application in a real case of radiographic room. His visual construction is done with the help of software used in modeling of indoor and outdoor. The construction of barriers for calculating program resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in September / 2011

  20. Validation of Neutron Calculation Codes and Models by means of benchmark cases in the frame of the Binational Commission of Nuclear Energy. Criticality Experiments

    International Nuclear Information System (INIS)

    In year 2008 the Atomic Energy National Commission (CNEA) of Argentina, and the Brazilian Institute of Energetic and Nuclear Research (IPEN), under the frame of Nuclear Energy Argentine Brazilian Agreement (COBEN), among many others, included the project “Validation and Verification of Calculation Methods used for Research and Experimental Reactors. At this time, it was established that the validation was to be performed with models implemented in the deterministic codes HUEMUL and PUMA (cell and reactor codes) developed by CNEA and those ones implemented in MCNP by CNEA and IPEN. The necessary data for these validations would correspond to theoretical-experimental reference cases in the research reactor IPEN/MB-01 located in São Paulo, Brazil. The staff of the group Reactor and Nuclear Power Studies (SERC) of CNEA, from the argentine side, performed calculations with deterministic models (HUEMUL-PUMA) and probabilistic methods (MCNP) modeling a great number of physical situations of de reactor, which previously have been studied and modeled by members of the Center of Nuclear Engineering of the IPEN, whose results were extensively provided to CNEA. In this paper results for critical configurations are shown. (author)

  1. HARAD: a computer code for calculating daughter concentrations in air following the atmospheric release of a parent radionuclide

    International Nuclear Information System (INIS)

    The HARAD computer code, written in FORTRAN IV, calculates concentrations of radioactive daughters in air following the atmospheric release of a parent radionuclide under a variety of meteorological conditions. It can be applied most profitably to the assessment of doses to man from the noble gases such as 222Rn, 220Rn, and Xe and Kr isotopes. These gases can produce significant quantities of short-lived particulate daughters in an airborne plume, which are the major contributors to dose from these chains with gaseous parent radionuclides. The simultaneous processes of radioactive decay, buildup, and environmental losses through wet and dry deposition on ground surfaces are calculated for a daughter chain in an airborne plume as it is dispersed downwind from a point of release of a parent. The code employs exact solutions of the differential equations describing the above processes over successive discrete segments of downwind distance. Average values for the dry deposition coefficients of the chain members over each of these distance segments were treated as constants in the equations. The advantage of HARAD is its short computing time

  2. Evaluation of ANGLE(R), a code for calculating HPGe detector efficiencies

    Energy Technology Data Exchange (ETDEWEB)

    Homan, Victoria M [Los Alamos National Laboratory

    2010-10-25

    This paper evaluates the ANGLE(reg sign) software package, an advanced efficiency calibration software for high purity germanium detectors that is distributed by ORTEC(reg sign). ANGLE(reg sign) uses a semi-empirical approach, by way of the efficiency transfer method, based on the calculated effective solid angle. This approach would have an advantage over the traditional relative and stochastic methods by decreasing the chances for systematic errors and reducing sensitivity to uncertainties in detector parameters. For experimental confirmation, a closed-end coaxial HPGe detector was used with sample geometries frequently encountered at the Los Alamos National Laboratory. The results obtained were sufficient for detector-source configurations which included intercepting layers of plexiglass and carbon graphite, but somewhat insufficient for bare source configurations.

  3. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

    2015-07-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  4. A computer code for calculation of solvent-extraction separation in a multicomponent system with reference to nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Nuclear technology development pointed out the need for a new assessment of the fuel cycle back-end. Treatment and disposal of radioactive wastes arising from nuclear fuel reprocessing is known as one of the problems not yet satisfactorily solved, together with separation process of uranium and plutonium from fission products in highly irradiated fuels. Aim of this work is to present an improvement of the computer code for solvent extraction process calculation previously designed by the authors. The modeling of the extraction system has been modified by introducing a new method for calculating the distribution coefficients. The new correlations were based on deriving empirical functions for not only the apparent equilibrium constants, but also the solvation number. The mathematical model derived for calculating separation performance has been then tested for up to ten components and twelve theoretical stages with minor modifications to the convergence criteria. Suitable correlations for the calculation of the distribution coefficients of Uranium, Plutonium, Nitric Acid and fission products were constructed and used to successfully simulate several experimental conditions. (Author)

  5. Calculation of conversion coefficients for effective dose for neutrons using a female voxel anthropomorphic model and the MCNPX code

    International Nuclear Information System (INIS)

    This work aims to calculate the fluence to effective dose conversion coefficients, (E/Φ), for monoenergetic neutrons from 10-9 to 20 MeV, based on the radiation (wR) and tissue (wT) weighting factors values recommended by ICRP publications numbers 60 and 103. The organs and tissues absorbed doses were calculated using the radiation transport code MCNPX and a female anthropomorphic voxel-based simulator, assuming whole-body irradiation by plane-parallel beams, on the geometries of the antero-posterior (AP) and postero-anterior (PA) irradiation. Dose calculations were performed for 21 selected organs of the body, for which the International Commission on Radiological Protection and the International Commission on Radiological Units and Measurements have set tissue weighting factors for the determination of the effective dose. From comparison between the dose results calculated and the data reported for the MIRD model, it can be concluded that, the fluence to effective dose conversion coefficients obtained using the voxel simulator are underestimated by a factor of up to 5 times when compared with the one obtained by ICRP 74, using mathematical simulators. (author)

  6. EFFI: a code for calculating the electromagnetic field, force, and inductance in coil systems of arbitrary geometry

    International Nuclear Information System (INIS)

    EFFI calculates the electromagnetic field and vector potential in coil systems of arbitrary geometry. The coils are made from circular arc and/or straight segments of rectangular cross-section conductor. EFFI can also calculate magnetic flux lines, magnetic force, and inductance. The methods used for the calculations are based on a combination of analytical and numerical integration of the Biot--Savart law for a volume distribution of current. These methods yield accurate field values inside and outside the conductor. All input to EFFI is format-free and is checked for validity before any calculations are done. Any errors detected during the check produce a diagnostic that lists the error, the code's objection to it, and the number of the offending data card. EFFI produces output in both printed and graphical form. Each page of output is labeled with the title of the problem, the time, the computer and date of the run, and the version number and compilation date for EFFI. In addition, each column of numbers on each page is appropriately labeled. Examples from the coil design for the Mirror Fusion Test Facility (MFTF) and a divertor design for a Tokamak reactor are used for illustration

  7. SIMPLE TRANSIENT CALCULATIONS OF CELL FLAMMABLE GAS CONCENTRATIONS

    Energy Technology Data Exchange (ETDEWEB)

    (NOEMAIL), J; David Allison (NOEMAIL), D; John Mccord, J

    2009-05-06

    The Saltstone Facility at Savannah River Site (SRS) mixes low-level radiological liquid waste with grout for permanent disposal as cement in vault cells. The grout mixture is poured into each cell in approximately 17 batches (8 to 10 hours duration). The grout mixture contains ten flammable gases of concern that are released from the mixture into the cell. Prior to operations, simple parametric transient calculations were performed to develop batch parameters (including schedule of batch pours) to support operational efficiency while ensuring that a flammable gas mixture does not develop in the cell vapor space. The analysis demonstrated that a nonflammable vapor space environment can be achieved, with workable operational constraints, without crediting the ventilation flow as a safety system control. Isopar L was identified as the primary flammable gas of concern. The transient calculations balanced inflows of the flammable gases into the vapor space with credited outflows of diurnal breathing through vent holes and displacement from new grout pours and gases generated. Other important features of the analyses included identifying conditions that inhibited a well-mixed vapor space, the expected frequency and duration of such conditions, and the estimated level of stratification that could develop.

  8. Implementation in the reaction code system EMPIRE-2.19 of an advanced formalism for fission cross-section calculation

    International Nuclear Information System (INIS)

    Full text: The implementation in the reaction code system EMPIRE-2.19 of an advanced formalism for fission cross-section calculation has been completed. The formalism is based on the optical model for fission and can be applied for nuclei exhibiting double- or triple-humped barrier starting from sub-barrier excitation energies. The optical model for fission, initially developed to describe the resonant structure of the fission cross section at sub-barrier excitation energies due to the vibrational states in the second well of a double-humped fission barrier, was extended to light actinides by including the relations for the transmission coefficients through a complex triple-humped fission barrier. The real part of the fission barrier is parameterised as a function of the nucleus deformation by five smoothly joined parabolas. The imaginary potential is introduced only in the deformation range corresponding to the second well because the tertiary well is supposed to be shallow enough to neglect the damping of class III vibrational states. The transition states are assumed to be rotational states built on vibrational or non-collective band-heads. As the excitation energy increases, the shell effect, which causes the splitting of the outer barrier, diminishes and the outer humps lump into a single one. Therefore, in the present formalism, triple-humped barriers are associated only to the discrete transition states; the contribution of continuum to the fission coefficients is calculated considering a double-humped barrier. The parameters of the second single barrier equivalent with the outer humps are being determined from the condition of equal transmission coefficients. The saddle-point transition states in continuum are described by level densities (BCS below the critical energy and a modified version of Fermi Gas above) accounting for collective enhancements specific to the nuclear shape asymmetry at each saddle point . The neutron cross sections of 232Th in the

  9. GNASH: a preequilibrium, statistical nuclear-model code for calculation of cross sections and emission spectra. [In FORTRAN for CDC 7600

    Energy Technology Data Exchange (ETDEWEB)

    Young, P.G.; Arthur, E.D.

    1977-11-01

    A new multistep Hauser--Feshbach code that includes corrections for preequilibrium effects is described. The code can calculate up to 60 decay reactions (cross sections and energy spectra) in one computation, and thereby provide considerable flexibility for handling processes with complicated reaction chains. Input parameter setup, problem output, and subroutine descriptions are given along with a sample problem calculation. A brief theoretical description is also included. 8 figures, 3 tables.

  10. Regulation of mammalian cell differentiation by long non-coding RNAs.

    Science.gov (United States)

    Hu, Wenqian; Alvarez-Dominguez, Juan R; Lodish, Harvey F

    2012-11-01

    Differentiation of specialized cell types from stem and progenitor cells is tightly regulated at several levels, both during development and during somatic tissue homeostasis. Many long non-coding RNAs have been recognized as an additional layer of regulation in the specification of cellular identities; these non-coding species can modulate gene-expression programmes in various biological contexts through diverse mechanisms at the transcriptional, translational or messenger RNA stability levels. Here, we summarize findings that implicate long non-coding RNAs in the control of mammalian cell differentiation. We focus on several representative differentiation systems and discuss how specific long non-coding RNAs contribute to the regulation of mammalian development.

  11. Verification of NUREC Code Transient Calculation Capability Using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Noh, Jae Man; Lee, Hyung Chul; Yoo, Jae Woon

    2006-01-15

    In this report, we verified the NUREC code transient calculation capability using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem. The benchmark problem consists of Part 1, a 2-D problem with given T/H conditions, Part 2, a 3-D problem at HFP condition, Part 3, a 3-D problem at HZP condition, and Part 4, a transient state initiated by a control rod ejection at HZP condition in Part 3. In Part 1, the results of NUREC code agreed well with the reference solution obtained from DeCART calculation except for the pin power distributions at the rodded assemblies. In Part 2, the results of NUREC code agreed well with the reference DeCART solutions. In Part 3, some results of NUREC code such as critical boron concentration and core averaged delayed neutron fraction agreed well with the reference PARCS 2G solutions. But the error of the assembly power at the core center was quite large. The pin power errors of NUREC code at the rodded assemblies was much smaller the those of PARCS code. The axial power distribution also agreed well with the reference solution. In Part 4, the results of NUREC code agreed well with those of PARCS 2G code which was taken as the reference solution. From the above results we can conclude that the results of NUREC code for steady states and transient states of the MOX loaded LWR core agree well with those of the other codes.

  12. Lessons learnt from application of the standardized cost calculation code OMEGA in decision making processes and planning in decommissioning

    International Nuclear Information System (INIS)

    Implementation of the standardised cost structure, as defined in 'A Proposed Standardised List of Costs Items for Decommissioning Purposes' (OECD/NEA, IAEA, EC, 1999), into the decommissioning costing, supports the harmonisation of decommissioning costs. The decision making processes in decommissioning planning can be more effective if there is the possibility to compare the calculated data with the data of other projects, structured in standardised cost structure. The results of the decision making process should be based on evaluation of such a set of decommissioning options which covers the methods of decommissioning, the selected strategy and existing or planned decommissioning infrastructure. Aspects such as impact of time, waste management scenarios, uncertainties of input data and other aspects should be also evaluated. These issues of decision making process were implemented into the decommissioning costing code OMEGA. All activities of a decommissioning project are involved within single compact standardised calculation structure including waste management. The resulting costs have standardised format and no additional data conversion is needed. The calculation process is nuclide resolved and internally linked in such a way that it models the material and radioactivity flow in the decommissioning process. The effect of decay of radioactivity is considered. The options are optimised in the standard MS-Project software as Gantt charts. The bi-directional data link between the standardised calculation structure and the Gantt chart supports the on-line optimisation of the Gantt chart structure. Multi-option work is applied, i.e. decommissioning options, which cover all decommissioning scenarios to be considered, are evaluated individually and multi-attribute analysis is applied for selecting the optimal one. Methods of sensitivity analysis and evaluation of uncertainties of calculated costs were developed for support the decision making process and for

  13. Construction of a computational exposure model for dosimetric calculations using the EGS4 Monte Carlo code and voxel phantoms

    International Nuclear Information System (INIS)

    The MAX phantom has been developed from existing segmented images of a male adult body, in order to achieve a representation as close as possible to the anatomical properties of the reference adult male specified by the ICRP. In computational dosimetry, MAX can simulate the geometry of a human body under exposure to ionizing radiations, internal or external, with the objective of calculating the equivalent dose in organs and tissues for occupational, medical or environmental purposes of the radiation protection. This study presents a methodology used to build a new computational exposure model MAX/EGS4: the geometric construction of the phantom; the development of the algorithm of one-directional, divergent, and isotropic radioactive sources; new methods for calculating the equivalent dose in the red bone marrow and in the skin, and the coupling of the MAX phantom with the EGS4 Monte Carlo code. Finally, some results of radiation protection, in the form of conversion coefficients between equivalent dose (or effective dose) and free air-kerma for external photon irradiation are presented and discussed. Comparing the results presented with similar data from other human phantoms it is possible to conclude that the coupling MAX/EGS4 is satisfactory for the calculation of the equivalent dose in radiation protection. (author)

  14. A computer code for forward calculation and inversion of the H/V spectral ratio under the diffuse field assumption

    CERN Document Server

    García-Jerez, Antonio; Sánchez-Sesma, Francisco J; Luzón, Francisco; Perton, Mathieu

    2016-01-01

    During a quarter of a century, the main characteristics of the horizontal-to-vertical spectral ratio of ambient noise HVSRN have been extensively used for site effect assessment. In spite of the uncertainties about the optimum theoretical model to describe these observations, several schemes for inversion of the full HVSRN curve for near surface surveying have been developed over the last decade. In this work, a computer code for forward calculation of H/V spectra based on the diffuse field assumption (DFA) is presented and tested.It takes advantage of the recently stated connection between the HVSRN and the elastodynamic Green's function which arises from the ambient noise interferometry theory. The algorithm allows for (1) a natural calculation of the Green's functions imaginary parts by using suitable contour integrals in the complex wavenumber plane, and (2) separate calculation of the contributions of Rayleigh, Love, P-SV and SH waves as well. The stability of the algorithm at high frequencies is preserv...

  15. Electron impact excitation of N IV: calculations with the DARC code and a comparison with ICFT results

    CERN Document Server

    Aggarwal, K M; Lawson, K D

    2016-01-01

    There have been discussions in the recent literature regarding the accuracy of the available electron impact excitation rates (equivalently effective collision strengths $\\Upsilon$) for transitions in Be-like ions. In the present paper we demonstrate, once again, that earlier results for $\\Upsilon$ are indeed overestimated (by up to four orders of magnitude), for over 40\\% of transitions and over a wide range of temperatures. To do this we have performed two sets of calculations for N~IV, with two different model sizes consisting of 166 and 238 fine-structure energy levels. As in our previous work, for the determination of atomic structure the GRASP (General-purpose Relativistic Atomic Structure Package) is adopted and for the scattering calculations (the standard and parallelised versions of) the Dirac Atomic R-matrix Code ({\\sc darc}) are employed. Calculations for collision strengths and effective collision strengths have been performed over a wide range of energy (up to 45~Ryd) and temperature (up to 2.0$...

  16. Electron impact excitation of N IV: calculations with the DARC code and a comparison with ICFT results

    Science.gov (United States)

    Aggarwal, K. M.; Keenan, F. P.; Lawson, K. D.

    2016-10-01

    There have been discussions in the recent literature regarding the accuracy of the available electron impact excitation rates (equivalently effective collision strengths Υ) for transitions in Be-like ions. In the present paper we demonstrate, once again, that earlier results for Υ are indeed overestimated (by up to four orders of magnitude), for over 40 per cent of transitions and over a wide range of temperatures. To do this we have performed two sets of calculations for N IV, with two different model sizes consisting of 166 and 238 fine-structure energy levels. As in our previous work, for the determination of atomic structure the GRASP (General-purpose Relativistic Atomic Structure Package) is adopted and for the scattering calculations (the standard and parallelised versions of) the Dirac Atomic R-matrix Code (DARC) are employed. Calculations for collision strengths and effective collision strengths have been performed over a wide range of energy (up to 45 Ryd) and temperature (up to 2.0 × 106 K), useful for applications in a variety of plasmas. Corresponding results for energy levels, lifetimes and A-values for all E1, E2, M1 and M2 transitions among 238 levels of N IV are also reported.

  17. AZIMUT code abstract

    International Nuclear Information System (INIS)

    The brief description of the AZIMUT code for calculation the neutron flux in a cluster cell is presented. Code takes into account 1 and 2 azimuthal harmonics in the one-group P3-approximation and uses the heterogeneous approach. 2 refs

  18. Codes and Standards Requirements for Deployment of Emerging Fuel Cell Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, R.; Buttner, W.; Riykin, C.

    2011-12-01

    The objective of this NREL report is to provide information on codes and standards (of two emerging hydrogen power fuel cell technology markets; forklift trucks and backup power units), that would ease the implementation of emerging fuel cell technologies. This information should help project developers, project engineers, code officials and other interested parties in developing and reviewing permit applications for regulatory compliance.

  19. Comparison study on cell calculation method of fast reactor

    International Nuclear Information System (INIS)

    Effective cross sections obtained by cell calculations are used in core calculations in current deterministic methods. Therefore, it is important to calculate the effective cross sections accurately and several methods have been proposed. In this study, some of the methods are compared to each other using a continuous energy Monte Carlo method as a reference. The result shows that the table look-up method used in Japan Nuclear Cycle Development Institute (JNC) sometimes has a difference over 10% in effective microscopic cross sections and be inferior to the sub-group method. The problem was overcome by introducing a new nuclear constant system developed in JNC, in which the ultra free energy group library is used. The system can also deal with resonance interaction effects between nuclides which are not able to be considered by other methods. In addition, a new method was proposed to calculate effective cross section accurately for power reactor fuel subassembly where the new nuclear constant system cannot be applied. This method uses the sub-group method and the ultra fine energy group collision probability method. The microscopic effective cross sections obtained by this method agree with the reference values within 5% difference. (author)

  20. Development of an integrated fission product release and transport code for spatially resolved full-core calculations of V/HTRs

    Energy Technology Data Exchange (ETDEWEB)

    Xhonneux, Andre, E-mail: a.xhonneux@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology RWTH-Aachen, 52064 Aachen (Germany); Allelein, Hans-Josef [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology RWTH-Aachen, 52064 Aachen (Germany)

    2014-05-01

    The computer codes FRESCO-I, FRESCO-II, PANAMA and SPATRA developed at Forschungszentrum Jülich in Germany in the early 1980s are essential tools to predict the fission product release from spherical fuel elements and the TRISO fuel performance, respectively, under given normal or accidental conditions. These codes are able to calculate a conservative estimation of the source term, i.e. quantity and duration of radionuclide release. Recently, these codes have been reversed engineered, modernized (FORTRAN 95/2003) and combined to form a consistent code named STACY (Source Term Analysis Code System). STACY will later become a module of the V/HTR Code Package (HCP). In addition, further improvements have been implemented to enable more detailed calculations. For example the distinct temperature profile along the pebble radius is now taken into account and coated particle failure rates can be calculated under normal operating conditions. In addition, the absolute fission product release of an V/HTR pebble bed core can be calculated by using the newly developed burnup code Topological Nuclide Transformation (TNT) replacing the former rudimentary approach. As a new functionality, spatially resolved fission product release calculations for normal operating conditions as well as accident conditions can be performed. In case of a full-core calculation, a large number of individual pebbles which follow a random path through the reactor core can be simulated. The history of the individual pebble is recorded, too. Main input data such as spatially resolved neutron fluxes and fluid dynamics data are provided by the VSOP code. Capabilities of the FRESCO-I and SPATRA code which allow for the simulation of the redistribution of fission products within the primary circuit and the deposition of fission products on graphitic and metallic surfaces are also available in STACY. In this paper, details of the STACY model and first results for its application to the 200 MW(th) HTR

  1. Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1

    International Nuclear Information System (INIS)

    This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50

  2. X-ray microbeam radiation therapy calculations, including polarisation effects, with the Monte Carlo code EGS5

    Energy Technology Data Exchange (ETDEWEB)

    Hugtenburg, Richard P., E-mail: r.p.hugtenburg@swansea.ac.u [School of Medicine, Swansea University, Swansea SA2 8PP (United Kingdom); Department of Medical Physics and Clinical Engineering, Abertawe Bro Morgannwg University, LHB, Swansea SA2 8QA (United Kingdom); Adegunloye, A.S.; Bradley, David A. [Department of Physics, Surrey University, Guildford (United Kingdom)

    2010-07-21

    Microbeam radiation therapy (MRT) is currently being considered for the treatment of glioblastoma multiforme. A high degree of dosimetric accuracy (around 5%) is known to be required for a successful outcome in conventional radiation therapy, Modelling of MRT beams, measurements and treatments have been performed with Monte Carlo methods using the code EGS5, which features improved physics models for low energy scattering processes including linear polarisation. Polarisation of the X-ray source leads to distortions in beam profiles that exceed the usual clinical tolerances. Changes in the energy spectrum also effect the response of many dosimetry systems. Anatomical (CT) data has been used in the dose calculations and the manipulation of dose data with the open-source software treatment planning system, PlanUNC, is demonstrated, in order that the therapeutic effects of the different components, e.g. the microbeam and scattered photons, can examined separately in relation to relevant anatomy.

  3. Calculations of neutron flux for BNCT facility of typical working core Multipurpose Reactor (RSG-GAS) using MCNP4B Code

    International Nuclear Information System (INIS)

    Calculation of neutron flux distributions of RSG-GAS typical working core using MCNP 4b Code has been done. Prior to the calculations, modelling of fuel element of meat as well as surfaces of cladding cell and geometry should be made. The model was then included water as a containment also developed. To achieve neutron flux behavior, it was simulated 200,000 to 2,000,000 neutrons. The calculation results indicated that the neutron flux in TWC core is in the order of 1014. Meanwhile, the best flux order for the BNCT facility should be in the order of 1010. With the use of any method, such as constructing of shielding and collimator, the order of neutron flux will decrease. In the previous research in 2001, the results showed the neutron flux in the order of 1010 by installing the collimator with 45 cm thick, made of Pb and 380 cm from the core centre. The results of this research completed with the research done in 2001, 2000 and 1999 certainly support the possibility to construct the BNCT facility in RSG-GAS reactor core

  4. Three-dimensional whole-core transport calculation of the OECD benchmark problem C5G7 MOX by the CRX code

    Energy Technology Data Exchange (ETDEWEB)

    Gil, Soo Lee; Nam, Zin Cho [Korea Advanced Institute of Science and Technology, Dept. of Nuclear and Quantum Engineering, Yusong-gu, Daejeon (Korea, Republic of)

    2005-07-01

    The OECD 3-dimensional benchmark problem C5G7 MOX was calculated by the CRX code. For 3-dimensional heterogeneous calculation, the CRX code uses a fusion technique of 2-dimensional/1-dimensional methods: the method of characteristics for radial 2-dimensional calculation and diamond difference scheme (DD) that is an S{sub N}-like method for axial 1-dimensional calculation. We improve the fusion method by using a linear characteristics (LC) solver in the 1-dimensional calculation. Here, we present brief structure of 2-dimensional/1-dimensional fusion method and the results of 3 configurations of benchmark problem. We also present results of several different 1-dimensional calculation options. Numerical results show that the LC scheme presents better performance than DD. In the results of the benchmark problem, k(eff) errors are less than 0.05% and the averages of pin power errors are less than 1% for all calculations.

  5. Code Betal to calculation Alpha/Beta activities in environmental samples; Programa de ordenador Betal para el calculo de la actividad Beta/Alfa de muestras ambientales

    Energy Technology Data Exchange (ETDEWEB)

    Romero, L.; Travesi, A.

    1983-07-01

    A codes, BETAL, was developed, written in FORTRAN IV, to automatize calculations and presentations of the result of the total alpha-beta activities measurements in environmental samples. This code performs the necessary calculations for transformation the activities measured in total counts, to pCi/1., bearing in mind the efficiency of the detector used and the other necessary parameters. Further more, it appraise the standard deviation of the result, and calculus the Lower limit of detection for each measurement. This code is written in iterative way by screen-operator dialogue, and asking the necessary data to perform the calculation of the activity in each case by a screen label. The code could be executed through any screen and keyboard terminal, (whose computer accepts Fortran IV) with a printer connected to the said computer. (Author) 5 refs.

  6. Monte-Carlo approach to calculate the ionization of warm dense matter within particle-in-cell simulations

    CERN Document Server

    Wu, D; Yu, W; Fritzsche, S

    2016-01-01

    A physical model based on Monte-Carlo approach is proposed to calculate the ionization dynamics of warm dense matters within particle-in-cell simulations, where impact ionization, electron-ion recombination and ionization potential depression (IPD) by surrounding plasmas are taken into consideration self-consistently. When compared with other models, which are applied in the literature for plasmas near thermal equilibrium, the temporal relaxation of ionizations can also be simulated by the proposed model with the final thermal equilibrium determined by the competition between impact ionization and its inverse process, i.e., electron-ion recombination. Our model is general and can be applied for both single elements and alloys with quite different compositions. The proposed model is implemented into a particle-in-cell (PIC) simulation code, and the average ionization degree of bulk aluminium varying with temperature is calculated, showing good agreement with the data provided by FLYCHK code.

  7. Integration of the DRAGON5/DONJON5 codes in the SALOME platform for performing multi-physics calculations in nuclear engineering

    Science.gov (United States)

    Hébert, Alain

    2014-06-01

    We are presenting the computer science techniques involved in the integration of codes DRAGON5 and DONJON5 in the SALOME platform. This integration brings new capabilities in designing multi-physics computational schemes, with the possibility to couple our reactor physics codes with thermal-hydraulics or thermo-mechanics codes from other organizations. A demonstration is presented where two code components are coupled using the YACS module of SALOME, based on the CORBA protocol. The first component is a full-core 3D steady-state neuronic calculation in a PWR performed using DONJON5. The second component implement a set of 1D thermal-hydraulics calculations, each performed over a single assembly.

  8. Adaptation of penelope Monte Carlo code system to the absorbed dose metrology: characterization of high energy photon beams and calculations of reference dosimeter correction factors; Adaptation du code Monte Carlo penelope pour la metrologie de la dose absorbee: caracterisation des faisceaux de photons X de haute energie et calcul de facteurs de correction de dosimetres de reference

    Energy Technology Data Exchange (ETDEWEB)

    Mazurier, J

    1999-05-28

    This thesis has been performed in the framework of national reference setting-up for absorbed dose in water and high energy photon beam provided with the SATURNE-43 medical accelerator of the BNM-LPRI (acronym for National Bureau of Metrology and Primary standard laboratory of ionising radiation). The aim of this work has been to develop and validate different user codes, based on PENELOPE Monte Carlo code system, to determine the photon beam characteristics and calculate the correction factors of reference dosimeters such as Fricke dosimeters and graphite calorimeter. In the first step, the developed user codes have permitted the influence study of different components constituting the irradiation head. Variance reduction techniques have been used to reduce the calculation time. The phase space has been calculated for 6, 12 and 25 MV at the output surface level of the accelerator head, then used for calculating energy spectra and dose distributions in the reference water phantom. Results obtained have been compared with experimental measurements. The second step has been devoted to develop an user code allowing calculation correction factors associated with both BNM-LPRI's graphite and Fricke dosimeters thanks to a correlated sampling method starting with energy spectra obtained in the first step. Then the calculated correction factors have been compared with experimental and calculated results obtained with the Monte Carlo EGS4 code system. The good agreement, between experimental and calculated results, leads to validate simulations performed with the PENELOPE code system. (author)

  9. GORGON - a computer code for the calculation of energy deposition and the slowing down of ions in cold materials and hot dense plasmas

    International Nuclear Information System (INIS)

    The computer code GORGON, which calculates the energy deposition and slowing down of ions in cold materials and hot plasmas is described, and analyzed in this report. This code is in a state of continuous development but an intermediate stage has been reached where it is considered useful to document the 'state of the art' at the present time. The GORGON code is an improved version of a code developed by Zinamon et al. as part of a more complex program system for studying the hydrodynamic motion of plane metal targets irradiated by intense beams of protons. The improvements made in the code were necessary to improve its usefulness for problems related to the design and burn of heavy ion beam driven inertial confinement fusion targets. (orig./GG)

  10. Calculation of the fuel composition and the thermo-neutronic parameters of the Bushehr’s VVER-1000 reactor during the initial startup and the first cycle using the WIMSD5-B, CITATION-LDI2 and WERL codes

    International Nuclear Information System (INIS)

    Highlights: ► In this paper, the changes of the thermo-neutronic parameters of a VVER 1000 reactor were studied during the first cycle. ► The coupling of neutronic and thermo-hydraulic codes was utilized. ► A computational program (WERL code) was designed to calculate the temperature distribution of the reactor core. ► To estimate the concentration of the released gaseous fission products, the Weisman model was used. ► The results of this study enjoyed the desirable accuracy. - Abstract: In this paper, the concentrations of fission products and fuel isotopes as well as the changes of the thermo-neutronic parameters of the Bushehr’s VVER-1000 reactor were studied during the initial startup and the first cycle. In order to perform the time-dependent cell calculations and obtain the concentration of fuel elements, the WIMSD5-B code was used. Besides, by utilizing the CITATION-LDI2 code, the effective multiplication factor and the thermal power distribution of the reactor were calculated. A computer program (WERL code) was designed in order to perform accurate calculation of the temperature distribution of the reactor core. For this purpose, the Ross–Stoute, Weisman, and Lee–Kesler models were used for calculating of the gap conductance coefficient, fission gas release and gap pressure, respectively. The results demonstrated that in designing the startup process, in addition to the role considered for overcoming the power defects and in preparing the required conditions for performing the safety-assurance tests, the flattening of the reactor’s power must be taken into account. Comparison between the results of this modeling and the final safety analysis report of this reactor showed that the results presented in this paper are satisfactorily accurate

  11. Coded illumination for motion-blur free imaging of cells on cell-phone based imaging flow cytometer

    Science.gov (United States)

    Saxena, Manish; Gorthi, Sai Siva

    2014-10-01

    Cell-phone based imaging flow cytometry can be realized by flowing cells through the microfluidic devices, and capturing their images with an optically enhanced camera of the cell-phone. Throughput in flow cytometers is usually enhanced by increasing the flow rate of cells. However, maximum frame rate of camera system limits the achievable flow rate. Beyond this, the images become highly blurred due to motion-smear. We propose to address this issue with coded illumination, which enables recovery of high-fidelity images of cells far beyond their motion-blur limit. This paper presents simulation results of deblurring the synthetically generated cell/bead images under such coded illumination.

  12. Re-evaluation of Assay Data of Spent Nuclear Fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

    International Nuclear Information System (INIS)

    Highlights: → The specifications required for the analyses of the destructive assay data taken from irradiated fuel in Ohi-1 and Ohi-2 PWRs were documented in this paper. → These data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. → These destructive assay data are suitable for the benchmarking of the burnup calculation code systems. - Abstract: The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, Assay Data of Spent Nuclear Fuel from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The destructive assay data from Ohi-2 have already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of destructive assay data and specifications required for analyses of isotopic composition from Ohi-1 and Ohi-2. For precise burnup analyses, the burnup values of destructive assay samples were re-evaluated in this study. These destructive assay data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of destructive assay data from Ohi-1 and Ohi-2 PWRs is high, and that these destructive assay data are suitable for the benchmarking of burnup calculation code systems.

  13. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  14. A general concurrent algorithm for plasma particle-in-cell simulation codes

    Science.gov (United States)

    Liewer, Paulett C.; Decyk, Viktor K.

    1989-01-01

    The general concurrent particle-in-cell (GCPIC) algorithm has been used to implement an electrostatic particle-in-cell code on a 32-node hypercube parallel computer. The GCPIC algorithm decomposes the PIC code by dividing the particle simulation physical domain into subdomains that are equal in number to the number of processors; all subdomains will accordingly possess approximately equal numbers of particles. The portion of the code which updates particle positions and velocities is nearly 100 percent efficient when the number of particles increases linearly with that of hypercube processors.

  15. User's manual to the ICRP Code: a series of computer programs to perform dosimetric calculations for the ICRP Committee 2 report

    Energy Technology Data Exchange (ETDEWEB)

    Watson, S.B.; Ford, M.R.

    1980-02-01

    A computer code has been developed that implements the recommendations of ICRP Committee 2 for computing limits for occupational exposure of radionuclides. The purpose of this report is to describe the various modules of the computer code and to present a description of the methods and criteria used to compute the tables published in the Committee 2 report. The computer code contains three modules of which: (1) one computes specific effective energy; (2) one calculates cumulated activity; and (3) one computes dose and the series of ICRP tables. The description of the first two modules emphasizes the new ICRP Committee 2 recommendations in computing specific effective energy and cumulated activity. For the third module, the complex criteria are discussed for calculating the tables of committed dose equivalent, weighted committed dose equivalents, annual limit of intake, and derived air concentration.

  16. User's manual to the ICRP Code: a series of computer programs to perform dosimetric calculations for the ICRP Committee 2 report

    International Nuclear Information System (INIS)

    A computer code has been developed that implements the recommendations of ICRP Committee 2 for computing limits for occupational exposure of radionuclides. The purpose of this report is to describe the various modules of the computer code and to present a description of the methods and criteria used to compute the tables published in the Committee 2 report. The computer code contains three modules of which: (1) one computes specific effective energy; (2) one calculates cumulated activity; and (3) one computes dose and the series of ICRP tables. The description of the first two modules emphasizes the new ICRP Committee 2 recommendations in computing specific effective energy and cumulated activity. For the third module, the complex criteria are discussed for calculating the tables of committed dose equivalent, weighted committed dose equivalents, annual limit of intake, and derived air concentration

  17. Computer code TERFOC-N to calculate doses to public using terrestrial foodchain models improved and extended for long-lived nuclides

    International Nuclear Information System (INIS)

    A computer code TERFOC-N has bee developed to calculate doses to the public due to atmospheric releases of radionuclides in normal operations of nuclear facilities. The code calculates the highest individual dose and the collective dose from four exposure pathways; internal doses due to ingestion and inhalation, external doses due to cloudshine and groundshine. A foodchain model, which is originally referred to the U.S.NRC Regulatory Guide 1.109, has been improved to apply to not only LWRs but also other nuclear facilities. This report describes the models employed and gives a sample run performed by the code. The parameters which were sensitive to ingestion dose were identified from the results of sensitivity analysis. The models which significantly contributed to the dose were identified among the models improved and extended here. (author)

  18. Calculation of absorbed dose for skin contamination imparted by beta radiation through the VARSKIN modified code for 122 interesting isotopes for nuclear medicine, nuclear power plants and research

    International Nuclear Information System (INIS)

    In this work the implementation of a modification of the VARSKIN code for calculation of absorbed dose for contamination in skin imparted by external radiation fields generated by Beta emitting is presented. The modification consists on the inclusion of 47 isotopes of interest even Nuclear Plants for the dose evaluation in skin generated by 'hot particles'. The approach for to add these isotopes is the correlation parameter F and the average energy of the Beta particle, with relationship to those 75 isotopes of the original code. The methodology of the dose calculation of the VARSKIN code is based on the interpolation, (and integration of the interest geometries: punctual or plane sources), of the distribution functions scaled doses in water for beta and electrons punctual sources, tabulated by Berger. Finally a brief discussion of the results for their interpretation and use with purposes of radiological protection (dose insurance in relation to the considered biological effects) is presented

  19. The Monte Carlo method for shielding calculations analysis by MORSE code of a streaming case in the CAORSO BWR power reactor shielding (Italy)

    International Nuclear Information System (INIS)

    In the field of shielding, the requirement of radiation transport calculations in severe conditions, characterized by irreducible three-dimensional geometries has increased the use of the Monte Carlo method. The latter has proved to be the only rigorous and appropriate calculational method in such conditions. However, further efforts at optimization are still necessary to render the technique practically efficient, despite recent improvements in the Monte Carlo codes, the progress made in the field of computers and the availability of accurate nuclear data. Moreover, the personal experience acquired in the field and the control of sophisticated calculation procedures are of the utmost importance. The aim of the work which has been carried out is the gathering of all the necessary elements and features that would lead to an efficient utilization of the Monte Carlo method used in connection with shielding problems. The study of the general aspects of the method and the exploitation techniques of the MORSE code, which has proved to be one of the most comprehensive of the Monte Carlo codes, lead to a successful analysis of an actual case. In fact, the severe conditions and difficulties met have been overcome using such a stochastic simulation code. Finally, a critical comparison between calculated and high-accuracy experimental results has allowed the final confirmation of the methodology used by us

  20. PN/S calculations for a fighter W/F at high-lift yaw conditions. [parabolized Navier-Stokes computer code

    Science.gov (United States)

    Wai, J. C.; Blom, G.; Yoshihara, H.; Chaussee, D.

    1986-01-01

    The NASA/Ames parabolized Navier/Stokes computer code was used to calculate the turbulent flow over the wing/fuselage for a generic fighter at M = 2.2. 18 deg angle-of-attack, and 0 and 5 deg yaw. Good test/theory agreement was achieved in the zero yaw case. No test data were available for the yaw case.

  1. Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

    International Nuclear Information System (INIS)

    Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

  2. Object-Oriented Parallel Particle-in-Cell Code for Beam Dynamics Simulation in Linear Accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Qiang, J.; Ryne, R.D.; Habib, S.; Decky, V.

    1999-11-13

    In this paper, we present an object-oriented three-dimensional parallel particle-in-cell code for beam dynamics simulation in linear accelerators. A two-dimensional parallel domain decomposition approach is employed within a message passing programming paradigm along with a dynamic load balancing. Implementing object-oriented software design provides the code with better maintainability, reusability, and extensibility compared with conventional structure based code. This also helps to encapsulate the details of communications syntax. Performance tests on SGI/Cray T3E-900 and SGI Origin 2000 machines show good scalability of the object-oriented code. Some important features of this code also include employing symplectic integration with linear maps of external focusing elements and using z as the independent variable, typical in accelerators. A successful application was done to simulate beam transport through three superconducting sections in the APT linac design.

  3. SOLANG: A user-friendly code to calculate the geometry factor using Monte Carlo simulations. Application to alpha-particle spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Cornejo Diaz, N.A. [Centro de Proteccion e Higiene de las Radiaciones, C.P. 6195, La Habana (Cuba); Martin Sanchez, A., E-mail: ams@unex.e [Departamento de Fisica, Universidad de Extremadura, E-06071 Badajoz (Spain); Torre Perez, J. de la [Departamento de Fisica, Universidad de Extremadura, E-06071 Badajoz (Spain)

    2011-05-15

    Monte Carlo simulation was applied to calculate the effective solid angle (or geometry factor) presented by a plane radioactive source at a detector entrance window. A fast and user-friendly computer program SOLANG was written to perform the calculations for disk or rectangular sources and circular non-coaxial detector disks. Results can be achieved with great precision, depending on the number of simulated trajectories. Some checks and applications to the calculation of efficiencies of semiconductor detectors and gas ionization chambers used to measure alpha particles are presented. Their results were very reliable. The code is available free of charge on request to the authors.

  4. Verification and sensitivity of the calculational methods used in the PATHRAE code to predict subsurface contaminant transport for risk assessments of SRP waste sites

    Energy Technology Data Exchange (ETDEWEB)

    Fjeld, R.A.; Elzerman, A.W.; Overcamp, T.J.; Giannopoulos, N.; Crider, S.; Sill, B.L.

    1986-10-01

    Presented in this report are an independent verification of the subsurface contaminant transport calculations contained in the code and an assessment of the sensitivity of predicted contaminant concentrations to uncertainties in transport parameters. The subsurface transport approximation incorporated in the PATHRAE risk assessment code was compared with alternate two-dimensional and three-dimensional approximations and with the EPA VHS model. Agreement between the PATHRAE approximation and the alternate two-dimensional approximation was good. Due to its neglect of vertical dispersion, the PATHRAE model predicted higher groundwater (undiluted) concentrations than the three-dimensional approximation and, for EPA parameters, the VHS model. The use of a value of zero for horizontal dispersivity, as specified for 1 m and 100 m wells in SPR waste site analyses, was found to add an additional degree of conservatism to PATHRAE estimates of groundwater concentration, yielding levels that were more than three orders of magnitude higher than those of the three-dimensional model for a 100 m well. Implementation of the transport approximation in the PATHRAE code was verified by comparing code generated concentrations with those of an independent calculation for wide ranges of the input parameters. Agreement between PATHRAE and the independent calculations was excellent.

  5. Post-test calculation of thermal-hydraulic behaviour in Demona experiment B3 with various computer codes used in EC Member States

    International Nuclear Information System (INIS)

    In 1986 the CEC sponsored a benchmark exercise on aerosol calculation based on the Demona B3 experiment. The results of this exercise were very sensitive to the calculation of energy and mass transfer between the phases. In view of the results of the study mentioned above, it had been decided to carry out a benchmark exercise for severe accident containment thermal-hydraulics codes. This exercise is based on experiment B3 in the Demona programme. The experiment B3 was a simulation of a late overpressure failure scenario in a PWR. The main objective of the benchmark exercise has been to assess the ability of the participating codes to predict atmosphere saturation levels and bulk condensation rates under conditions similar to those predicted to follow a severe accident in a PWR. Several research organizations of the Community member countries have participated at this benchmark exercise. The paper presents the comparison of the experimental results with the following calculated quantities: total system pressure, steam pressure, containment temperature, saturation ratio, steam condensation rate on walls and in the bulk volume, water mass in sump, sump temperature, heat transfer coefficient, steam mass flow rate and structural temperature for a real time period of up to 90 hours. The paper also presents the major conclusions from the exercise in order to identify the status of present codes versus the requirements needed as input and for coupling with aerosol analysis codes

  6. Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

    International Nuclear Information System (INIS)

    All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important

  7. TRANGE: computer code to calculate the energy beam degradation in target stack; TRANGE: programa para calcular a degradacao de energia de particulas carregadas em alvos

    Energy Technology Data Exchange (ETDEWEB)

    Bellido, Luis F.

    1995-07-01

    A computer code to calculate the projectile energy degradation along a target stack was developed for an IBM or compatible personal microcomputer. A comparison of protons and deuterons bombarding uranium and aluminium targets was made. The results showed that the data obtained with TRANGE were in agreement with other computers code such as TRIM, EDP and also using Williamsom and Janni range and stopping power tables. TRANGE can be used for any charged particle ion, for energies between 1 to 100 MeV, in metal foils and solid compounds targets. (author). 8 refs., 2 tabs.

  8. Validation study of SRAC2006 code system based on evaluated nuclear data libraries for TRIGA calculations by benchmarking integral parameters of TRX and BAPL lattices of thermal reactors

    International Nuclear Information System (INIS)

    Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations of

  9. Benchmarking of calculated projectile fragmentation cross-sections using the 3-D, MC codes PHITS, FLUKA, HETC-HEDS, MCNPX_HI, and NUCFRG2

    Science.gov (United States)

    Sihver, L.; Mancusi, D.; Niita, K.; Sato, T.; Townsend, L.; Farmer, C.; Pinsky, L.; Ferrari, A.; Cerutti, F.; Gomes, I.

    Particles and heavy ions are used in various fields of nuclear physics, medical physics, and material science, and their interactions with different media, including human tissue and critical organs, have therefore carefully been investigated both experimentally and theoretically since the 1930s. However, heavy-ion transport includes many complex processes and measurements for all possible systems, including critical organs, would be impractical or too expensive; e.g. direct measurements of dose equivalents to critical organs in humans cannot be performed. A reliable and accurate particle and heavy-ion transport code is therefore an essential tool in the design study of accelerator facilities as well as for other various applications. Recently, new applications have also arisen within transmutation and reactor science, space and medicine, especially radiotherapy, and several accelerator facilities are operating or planned for construction. Accurate knowledge of the physics of interaction of particles and heavy ions is also necessary for estimating radiation damage to equipment used on space vehicles, to calculate the transport of the heavy ions in the galactic cosmic ray (GCR) through the interstellar medium, and the evolution of the heavier elements after the Big Bang. Concerns about the biological effect of space radiation and space dosimetry are increasing rapidly due to the perspective of long-duration astronaut missions, both in relation to the International Space Station and to manned interplanetary missions in near future. Radiation protection studies for crews of international flights at high altitude have also received considerable attention in recent years. There is therefore a need to develop accurate and reliable particle and heavy-ion transport codes. To be able to calculate complex geometries, including production and transport of protons, neutrons, and alpha particles, 3-dimensional transport using Monte Carlo (MC) technique must be used. Today

  10. Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh; Calculs de reference avec un maillage multigroupe fin sur des assemblages critiques par Apollo2

    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)

  11. SYSMOD: user-interface for data processing, calculation codes and analysis of PWR lattices; SYSMOD: una interfase-usuario para el procesamiento, calculo y analysis de redes PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, Alejandro; Milian, Daniel [Instituto Superior de Ciencias y Tecnologias Nucleares (ISCTN), La Habana (Cuba). E-mail: agg@ctn.isctn.edu.cu

    2000-07-01

    The task of the physical calculation of the reactor demand of the management of a great volume of information and inclose the stages for processing of data, calculations and analysis of their results. These stages are highly sensible to human mistakes, that's why is imprescindible that them undergo automatization, doing tracked all the process against mistake or unexpected result. The user-interface SYSMOD was developed over the platform IDE Delphi 3.0, visual language driven to events. It to consist in of the principal menu, which inclose between its options the preparation of the input data (File and Edit) to the pre-processors for the calculation codes of reactors. The output information may be showed in graphic and/or alphanumeric format (Data-Process). SYSMOD endures two applications for the management of the data base for the data during the preparation of the input for the pre-processors of the spectral calculation, so as for the organization, conservation and presentation for the obtained results. The carried out of the lattices and global codes, takes place from this application, over the platform MS-DOS (Run). SYSMOD regards the possibility for the debugging of the codes (Debugging), so as the benchmarks qualified to so effect (Benchmark). SYSMOD has been applied for the analysis of te WWER-440 of the first unity of Juragua Nuclear Power Plant. (author)

  12. IR approximation for calculating sensitivity and uncertainty of PWR cells by taking account of self-shielding effect

    International Nuclear Information System (INIS)

    A new improved method has been developed for calculating sensitivity coefficients of neutronics parameters in pressurized water reactor cells relative to infinite dilution cross-sections by taking account of resonance self-shielding effect. In our paper, the IR approximation is used in order to get accurate results in both high and low energy groups. This method is applied to UO2 and MOX fueled PWR cells to calculate sensitivity coefficients and uncertainties of eigenvalue responses. We have verified the improved method by comparing the sensitivities with MCNP code and good agreement is found. For uncertainty, the improved results are compared with TSUNAMI-1D, and demonstrate that the differences are caused by the use of different covariance matrix. (author)

  13. URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

    International Nuclear Information System (INIS)

    The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling by Doppler broadened cross-sections. The various self-shielding factors are computer numerically as Lebesgue integrals over the cross-section probability tables

  14. GPU Acceleration of the Locally Selfconsistent Multiple Scattering Code for First Principles Calculation of the Ground State and Statistical Physics of Materials

    Energy Technology Data Exchange (ETDEWEB)

    Eisenbach, Markus [ORNL; Larkin, Jeff [NVIDIA, Santa Clara, CA; Lutjens, Justin [NVIDIA, Santa Clara, CA; Rennich, Steven [NVIDIA, Santa Clara, CA; Rogers, James H [ORNL

    2016-01-01

    The Locally Self-consistent Multiple Scattering (LSMS) code solves the first principles Density Functional theory Kohn-Sham equation for a wide range of materials with a special focus on metals, alloys and metallic nano-structures. It has traditionally exhibited near perfect scalability on massively parallel high performance computer architectures. We present our efforts to exploit GPUs to accelerate the LSMS code to enable first principles calculations of O(100,000) atoms and statistical physics sampling of finite temperature properties. Using the Cray XK7 system Titan at the Oak Ridge Leadership Computing Facility we achieve a sustained performance of 14.5PFlop/s and a speedup of 8.6 compared to the CPU only code.

  15. Particle in cell calculation of plasma force on a small grain in a non-uniform collisional sheath

    CERN Document Server

    Hutchinson, I H

    2013-01-01

    The plasma force on grains of specified charge and height in a collisional plasma sheath are calculated using the multidimensional particle in cell code COPTIC. The background ion velocity distribution functions for the unperturbed sheath vary substantially with collisionality. The grain force is found to agree quite well with a combination of background electric field force plus ion drag force. However, the drag force must take account of the non-Maxwellian (and spatially varying) ion distribution function, and the collisional drag enhancement. It is shown how to translate the dimensionless results into practical equilibrium including other forces such as gravity.

  16. Monte-Carlo approach to calculate the proton stopping in warm dense matter within particle-in-cell simulations

    OpenAIRE

    Wu, D; X. T. He; Yu, W.; Fritzsche, S.

    2016-01-01

    A Monte-Carlo approach to proton stopping in warm dense matter is implemented into an existing particle-in-cell code. The model is based on multiple binary-collisions among electron-electron, electron-ion and ion-ion, taking into account contributions from both free and bound electrons, and allows to calculate particle stopping in much more natural manner. At low temperature limit, when ``all'' electron are bounded at the nucleus, the stopping power converges to the predictions of Bethe-Bloch...

  17. Particle-in-Cell Codes for plasma-based particle acceleration

    CERN Document Server

    Pukhov, Alexander

    2016-01-01

    Basic principles of particle-in-cell (PIC ) codes with the main application for plasma-based acceleration are discussed. The ab initio full electromagnetic relativistic PIC codes provide the most reliable description of plasmas. Their properties are considered in detail. Representing the most fundamental model, the full PIC codes are computationally expensive. The plasma-based acceler- ation is a multi-scale problem with very disparate scales. The smallest scale is the laser or plasma wavelength (from one to hundred microns) and the largest scale is the acceleration distance (from a few centimeters to meters or even kilometers). The Lorentz-boost technique allows to reduce the scale disparity at the costs of complicating the simulations and causing unphysical numerical instabilities in the code. Another possibility is to use the quasi-static approxi- mation where the disparate scales are separated analytically.

  18. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1997-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  19. Criticality calculations on pebble-bed HTR-PROTEUS configuration as a validation for the pseudo-scattering tracking method implemented in the MORET 5 Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Forestier, Benoit; Miss, Joachim; Bernard, Franck; Dorval, Aurelien [Institut de Radioprotection et Surete Nucleaire, Fontenay aux Roses (France); Jacquet, Olivier [Independent consultant (France); Verboomen, Bernard [Belgian Nuclear Research Center - SCK-CEN (Belgium)

    2008-07-01

    The MORET code is a three dimensional Monte Carlo criticality code. It is designed to calculate the effective multiplication factor (k{sub eff}) of any geometrical configuration as well as the reaction rates in the various volumes and the neutron leakage out of the system. A recent development for the MORET code consists of the implementation of an alternate neutron tracking method, known as the pseudo-scattering tracking method. This method has been successfully implemented in the MORET code and its performances have been tested by mean of an extensive parametric study on very simple geometrical configurations. In this context, the goal of the present work is to validate the pseudo-scattering method against realistic configurations. In this perspective, pebble-bed cores are particularly well-adapted cases to model, as they exhibit large amount of volumes stochastically arranged on two different levels (the pebbles in the core and the TRISO particles inside each pebble). This paper will introduce the techniques and methods used to model pebble-bed cores in a realistic way. The results of the criticality calculations, as well as the pseudo-scattering tracking method performance in terms of computation time, will also be presented. (authors)

  20. A New Method for Calculating Counts in Cells

    CERN Document Server

    Szapudi, I

    1997-01-01

    In the near future a new generation of CCD based galaxy surveys will enable high precision determination of the N-point correlation functions. The resulting information will help to resolve the ambiguities associated with two-point correlation functions thus constraining theories of structure formation, biasing, and Gaussianity of initial conditions independently of the value of $\\Omega$. As one the most successful methods to extract the amplitude of higher order correlations is based on measuring the distribution of counts in cells, this work presents an advanced way of measuring it with unprecedented accuracy. Szapudi and Colombi (1996, hereafter \\cite{sc96}) identified the main sources of theoretical errors in extracting counts in cells from galaxy catalogs. One of these sources, termed as measurement error, stems from the fact that conventional methods use a finite number of sampling cells to estimate counts in cells. This effect can be circumvented by using an infinite number of cells. This paper present...

  1. photon-plasma: A modern high-order particle-in-cell code

    Energy Technology Data Exchange (ETDEWEB)

    Haugbølle, Troels [Centre for Star and Planet Formation, Natural History Museum of Denmark, University of Copenhagen, Øster Voldgade 5-7, DK-1350 Copenhagen (Denmark); Frederiksen, Jacob Trier [Niels Bohr Institute, University of Copenhagen, Juliane Maries Vej 30, DK-2100 Copenhagen (Denmark); Nordlund, Åke [Niels Bohr Institute, University of Copenhagen, Juliane Maries Vej 30, DK-2100 Copenhagen (Denmark); Centre for Star and Planet Formation, Natural History Museum of Denmark, University of Copenhagen, Øster Voldgade 5-7, DK-1350 Copenhagen (Denmark)

    2013-06-15

    We present the photon-plasma code, a modern high order charge conserving particle-in-cell code for simulating relativistic plasmas. The code is using a high order implicit field solver and a novel high order charge conserving interpolation scheme for particle-to-cell interpolation and charge deposition. It includes powerful diagnostics tools with on-the-fly particle tracking, synthetic spectra integration, 2D volume slicing, and a new method to correctly account for radiative cooling in the simulations. A robust technique for imposing (time-dependent) particle and field fluxes on the boundaries is also presented. Using a hybrid OpenMP and MPI approach, the code scales efficiently from 8 to more than 250.000 cores with almost linear weak scaling on a range of architectures. The code is tested with the classical benchmarks particle heating, cold beam instability, and two-stream instability. We also present particle-in-cell simulations of the Kelvin-Helmholtz instability, and new results on radiative collisionless shocks.

  2. photon-plasma: A modern high-order particle-in-cell code

    International Nuclear Information System (INIS)

    We present the photon-plasma code, a modern high order charge conserving particle-in-cell code for simulating relativistic plasmas. The code is using a high order implicit field solver and a novel high order charge conserving interpolation scheme for particle-to-cell interpolation and charge deposition. It includes powerful diagnostics tools with on-the-fly particle tracking, synthetic spectra integration, 2D volume slicing, and a new method to correctly account for radiative cooling in the simulations. A robust technique for imposing (time-dependent) particle and field fluxes on the boundaries is also presented. Using a hybrid OpenMP and MPI approach, the code scales efficiently from 8 to more than 250.000 cores with almost linear weak scaling on a range of architectures. The code is tested with the classical benchmarks particle heating, cold beam instability, and two-stream instability. We also present particle-in-cell simulations of the Kelvin-Helmholtz instability, and new results on radiative collisionless shocks

  3. The fourth research co-ordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors reactivity effects'. Working material

    International Nuclear Information System (INIS)

    The fourth Research Co-ordination Meeting (RCM) of the Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effect' was held during 19-23 May, 2003 in Obninsk, Russian Federation. The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The first RCM took place in Vienna on 24 - 26 November 1999. The meeting was attended by 19 participants from 7 Member States and one from an international organization (France, Germany, India, Japan, Rep. of Korea, Russian Federation, the United Kingdom, and IAEA). The participants from two Member States (China and the U.S.A.) provided their results and presentation materials even though being absent at the meeting. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN- 600 core were evaluated. Contributions of the participants in the benchmark analyses is shown. This report first addresses the benchmark definitions and specifications given for each Phase and briefly introduces the basic data, computer codes, and methodologies applied to the benchmark analyses by various participants. Then, the results obtained by the participants in terms of calculational uncertainty and their effect on the core transient behavior are intercompared. Finally it addresses some conclusions drawn in the benchmarks

  4. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

    OpenAIRE

    Gholamzadeh Zohreh; Hossein Feghhi Seyed Amir; Soltani Leila; Rezazadeh Marzieh; Tenreiro Claudio; Joharifard Mahdi

    2014-01-01

    Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. N...

  5. Pre-test calculation of reflooding experiments with wider lattice in APWR-geometry (FLORESTAN 2) using the advanced computer code FLUT-FDWR

    International Nuclear Information System (INIS)

    After the reflooding tests in an extremely tight bundle (p/d=1.06, FLORESTAN 1) have been completed, new experiments for a wider lattice (p/d=1.242, FLORESTAN 2), which is employed in the recent APWR design of KfK, are planned at KfK to obtain the benchmark data for validation and improvement of calculation methods. This report presents the results of pre-test calculations for the FLORESTAN 2 experiment using FLUT-FDWR, a modified version of the GRS computer code FLUT for analysis of the most important behaviour during the reflooding phase after a LOCA in the APWR design. (orig.)

  6. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 2

    International Nuclear Information System (INIS)

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP)

  7. Development of the neutron-transport code TransRay and studies on the two- and three-dimensional calculation of effective group cross sections; Entwicklung des Neutronentransportcodes TransRay und Untersuchungen zur zwei- und dreidimensionalen Berechnung effektiver Gruppenwirkungsquerschnitte

    Energy Technology Data Exchange (ETDEWEB)

    Beckert, C.

    2007-12-19

    Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed. [German] Standardmaessig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte fuer

  8. Relay-aided multi-cell broadcasting with random network coding

    DEFF Research Database (Denmark)

    Lu, Lu; Sun, Fan; Xiao, Ming;

    2010-01-01

    We investigate a relay-aided multi-cell broadcasting system using random network codes, where the focus is on devising efficient scheduling algorithms between relay and base stations. Two scheduling algorithms are proposed based on different feedback strategies; namely, a one-step scheduling...

  9. Recent progress in linear-scaling density functional calculations with plane waves and pseudopotentials: the ONETEP code

    International Nuclear Information System (INIS)

    The ONETEP program employs the single-particle density matrix reformulation of Kohn-Sham density functional theory to achieve computational cost and memory requirements which increase only linearly with the number of atoms. As the code employs a plane wave basis set (in the form of periodic sinc functions) and pseudopotentials it is able to achieve levels of accuracy and systematic improvability comparable to those of conventional cubic-scaling plane wave approaches. The code has been developed with the aim of running efficiently on a variety of parallel architectures ranging from commodity clusters with tens of processors to large national facilities with thousands of processors. Recent and ongoing studies which we are performing with ONETEP involve problems ranging from materials to biomolecules to nanostructures

  10. Quality control of the treatment planning systems dose calculations in external radiation therapy using the Penelope Monte Carlo code; Controle qualite des systemes de planification dosimetrique des traitements en radiotherapie externe au moyen du code Monte-Carlo Penelope

    Energy Technology Data Exchange (ETDEWEB)

    Blazy-Aubignac, L

    2007-09-15

    The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)

  11. FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

    International Nuclear Information System (INIS)

    In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs

  12. Comparison of depth-dose distributions of proton therapeutic beams calculated by means of logical detectors and ionization chamber modeled in Monte Carlo codes

    Science.gov (United States)

    Pietrzak, Robert; Konefał, Adam; Sokół, Maria; Orlef, Andrzej

    2016-08-01

    The success of proton therapy depends strongly on the precision of treatment planning. Dose distribution in biological tissue may be obtained from Monte Carlo simulations using various scientific codes making it possible to perform very accurate calculations. However, there are many factors affecting the accuracy of modeling. One of them is a structure of objects called bins registering a dose. In this work the influence of bin structure on the dose distributions was examined. The MCNPX code calculations of Bragg curve for the 60 MeV proton beam were done in two ways: using simple logical detectors being the volumes determined in water, and using a precise model of ionization chamber used in clinical dosimetry. The results of the simulations were verified experimentally in the water phantom with Marcus ionization chamber. The average local dose difference between the measured relative doses in the water phantom and those calculated by means of the logical detectors was 1.4% at first 25 mm, whereas in the full depth range this difference was 1.6% for the maximum uncertainty in the calculations less than 2.4% and for the maximum measuring error of 1%. In case of the relative doses calculated with the use of the ionization chamber model this average difference was somewhat greater, being 2.3% at depths up to 25 mm and 2.4% in the full range of depths for the maximum uncertainty in the calculations of 3%. In the dose calculations the ionization chamber model does not offer any additional advantages over the logical detectors. The results provided by both models are similar and in good agreement with the measurements, however, the logical detector approach is a more time-effective method.

  13. Role of long non-coding RNAs in the determination of β-cell identity.

    Science.gov (United States)

    Motterle, A; Sanchez-Parra, C; Regazzi, R

    2016-09-01

    Pancreatic β-cells are highly specialized cells committed to secrete insulin in response to changes in the level of nutrients, hormones and neurotransmitters. Chronic exposure to elevated concentrations of glucose, fatty acids or inflammatory mediators can result in modifications in β-cell gene expression that alter their functional properties. This can lead to the release of insufficient amount of insulin to cover the organism's needs, and thus to the development of diabetes mellitus. Although most of the studies carried out in the last decades to elucidate the causes of β-cell dysfunction under disease conditions have focused on protein-coding genes, we now know that insulin-secreting cells also contain thousands of molecules of RNA that do not encode polypeptides but play key roles in the acquisition and maintenance of a highly differentiated state. In this review, we will highlight the involvement of long non-coding RNAs (lncRNAs), a particular class of non-coding transcripts, in the differentiation of β-cells and in the regulation of their specialized tasks. We will also discuss the crosstalk between the activities of lncRNAs and microRNAs and present the emerging evidence of a potential contribution of particular lncRNAs to the development of both type 1 and type 2 diabetes. PMID:27615130

  14. ANGIOGENES: knowledge database for protein-coding and noncoding RNA genes in endothelial cells.

    Science.gov (United States)

    Müller, Raphael; Weirick, Tyler; John, David; Militello, Giuseppe; Chen, Wei; Dimmeler, Stefanie; Uchida, Shizuka

    2016-09-01

    Increasing evidence indicates the presence of long noncoding RNAs (lncRNAs) is specific to various cell types. Although lncRNAs are speculated to be more numerous than protein-coding genes, the annotations of lncRNAs remain primitive due to the lack of well-structured schemes for their identification and description. Here, we introduce a new knowledge database "ANGIOGENES" (http://angiogenes.uni-frankfurt.de) to allow for in silico screening of protein-coding genes and lncRNAs expressed in various types of endothelial cells, which are present in all tissues. Using the latest annotations of protein-coding genes and lncRNAs, publicly-available RNA-seq data was analyzed to identify transcripts that are expressed in endothelial cells of human, mouse and zebrafish. The analyzed data were incorporated into ANGIOGENES to provide a one-stop-shop for transcriptomics data to facilitate further biological validation. ANGIOGENES is an intuitive and easy-to-use database to allow in silico screening of expressed, enriched and/or specific endothelial transcripts under various conditions. We anticipate that ANGIOGENES serves as a starting point for functional studies to elucidate the roles of protein-coding genes and lncRNAs in angiogenesis.

  15. ANGIOGENES: knowledge database for protein-coding and noncoding RNA genes in endothelial cells.

    Science.gov (United States)

    Müller, Raphael; Weirick, Tyler; John, David; Militello, Giuseppe; Chen, Wei; Dimmeler, Stefanie; Uchida, Shizuka

    2016-01-01

    Increasing evidence indicates the presence of long noncoding RNAs (lncRNAs) is specific to various cell types. Although lncRNAs are speculated to be more numerous than protein-coding genes, the annotations of lncRNAs remain primitive due to the lack of well-structured schemes for their identification and description. Here, we introduce a new knowledge database "ANGIOGENES" (http://angiogenes.uni-frankfurt.de) to allow for in silico screening of protein-coding genes and lncRNAs expressed in various types of endothelial cells, which are present in all tissues. Using the latest annotations of protein-coding genes and lncRNAs, publicly-available RNA-seq data was analyzed to identify transcripts that are expressed in endothelial cells of human, mouse and zebrafish. The analyzed data were incorporated into ANGIOGENES to provide a one-stop-shop for transcriptomics data to facilitate further biological validation. ANGIOGENES is an intuitive and easy-to-use database to allow in silico screening of expressed, enriched and/or specific endothelial transcripts under various conditions. We anticipate that ANGIOGENES serves as a starting point for functional studies to elucidate the roles of protein-coding genes and lncRNAs in angiogenesis. PMID:27582018

  16. Time-Dependent Distribution Functions in C-Mod Calculated with the CQL3D-Hybrid-FOW, AORSA Full-Wave, and DC Lorentz Codes

    Science.gov (United States)

    Harvey, R. W. (Bob); Petrov, Yu. V.; Jaeger, E. F.; Berry, L. A.; Bonoli, P. T.; Bader, A.

    2015-11-01

    A time-dependent simulation of C-Mod pulsed ICRF power is made calculating minority hydrogen ion distribution functions with the CQL3D-Hybrid-FOW finite-orbit-width Fokker-Planck code. ICRF fields are calculated with the AORSA full wave code, and RF diffusion coefficients are obtained from these fields using the DC Lorentz gyro-orbit code. Prior results with a zero-banana-width simulation using the CQL3D/AORSA/DC time-cycles showed a pronounced enhancement of the H distribution in the perpendicular velocity direction compared to results obtained from Stix's quasilinear theory, in general agreement with experiment. The present study compares the new FOW results, including relevant gyro-radius effects, to determine the importance of these effects on the the NPA synthetic diagnostic time-dependence. The new NPA results give increased agreement with experiment, particularly in the ramp-down time after the ICRF pulse. Funded, through subcontract with Massachusetts Institute of Technology, by USDOE sponsored SciDAC Center for Simulation of Wave-Plasma Interactions.

  17. Application of the Synthesis method to the calculations of neutron flow in 3D in the enveloping of a BWR reactor with the DORT code

    International Nuclear Information System (INIS)

    The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)

  18. Exit of a blast wave from a conical nozzle. [flow field calculations by Eulerian computer code DORF

    Science.gov (United States)

    Kim, K.; Johnson, W. E.

    1976-01-01

    The Eulerian computer code DORF was used in the analysis of a two-dimensional, unsteady flow field resulting from semi-confined explosions for propulsive applications. Initially, the ambient gas inside the conical shaped nozzle is set into motion due to the expansion of the explosion product gas, forming a shock wave. When this shock front exits the nozzle, it takes almost a spherical form while a complex interaction between the nozzle and compression and rarefaction waves takes place behind the shock. The results show an excellent agreement with experimental data.

  19. Benchmarking a modified version of the civ3 nonrelativistic atomic-structure code within Na-like-tungsten R -matrix calculations

    Science.gov (United States)

    Turkington, M. D.; Ballance, C. P.; Hibbert, A.; Ramsbottom, C. A.

    2016-08-01

    In this work we explore the validity of employing a modified version of the nonrelativistic structure code civ3 for heavy, highly charged systems, using Na-like tungsten as a simple benchmark. Consequently, we present radiative and subsequent collisional atomic data compared with corresponding results from a fully relativistic structure and collisional model. Our motivation for this line of study is to benchmark civ3 against the relativistic grasp0 structure code. This is an important study as civ3 wave functions in nonrelativistic R -matrix calculations are computationally less expensive than their Dirac counterparts. There are very few existing data for the W LXIV ion in the literature with which we can compare except for an incomplete set of energy levels available from the NIST database. The overall accuracy of the present results is thus determined by the comparison between the civ3 and grasp0 structure codes alongside collisional atomic data computed by the R -matrix Breit-Pauli and Dirac codes. It is found that the electron-impact collision strengths and effective collision strengths computed by these differing methods are in good general agreement for the majority of the transitions considered, across a broad range of electron temperatures.

  20. Classification and modelling of functional outputs of computation codes. Application to accidental thermal-hydraulic calculations in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    This research thesis has been made within the frame of a project on nuclear reactor vessel life. It deals with the use of numerical codes aimed at estimating probability densities for every input parameter in order to calculate probability margins at the output level. More precisely, it deals with codes with one-dimensional functional responses. The author studies the numerical simulation of a pressurized thermal shock on a nuclear reactor vessel, i.e. one of the possible accident types. The study of the vessel integrity relies on a thermal-hydraulic analysis and on a mechanical analysis. Algorithms are developed and proposed for each of them. Input-output data are classified using a clustering technique and a graph-based representation. A method for output dimension reduction is proposed, and a regression is applied between inputs and reduced representations. Applications are discussed in the case of modelling and sensitivity analysis for the CATHARE code (a code used at the CEA for the thermal-hydraulic analysis)

  1. New Particle-in-Cell Code for Numerical Simulation of Coherent Synchrotron Radiation

    Energy Technology Data Exchange (ETDEWEB)

    Balsa Terzic, Rui Li

    2010-05-01

    We present a first look at the new code for self-consistent, 2D simulations of beam dynamics affected by the coherent synchrotron radiation. The code is of the particle-in-cell variety: the beam bunch is sampled by point-charge particles, which are deposited on the grid; the corresponding forces on the grid are then computed using retarded potentials according to causality, and interpolated so as to advance the particles in time. The retarded potentials are evaluated by integrating over the 2D path history of the bunch, with the charge and current density at the retarded time obtained from interpolation of the particle distributions recorded at discrete timesteps. The code is benchmarked against analytical results obtained for a rigid-line bunch. We also outline the features and applications which are currently being developed.

  2. Comparison of particle tracks calculated by Monte Carlo computer codes with experimental tracks observed with the Harwell low pressure cloud chamber

    International Nuclear Information System (INIS)

    Photographs of the droplets associated with the ionisations caused by charged particle tracks in the Harwell low pressure cloud chamber have been analysed. The radiation types these represent are alpha particles, protons and low energy X rays (carbon and aluminium) in either a tissue-equivalent gas or water vapour. The tracks were used to test the validity of two Monte Carlo codes developed by Wilson and Paratzke, namely MOCA14 for the generation of proton and alpha particle tracks, and MOCA8 for the generation of electron tracks. The comparisons showed that the code MOCA14 would appear to be valid for protons with energies greater than about 390 keV, and for alpha particles with energies greater than 1.6 MeV. No disagreement was found between the low energy X ray tracks from the cloud chamber and the tracks calculated from MOCA8, although this comparison was severely limited by droplet diffusion. (author)

  3. One-group transport theory calculation for three slabs cells

    International Nuclear Information System (INIS)

    As an idealized model of plate type fuel assemblies for nuclear reactors, three-slab cells are analysed numerically based on the exact solution of the transport equation in the one-group isotropic scattering model. From the equations describing the interface conditions, a set of regular integral equations for the coefficients of the singular eigenfunctions expansions is derived using the half-range orthogonality relations of the eigenfunctions and the recently developed method of regularization. Numerical solutions are obtained by solving this set of equations iteratively. The thermal utilization factor and thermal disadvantage factors as well as flux and current distributions are reported for the first time for various sets of parameters. The accuracy of the P sub(N) approximations is also analysed compared to the exact results. (Author)

  4. URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

    International Nuclear Information System (INIS)

    The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self- indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling the Doppler broadened cross-section. The various shelf-shielded factors are computed numerically as Lebesgue integrals over the cross-section probability tables. 6 refs

  5. PCRELAP5: data calculation program for RELAP 5 code; PCRELAP5: programa de calculo dos dados de entrada para o codigo RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Silvestre, Larissa Jacome Barros

    2016-07-01

    Nuclear accidents in the world led to the establishment of rigorous criteria and requirements for nuclear power plant operations by the international regulatory bodies. By using specific computer programs, simulations of various accidents and transients likely to occur at any nuclear power plant are required for certifying and licensing a nuclear power plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most widely used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors in Brazil and worldwide. A major difficulty in the simulation by using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data requires a great number of mathematical operations to calculate the geometry of the components. Thus, for those calculations performance and preparation of RELAP5 input data, a friendly mathematical preprocessor was designed. The Visual Basic for Application (VBA) for Microsoft Excel demonstrated to be an effective tool to perform a number of tasks in the development of the program. In order to meet the needs of RELAP5 users, the RELAP5 Calculation Program (Programa de Calculo do RELAP5 - PCRELAP5) was designed. The components of the code were codified; all entry cards including the optional cards of each one have been programmed. In addition, an English version for PCRELAP5 was provided. Furthermore, a friendly design was developed in order to minimize the time of preparation of input data and errors committed by users. In this work, the final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra 2. (author)

  6. Calculations of criticality, nuclide compositions, decay heat and sources for WWER-440 fuel by new version of the SCALE 5 code

    International Nuclear Information System (INIS)

    In this article are compared theoretical results by new version of the SCALE5 code with experiments or other theoretical calculation for: 1. criticality: -measurement on ZR-6 and LR-0; - numerical benchmark No. 1.3 and 4 (CB1, CB3, CB4); 2. nuclide compositions: - measurement in Kurchatov institute for 3,6 %; - measurement in JAERI(PWR 17x17); - numerical benchmark No. 2-Source (CB2); 3. sources and decay heat:- numerical benchmark No.2-Source (CB2-S); The focus is on modules KENO, TRITON and ORIGEN-S (Authors)

  7. In-Depth Analysis of Simulation Engine Codes for Comparison with DOE s Roof Savings Calculator and Measured Data

    Energy Technology Data Exchange (ETDEWEB)

    New, Joshua Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Levinson, Ronnen [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Huang, Yu [White Box Technologies, Salt Lake City, UT (United States); Sanyal, Jibonananda [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mellot, Joe [The Garland Company, Cleveland, OH (United States); Childs, Kenneth W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kriner, Scott [Green Metal Consulting, Inc., Macungie, PA (United States)

    2014-06-01

    The Roof Savings Calculator (RSC) was developed through collaborations among Oak Ridge National Laboratory (ORNL), White Box Technologies, Lawrence Berkeley National Laboratory (LBNL), and the Environmental Protection Agency in the context of a California Energy Commission Public Interest Energy Research project to make cool-color roofing materials a market reality. The RSC website and a simulation engine validated against demonstration homes were developed to replace the liberal DOE Cool Roof Calculator and the conservative EPA Energy Star Roofing Calculator, which reported different roof savings estimates. A preliminary analysis arrived at a tentative explanation for why RSC results differed from previous LBNL studies and provided guidance for future analysis in the comparison of four simulation programs (doe2attic, DOE-2.1E, EnergyPlus, and MicroPas), including heat exchange between the attic surfaces (principally the roof and ceiling) and the resulting heat flows through the ceiling to the building below. The results were consolidated in an ORNL technical report, ORNL/TM-2013/501. This report is an in-depth inter-comparison of four programs with detailed measured data from an experimental facility operated by ORNL in South Carolina in which different segments of the attic had different roof and attic systems.

  8. Calculation of response of Chinese hamster cells to ions based on track structure theory

    Institute of Scientific and Technical Information of China (English)

    LiuXiao-Wei; ZhangChun-Xiang

    1997-01-01

    Considering biological cells as single target two-hit detectors,an analytic formula to calculate the response of cells to ions is developed based on track structure theory.In the calculation,the splitting deposition energy between ion kill mode and γ kill mode is not used.The results of calculation are in agreement with the experimental data for response of Chinese hamster cells,whose response to γ rays can be described by the response function of single target two hit detector to ions.

  9. Fluence-to-dose conversion coefficients for heavy ions calculated using the PHITS code and the ICRP/ICRU adult reference computational phantoms

    Science.gov (United States)

    Sato, Tatsuhiko; Endo, Akira; Niita, Koji

    2010-04-01

    The fluence to organ-absorbed-dose and effective-dose conversion coefficients for heavy ions with atomic numbers up to 28 and energies from 1 MeV/nucleon to 100 GeV/nucleon were calculated using the PHITS code coupled to the ICRP/ICRU adult reference computational phantoms, following the instruction given in ICRP Publication 103 (2007 (Oxford: Pergamon)). The conversion coefficients for effective dose equivalents derived using the radiation quality factors of both Q(L) and Q(y) relationships were also estimated, utilizing the functions for calculating the probability densities of absorbed dose in terms of LET (L) and lineal energy (y), respectively, implemented in PHITS. The calculation results indicate that the effective dose can generally give a conservative estimation of the effective dose equivalent for heavy-ion exposure, although it is occasionally too conservative especially for high-energy lighter-ion irradiations. It is also found from the calculation that the conversion coefficients for the Q(y)-based effective dose equivalents are generally smaller than the corresponding Q(L)-based values because of the conceptual difference between LET and y as well as the numerical incompatibility between the Q(L) and Q(y) relationships. The calculated data of these dose conversion coefficients are very useful for the dose estimation of astronauts due to cosmic-ray exposure.

  10. Long non-coding RNA profiling of human lymphoid progenitor cells reveals transcriptional divergence of B cell and T cell lineages.

    Science.gov (United States)

    Casero, David; Sandoval, Salemiz; Seet, Christopher S; Scholes, Jessica; Zhu, Yuhua; Ha, Vi Luan; Luong, Annie; Parekh, Chintan; Crooks, Gay M

    2015-12-01

    To elucidate the transcriptional 'landscape' that regulates human lymphoid commitment during postnatal life, we used RNA sequencing to assemble the long non-coding transcriptome across human bone marrow and thymic progenitor cells spanning the earliest stages of B lymphoid and T lymphoid specification. Over 3,000 genes encoding previously unknown long non-coding RNAs (lncRNAs) were revealed through the analysis of these rare populations. Lymphoid commitment was characterized by lncRNA expression patterns that were highly stage specific and were more lineage specific than those of protein-coding genes. Protein-coding genes co-expressed with neighboring lncRNA genes showed enrichment for ontologies related to lymphoid differentiation. The exquisite cell-type specificity of global lncRNA expression patterns independently revealed new developmental relationships among the earliest progenitor cells in the human bone marrow and thymus.

  11. The chemical coding of 5-hydroxytryptamine containing enteroendocrine cells in the mouse gastrointestinal tract.

    Science.gov (United States)

    Reynaud, Yohan; Fakhry, Josiane; Fothergill, Linda; Callaghan, Brid; Ringuet, Mitchell; Hunne, Billie; Bravo, David M; Furness, John B

    2016-06-01

    The majority of 5-HT (serotonin) in the body is contained in enteroendocrine cells of the gastrointestinal mucosa. From the time of their discovery over 80 years ago, the 5-HT-containing cells have been regarded as a class of cell that is distinct from enteroendocrine cells that contain peptide hormones. However, recent studies have cast doubt on the concept of there being distinct classes of enteroendocrine cells, each containing a single hormone or occasionally more than one hormone. Instead, data are rapidly accumulating that there are complex patterns of colocalisation of hormones that identify multiple subclasses of enteroendocrine cells. In the present work, multiple labelling immunohistochemistry is used to investigate patterns of colocalisation of 5-HT with enteric peptide hormones. Over 95 % of 5-HT cells in the duodenum also contained cholecystokinin and about 40 % of them also contained secretin. In the jejunum, about 75 % of 5-HT cells contained cholecystokinin but not secretin and 25 % contained 5-HT plus both cholecystokinin and secretin. Small proportions of 5-HT cells contained gastrin or somatostatin in the stomach, PYY or GLP-1 in the small intestine and GLP-1 or somatostatin in the large intestine. Rare or very rare 5-HT cells contained ghrelin (stomach), neurotensin (small and large intestines), somatostatin (small intestine) and PYY (in the large intestine). It is concluded that 5-HT-containing enteroendocrine cells are heterogeneous in their chemical coding and by implication in their functions. PMID:26803512

  12. Development and validation of CONV-3D code for calculation of thermal and hydrodynamics of Fast Reactor at the Supercomputer

    International Nuclear Information System (INIS)

    In IBRAE 3D CFD modules (CONV code) for safety analysis of the operated Nuclear Power Plants (NPPs) are developed. These modules are based on the developed algorithms with small scheme diffusion, for which the discrete approximations are constructed with use of finite-volume methods and fully staggered grids. For solving of convection problem the regularized nonlinear monotonic operator-splitting scheme is developed. The Richardson iterative method with iterative Fast Fourier Transformation (FFT) solver for Laplace’s operator as preconditioner is applied for solving pressure equation. Such approach for solving of the elliptical equations with variable coefficients gives multiple acceleration in a comparison with a usual method of conjugate gradients. For modeling of 3D turbulent single-phase flows Quasi DNS approach is used. The CONV code is fully parallelized and highly effective at the high performance computers such as “Chebyshev”, “Lomonosov” (Moscow State University). The developed modules were validated on a series of the well known tests in a wide range of Rayleigh numbers from a range 106-1016 and Reynolds numbers from a range 103-105. The software has been applied to the analysis results of test LIVE-L1 (L1 is aimed at investigating the melt pool and crust behaviour during the stages of air circulation at the outer RPV surface with subsequent flooding of the lower head) and joint analyses on transient molten pool thermal hydraulics in the LIVE facility in the framework of ISTC project. Moreover CONV was validated successfully on a series of the experimental tests as: the blind test on simulation of flows in T-junction (OECD/NEA), ERCOFTAC experiment (world database on turbulent flows) natural convection in the closures under extremely high Rayleigh numbers. In all cases the good coincidence of numerical predictions with experimental data was reached, that specifies a possibility of application of the developed approach for a prediction of CFD

  13. First experience with particle-in-cell plasma physics code on ARM-based HPC systems

    Science.gov (United States)

    Sáez, Xavier; Soba, Alejandro; Sánchez, Edilberto; Mantsinen, Mervi; Mateo, Sergi; Cela, José M.; Castejón, Francisco

    2015-09-01

    In this work, we will explore the feasibility of porting a Particle-in-cell code (EUTERPE) to an ARM multi-core platform from the Mont-Blanc project. The used prototype is based on a system-on-chip Samsung Exynos 5 with an integrated GPU. It is the first prototype that could be used for High-Performance Computing (HPC), since it supports double precision and parallel programming languages.

  14. Use of fluorescent proteins and color-coded imaging to visualize cancer cells with different genetic properties.

    Science.gov (United States)

    Hoffman, Robert M

    2016-03-01

    Fluorescent proteins are very bright and available in spectrally-distinct colors, enable the imaging of color-coded cancer cells growing in vivo and therefore the distinction of cancer cells with different genetic properties. Non-invasive and intravital imaging of cancer cells with fluorescent proteins allows the visualization of distinct genetic variants of cancer cells down to the cellular level in vivo. Cancer cells with increased or decreased ability to metastasize can be distinguished in vivo. Gene exchange in vivo which enables low metastatic cancer cells to convert to high metastatic can be color-coded imaged in vivo. Cancer stem-like and non-stem cells can be distinguished in vivo by color-coded imaging. These properties also demonstrate the vast superiority of imaging cancer cells in vivo with fluorescent proteins over photon counting of luciferase-labeled cancer cells.

  15. Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

    Energy Technology Data Exchange (ETDEWEB)

    Surkov, A. V., E-mail: surkov.andrew@gmail.com; Kochkin, V. N.; Pesnya, Yu. E.; Nasonov, V. A.; Vihrov, V. I.; Erak, D. Yu. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields in the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.

  16. OCA-II, a code for calculating the behavior of 2-D and 3-D surface flaws in a pressure vessel subjected to temperature and pressure transients

    International Nuclear Information System (INIS)

    The OCA-II computer code, like its predecessor OCA-I, performs the thermal, stress, and linear elastic fracture-mechanics analysis for long flaws on the surface of a cylinder that is subjected to thermal and pressure transients. OCA-II represents a revised and expanded version of OCA-I and includes as new features (1) cladding as a discrete region, (2) a finite-element subroutine for calculating the stresses, and (3) the ability to calculate stress intensity factors for certain three-dimensional flaws, for two-dimensional circumferential flaws on the inner surface, and for both axial and circumferential flaws on the outer surface. OCA-I considered only inner-surface flaws. An option is included in OCA-II that permits a search for critical values of fluence or nil-ductility reference temperature corresponding to a specified failure criterion. These and other features of OCA-II are described in the report, which also includes user instructions for the code

  17. Sci—Thur AM: YIS - 03: irtGPUMCD: a new GPU-calculated dosimetry code for {sup 177}Lu-octreotate radionuclide therapy of neuroendocrine tumors

    Energy Technology Data Exchange (ETDEWEB)

    Montégiani, Jean-François; Gaudin, Émilie; Després, Philippe [Physics, Engineering Physics and Optics, Université Laval, Quebec City, QC (Canada); Jackson, Price A. [Molecular Imaging and Targeted Therapeutics, Peter MacCallum Cancer Centre, Melbourne, VIC (Australia); Beauregard, Jean-Mathieu [Radiology, Université Laval, Quebec City, QC (Canada)

    2014-08-15

    In peptide receptor radionuclide therapy (PRRT), huge inter-patient variability in absorbed radiation doses per administered activity mandates the utilization of individualized dosimetry to evaluate therapeutic efficacy and toxicity. We created a reliable GPU-calculated dosimetry code (irtGPUMCD) and assessed {sup 177}Lu-octreotate renal dosimetry in eight patients (4 cycles of approximately 7.4 GBq). irtGPUMCD was derived from a brachytherapy dosimetry code (bGPUMCD), which was adapted to {sup 177}Lu PRRT dosimetry. Serial quantitative single-photon emission computed tomography (SPECT) images were obtained from three SPECT/CT acquisitions performed at 4, 24 and 72 hours after {sup 177}Lu-octreotate administration, and registered with non-rigid deformation of CT volumes, to obtain {sup 177}Lu-octreotate 4D quantitative biodistribution. Local energy deposition from the β disintegrations was assumed. Using Monte Carlo gamma photon transportation, irtGPUMCD computed dose rate at each time point. Average kidney absorbed dose was obtained from 1-cm{sup 3} VOI dose rate samples on each cortex, subjected to a biexponential curve fit. Integration of the latter time-dose rate curve yielded the renal absorbed dose. The mean renal dose per administered activity was 0.48 ± 0.13 Gy/GBq (range: 0.30–0.71 Gy/GBq). Comparison to another PRRT dosimetry code (VRAK: Voxelized Registration and Kinetics) showed fair accordance with irtGPUMCD (11.4 ± 6.8 %, range: 3.3–26.2%). These results suggest the possibility to use the irtGPUMCD code in order to personalize administered activity in PRRT. This could allow improving clinical outcomes by maximizing per-cycle tumor doses, without exceeding the tolerable renal dose.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

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

    Science.gov (United States)

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

    2016-02-01

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

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

    Directory of Open Access Journals (Sweden)

    Dzhalandinov A.

    2016-01-01

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

  1. Identification of a long non-coding RNA gene, growth hormone secretagogue receptor opposite strand, which stimulates cell migration in non-small cell lung cancer cell lines.

    Science.gov (United States)

    Whiteside, Eliza J; Seim, Inge; Pauli, Jana P; O'Keeffe, Angela J; Thomas, Patrick B; Carter, Shea L; Walpole, Carina M; Fung, Jenny N T; Josh, Peter; Herington, Adrian C; Chopin, Lisa K

    2013-08-01

    The molecular mechanisms involved in non‑small cell lung cancer tumourigenesis are largely unknown; however, recent studies have suggested that long non-coding RNAs (lncRNAs) are likely to play a role. In this study, we used public databases to identify an mRNA-like, candidate long non-coding RNA, GHSROS (GHSR opposite strand), transcribed from the antisense strand of the ghrelin receptor gene, growth hormone secretagogue receptor (GHSR). Quantitative real-time RT-PCR revealed higher expression of GHSROS in lung cancer tissue compared to adjacent, non-tumour lung tissue. In common with many long non-coding RNAs, GHSROS is 5' capped and 3' polyadenylated (mRNA-like), lacks an extensive open reading frame and harbours a transposable element. Engineered overexpression of GHSROS stimulated cell migration in the A549 and NCI-H1299 non-small cell lung cancer cell lines, but suppressed cell migration in the Beas-2B normal lung-derived bronchoepithelial cell line. This suggests that GHSROS function may be dependent on the oncogenic context. The identification of GHSROS, which is expressed in lung cancer and stimulates cell migration in lung cancer cell lines, contributes to the growing number of non-coding RNAs that play a role in the regulation of tumourigenesis and metastatic cancer progression.

  2. Expression-guided in silico evaluation of candidate cis regulatory codes for Drosophila muscle founder cells.

    Directory of Open Access Journals (Sweden)

    Anthony A Philippakis

    2006-05-01

    Full Text Available While combinatorial models of transcriptional regulation can be inferred for metazoan systems from a priori biological knowledge, validation requires extensive and time-consuming experimental work. Thus, there is a need for computational methods that can evaluate hypothesized cis regulatory codes before the difficult task of experimental verification is undertaken. We have developed a novel computational framework (termed "CodeFinder" that integrates transcription factor binding site and gene expression information to evaluate whether a hypothesized transcriptional regulatory model (TRM; i.e., a set of co-regulating transcription factors is likely to target a given set of co-expressed genes. Our basic approach is to simultaneously predict cis regulatory modules (CRMs associated with a given gene set and quantify the enrichment for combinatorial subsets of transcription factor binding site motifs comprising the hypothesized TRM within these predicted CRMs. As a model system, we have examined a TRM experimentally demonstrated to drive the expression of two genes in a sub-population of cells in the developing Drosophila mesoderm, the somatic muscle founder cells. This TRM was previously hypothesized to be a general mode of regulation for genes expressed in this cell population. In contrast, the present analyses suggest that a modified form of this cis regulatory code applies to only a subset of founder cell genes, those whose gene expression responds to specific genetic perturbations in a similar manner to the gene on which the original model was based. We have confirmed this hypothesis by experimentally discovering six (out of 12 tested new CRMs driving expression in the embryonic mesoderm, four of which drive expression in founder cells.

  3. Naturally Occurring Self-Reactive CD4+CD25+ Regulatory T Cells: Universal Immune Code

    Institute of Scientific and Technical Information of China (English)

    Nafiseh Pakravan; Agheel Tabar Molla Hassan; Zuhair Muhammad Hassan

    2007-01-01

    Naturally occurring thymus-arisen CD4+CD25+ regulatory T (Treg) cells are considered to play a central role in self-tolerance. Precise signals that promote the development of Treg cells remain elusive, but considerable evidence suggests that costimulatory molecules, cytokines, the nature of the TCR and the niche or the context in which the T cell encounters antigen in the thymus play important roles. Analysis of TCR from Treg cells has demonstrated that a large proportion of this population has a higher avidity to self-antigen in comparison with TCR from CD4+CD25- cells and that peripheral antigen is required for their development, maintenance, or expansion. Treg cells have been shown to undergo expansion in the periphery, likely regulated by the presence of self-antigen. Many studies have shown that the involvement of Treg cells in the tolerance induction is antigen-specific, even with MHC-mismatched,in transplantation/graft versus host disease (GVHD), autoimmunity, cancer, and pregnancy. Theses studies concluded a vital role for self-reactive Treg cells in maintenance of the body integrity. Based on those studies, we hypothesize that self-reactive Treg cells are shared among all healthy individuals and recognize same self-antigens and their TCR encodes for few dominant antigens of each organ which defines the healthy self. These dominant self antigens can be regarded as "universal immune code".

  4. Regulations and ethical codes for clinical cell therapy trials in Iran

    Institute of Scientific and Technical Information of China (English)

    Hooshang Saberi; Nazi Derakhshanrad; Babak Arjmand; Jafar Ai; Masoud Soleymani; Amir Ali Hamidieh; Mohammad Taghi Joghataei; Zahid Hussain Khan; Seyed Hassan Emami Razavi

    2015-01-01

    Objective:The local regulations for conducting experimental and clinical cell therapy studies are dependent on the national and cultural approach to the issue, and may have many common aspects as well as differences with the regulations in other countries. The study reflects the latest national aspects of cell therapy in Iran and relevant regulations. Methods:The following topics are discussed in the article including sources of cell harvest, regulations for cell disposal, stem cell manufacturing, and economic aspects of stem cell, based on current practice in Iran. Results:All cell therapy trials in Iran are required to strictly abide with the ethical codes, national and local regulations, and safety requirements, as well as considering human rights and respect. Adherence to these standards has facilitated the conduct of human cell therapy trials for research, academic advancement, and therapy. Conclusions:The cell therapy trials based on the aforementioned regulations may be assumed to be ethical and they are candidates for clinical translations based on safety and efficacy issues.

  5. Chemical coding and chemosensory properties of cholinergic brush cells in the mouse gastrointestinal and biliary tract

    Directory of Open Access Journals (Sweden)

    Burkhard eSchütz

    2015-03-01

    Full Text Available The mouse gastro-intestinal and biliary tract mucosal epithelia harbor choline acetyltransferase (ChAT-positive brush cells with taste cell-like traits. With the aid of two transgenic mouse lines that express green fluorescent protein (EGFP under the control of the ChAT promoter (EGFPChAT and by using in situ hybridization and immunohistochemistry we found that EGFPChAT cells were clustered in the epithelium lining the gastric groove. EGFPChAT cells were numerous in the gall bladder and bile duct, and found scattered as solitary cells along the small and large intestine. While all EGFPChAT cells were also ChAT-positive, expression of the high-affinity choline transporter (ChT1 was never detected. Except for the proximal colon, EGFPChAT cells also lacked detectable expression of the vesicular acetylcholine transporter (VAChT. EGFPChAT cells were found to be separate from enteroendocrine cells, however they were all immunoreactive for cytokeratin 18 (CK18, transient receptor potential melastatin-like subtype 5 channel (TRPM5, and for cyclooxygenases 1 (COX1 and 2 (COX2. The ex vivo stimulation of colonic EGFPChAT cells with the bitter substance denatonium resulted in a strong increase in intracellular calcium, while in other epithelial cells such an increase was significantly weaker and also timely delayed. Subsequent stimulation with cycloheximide was ineffective in both cell populations. Given their chemical coding and chemosensory properties, EGFPChAT brush cells thus may have integrative functions and participate in induction of protective reflexes and inflammatory events by utilizing ACh and prostaglandins for paracrine signaling.

  6. Mesenchymal stem cells from different organs are characterized by distinct topographic Hox codes.

    Science.gov (United States)

    Ackema, Karin B; Charité, Jeroen

    2008-10-01

    Mesenchymal stem cells (MSC) are multipotent cells found as part of the stromal compartment of the bone marrow and in many other organs. They can be identified in vitro as CFU-F (colony forming unit-fibroblast) based on their ability to form adherent colonies of fibroblast-like cells in culture. MSC expanded in vitro retain characteristics appropriate to their tissue of origin. This is reflected in their propensity for differentiating towards specific lineages, and their capacity to generate, upon retransplantation in vivo, a stroma supporting typical lineages of hematopoietic cells. Hox genes encode master regulators of regional specification and organ development in the embryo and are widely expressed in the adult. We investigated whether they could be involved in determining tissue-specific properties of MSC. Hox gene expression profiles of individual CFU-F colonies derived from various organs and anatomical locations were generated, and the relatedness between these profiles was determined using hierarchical cluster analysis. This revealed that CFU-F have characteristic Hox expression signatures that are heterogeneous but highly specific for their anatomical origin. The topographic specificity of these Hox codes is maintained during differentiation, suggesting that they are an intrinsic property of MSC. Analysis of Hox codes of CFU-F from vertebral bone marrow suggests that MSC originate over a large part of the anterioposterior axis, but may not originate from prevertebral mesenchyme. These data are consistent with a role for Hox proteins in specifying cellular identity of MSC.

  7. Calculation of response and thindown of V-79 cell for ion irradiation

    Institute of Scientific and Technical Information of China (English)

    CHEN Li-Xin; LIU Xiao-Wei

    2004-01-01

    A cellular survival model and the cross section calculation with low and high LET for ion irradiation were presented. Based on our formula of surviving fraction calculation, the survival data of Chinese hamster cell (V-79)for ion irradiation including He, Li, B, C, O, Ne and Ar were calculated; the cross sections for ion irradiation including He, Ni, C, Ar, Kr, Xe and U were shown. The calculated results show that the presented model is a good description of radiation effects of V-79 cell for different ion irradiation. In this model splitting energy between ion-kill mode and gamma-kill model is avoided, the calculated results of cross section needn't be multiplied by a factor to fit the experimental data.

  8. A highly optimized code for calculating atomic data at neutron star magnetic field strengths using a doubly self-consistent Hartree-Fock-Roothaan method

    Science.gov (United States)

    Schimeczek, C.; Engel, D.; Wunner, G.

    2014-05-01

    Our previously published code for calculating energies and bound-bound transitions of medium-Z elements at neutron star magnetic field strengths [D. Engel, M. Klews, G. Wunner, Comp. Phys. Comm. 180, 3-2-311 (2009)] was based on the adiabatic approximation. It assumes a complete decoupling of the (fast) gyration of the electrons under the action of the magnetic field and the (slow) bound motion along the field under the action of the Coulomb forces. For the single-particle orbitals this implied that each is a product of a Landau state and an (unknown) longitudinal wave function whose B-spline coefficients were determined self-consistently by solving the Hartree-Fock equations for the many-electron problem on a finite-element grid. In the present code we go beyond the adiabatic approximation, by allowing the transverse part of each orbital to be a superposition of Landau states, while assuming that the longitudinal part can be approximated by the same wave function in each Landau level. Inserting this ansatz into the energy variational principle leads to a system of coupled equations in which the B-spline coefficients depend on the weights of the individual Landau states, and vice versa, and which therefore has to be solved in a doubly self-consistent manner. The extended ansatz takes into account the back-reaction of the Coulomb motion of the electrons along the field direction on their motion in the plane perpendicular to the field, an effect which cannot be captured by the adiabatic approximation. The new code allows for the inclusion of up to 8 Landau levels. This reduces the relative error of energy values as compared to the adiabatic approximation results by typically a factor of three (1/3 of the original error) and yields accurate results also in regions of lower neutron star magnetic field strengths where the adiabatic approximation fails. Further improvements in the code are a more sophisticated choice of the initial wave functions, which takes into

  9. Comparison of the Hadley cells calculated from two reanalysis data sets

    Institute of Scientific and Technical Information of China (English)

    QIN Yujing; WANG Panxing; GUAN Zhaoyong; YUE Yang

    2006-01-01

    The mass stream function of mean meridional circulation is calculated from the ECMWF and NCEP/NCAR reanalysis data sets using a superposition computation scheme. The comparison of results shows that the common ascending leg of the Hadley cell calculated from the ECMWF data is strong and narrow, and averagely lies more north of the equator in comparison with its counterpart from the NCEP/NCAR data, and furthermore the Hadley cell from the ECMWF data shows an obvious double-layer structure. Therefore, there are obvious differences between Hadley cells displayed by the two objective analysis data sets.

  10. Shielding Calculations for Industrial 5/7.5MeV Electron Accelerators Using the MCNP Monte Carlo Code

    International Nuclear Information System (INIS)

    High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, in order to extend the shelf life of products. High energy photons can cause food activation due to (D3,n) reactions. Until 2004, to eliminate the possibility of food activation, the electron energy was limited to 5 MeV X-rays for food irradiation. In 2004, the FDA approved the usage of up to 7.5 MeV, but only with tantalum and gold targets (1). Higher X-ray energy results an increased flux of X-rays in the forward direction, increased penetration, and higher photon dose rate due to better electron-to-photon conversion. These improvements could decrease the irradiation time and allow irradiation of larger packages, thereby providing higher production rates with lower treatment cost. Medical accelerators usually work with 6-18 MV electron energy with tungsten target to convert the electron beam to X-rays. In order to protect the patients, the accelerator head is protected with a heavy lead shielding; therefore, the bremsstrahlung is emitted only in the forward direction. There are many publications and standards that guide how to design optimal shielding for medical accelerator rooms. The shielding data for medical accelerators is not applicable for industrial accelerators, since the data is for different conversion targets, different X-Ray energies, and only for the forward direction. Collimators are not always in use in industrial accelerators, and therefore bremsstrahlung photons can be emitted in all directions. The bremsstrahlung spectrum and dose rate change as a function of the emission angle. The dose rate decreases from maximum in the forward direction (0°) to minimum at 180° by 1-2 orders of magnitude. In order to design and calculate optimal shielding for food accelerator rooms, there is a need to have the bremsstrahlung spectrum data, dose rates and concrete attenuation data in all emission directions

  11. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Morgan C. White

    2000-07-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second

  12. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    International Nuclear Information System (INIS)

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to

  13. Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

    International Nuclear Information System (INIS)

    The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 k W. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 k W. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively. (author)

  14. Dose estimation for astronauts using dose conversion coefficients calculated with the PHITS code and the ICRP/ICRU adult reference computational phantoms.

    Science.gov (United States)

    Sato, Tatsuhiko; Endo, Akira; Sihver, Lembit; Niita, Koji

    2011-03-01

    Absorbed-dose and dose-equivalent rates for astronauts were estimated by multiplying fluence-to-dose conversion coefficients in the units of Gy.cm(2) and Sv.cm(2), respectively, and cosmic-ray fluxes around spacecrafts in the unit of cm(-2) s(-1). The dose conversion coefficients employed in the calculation were evaluated using the general-purpose particle and heavy ion transport code system PHITS coupled to the male and female adult reference computational phantoms, which were released as a common ICRP/ICRU publication. The cosmic-ray fluxes inside and near to spacecrafts were also calculated by PHITS, using simplified geometries. The accuracy of the obtained absorbed-dose and dose-equivalent rates was verified by various experimental data measured both inside and outside spacecrafts. The calculations quantitatively show that the effective doses for astronauts are significantly greater than their corresponding effective dose equivalents, because of the numerical incompatibility between the radiation quality factors and the radiation weighting factors. These results demonstrate the usefulness of dose conversion coefficients in space dosimetry. PMID:20835833

  15. U.S. Department Of Energy's nuclear engineering education research: highlights of recent and current research-II. 1. Comparison of Angular Approximations for PWR Cell Calculations

    International Nuclear Information System (INIS)

    Increasing computer power is allowing higher-order angular approximations to replace diffusion theory methods in whole core reactor physics computations. Spherical harmonic (Pn), simplified spherical harmonic (SPn), and discrete ordinates (Sn) methods are capable of performing such calculations in three dimensions. Most advantages of such transport methods are gained by eliminating fuel assembly homogenization, thus allowing pin powers to be calculated directly. A further step, currently under investigation, is the elimination of spatial homogenization at the pin cell level as well. The fuel-moderator interfaces may be treated explicitly in Pn, Sn, or SPn calculations by applying triangular finite elements (FEM) to the spatial variables. Early results using a modified form of the VARIANT code, however, indicate that without pin cell homogenization, high-order angular approximations may be required to represent the lattice effects accurately within the whole-core calculations. To examine these lattice effects further, a modified form of VARIANT was created to use the spatial triangular finite element scheme. The program was set up to treat a single heterogeneous pin cell coupled with Pn, SPn, or Sn angular approximations. Additional modifications replaced the nodal interface approximations with exact reflected boundary conditions to increase the accuracy of the results. Several pressurized water reactor pin cells, taken from a previous benchmark specification, were examined. However, the results shown here focus only on the most severe case, i.e., a pin cell containing 8.7% mixed-oxide enriched fuel. The DRAGON collision probability code was used to collapse a 69-group cross-section library to a more manageable 7-group library that contained cross sections for the fuel-cladding mixture and for the water. Eigenvalue results are shown in Figs. 1 and 2 using the modified VARIANT code with Pn, SPn, and Sn angular approximations. A 7-group MCNP Monte Carlo solution and a

  16. The STATFLUX code: a statistical method for calculation of flow and set of parameters, based on the Multiple-Compartment Biokinetical Model

    Science.gov (United States)

    Garcia, F.; Mesa, J.; Arruda-Neto, J. D. T.; Helene, O.; Vanin, V.; Milian, F.; Deppman, A.; Rodrigues, T. E.; Rodriguez, O.

    2007-03-01

    The code STATFLUX, implementing a new and simple statistical procedure for the calculation of transfer coefficients in radionuclide transport to animals and plants, is proposed. The method is based on the general multiple-compartment model, which uses a system of linear equations involving geometrical volume considerations. Flow parameters were estimated by employing two different least-squares procedures: Derivative and Gauss-Marquardt methods, with the available experimental data of radionuclide concentrations as the input functions of time. The solution of the inverse problem, which relates a given set of flow parameter with the time evolution of concentration functions, is achieved via a Monte Carlo simulation procedure. Program summaryTitle of program:STATFLUX Catalogue identifier:ADYS_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADYS_v1_0 Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Licensing provisions: none Computer for which the program is designed and others on which it has been tested:Micro-computer with Intel Pentium III, 3.0 GHz Installation:Laboratory of Linear Accelerator, Department of Experimental Physics, University of São Paulo, Brazil Operating system:Windows 2000 and Windows XP Programming language used:Fortran-77 as implemented in Microsoft Fortran 4.0. NOTE: Microsoft Fortran includes non-standard features which are used in this program. Standard Fortran compilers such as, g77, f77, ifort and NAG95, are not able to compile the code and therefore it has not been possible for the CPC Program Library to test the program. Memory required to execute with typical data:8 Mbytes of RAM memory and 100 MB of Hard disk memory No. of bits in a word:16 No. of lines in distributed program, including test data, etc.:6912 No. of bytes in distributed program, including test data, etc.:229 541 Distribution format:tar.gz Nature of the physical problem:The investigation of transport mechanisms for

  17. Molecular codes for neuronal individuality and cell assembly in the brain

    Directory of Open Access Journals (Sweden)

    Takeshi eYagi

    2012-04-01

    Full Text Available The brain contains an enormous, but finite, number of neurons. The ability of this limited number of neurons to produce nearly limitless neural information over a lifetime is typically explained by combinatorial explosion; that is, by the exponential amplification of each neuron’s contribution through its incorporation into cell assemblies and neural networks. In development, each neuron expresses diverse cellular recognition molecules that permit the formation of the appropriate neural cell assemblies to elicit various brain functions. The mechanism for generating neuronal assemblies and networks must involve molecular codes that give neurons individuality and allow them to recognize one another and join appropriate networks. The extensive molecular diversity of cell-surface proteins on neurons is likely to contribute to their individual identities. The cadherin-related neuronal receptors and clustered protocadherins (CNR/Pcdh is a large subfamily within the diverse cadherin superfamily. The CNR/Pcdh genes are encoded in tandem by three gene clusters, and are present in all known vertebrate genomes. The set of CNR/Pcdh genes is expressed in a random and combinatorial manner in each neuron. In addition, cis-tetramers composed of heteromultimeric CNR/Pcdh isoforms represent selective binding units for cell-cell interactions. Here I present the mathematical probabilities for neuronal individuality based on the random and combinatorial expression of CNR/Pcdh isoforms and their formation of cis-tetramers in each neuron. Notably, CNR/Pcdh gene products are known to play crucial roles in correct axonal projections, synaptic formation, and neuronal survival. Their molecular and biological features suggest that the diverse CNR/Pcdh molecules provide the molecular code by which neuronal individuality and cell assembly permit the combinatorial explosion of networks that supports enormous processing capability and plasticity of the brain.

  18. Thermal-Hydraulic Calculation for Simplified Fuel Assembly of Super Fast Reactor Using Two-Fluid Model Analysis Code ACE-3D

    International Nuclear Information System (INIS)

    To evaluate thermal hydraulic characteristics of a fuel assembly of supercritical water-cooled fast reactor (Super Fast Reactor), a simplified fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D which has been enhanced by Japan Atomic Energy Agency. In the ACE-3D code, the two-phase flow turbulent model based on the k-ε model were adopted. The analytical geometry simulates a 19-rod fuel assembly, which is a simplified geometry of the 271-rod fuel assembly and includes all three kinds of different subchannel types; (1): adjoining to the channel box, (2): next to type (1), and (3): located inside types (1) and (2). In this calculation, one-twelfth model is adopted as the computational domain taking advantage of symmetry. As the boundary conditions, mass velocity, inlet enthalpy and power per rod are to be the same as the steady state condition of the Super Fast Reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. Rod surface temperatures take peak values near the top of the rods. Maximum clad surface temperature (MCST) is observed at the position facing to the narrowest gap on the center rod near the outlet and the value is 902 K (629 deg. C). It was confirmed that the predicted MCST satisfies a thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650 deg. C. (author)

  19. Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Nasrabadi, M.N. [Department of Physics, Faculty of Science, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)], E-mail: mnnasri@kashanu.ac.ir; Jalali, M. [Isfahan Nuclear Science and Technology Research Institute, Atomic Energy organization of Iran (Iran, Islamic Republic of); Mohammadi, A. [Department of Physics, Faculty of Science, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)

    2007-10-15

    In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF{sub 3} detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.

  20. Analysis of core physics test data and sodium void reactivity worth calculation for MONJU core with ARCADIAN-FBR computer code system

    International Nuclear Information System (INIS)

    In order to evaluate core characteristics of fast reactors, a computer code system ARCADIAN-FBR has been developed by utilizing the existing analysis codes and the latest nuclear data library JENDL-3.3. The validity of ARCADIAN-FBR was verified by using the experimental data obtained in the MONJU core physics tests. The results of analyses are in good agreement with the experimental data and the applicability of ARCADIAN-FBR for fast reactor core analysis is confirmed. Using ARCADIAN-FBR, the sodium void reactivity worth, which is an important parameter in the safety analysis of fast reactors, was analyzed for MONJU core. 241Pu in the core fuel is transmuted to 241Am due to disintegrations. Therefore, the effect of 241Am accumulation on the sodium void reactivity worth was evaluated for MONJU core. As a result of calculation, it was confirmed that the accumulation of 241Am significantly influences on the sodium void reactivity worth and hence on the safety analysis of sodium-cooled fast reactors. (author)

  1. Parallel treatment of simulation particles in particle-in-cell codes on SUPRENUM

    International Nuclear Information System (INIS)

    This report contains the program documentation and description of the program package 2D-PLAS, which has been developed at the Nuclear Research Center Karlsruhe in the Institute for Data Processing in Technology (IDT) under the auspices of the BMFT. 2D-PLAS is a parallel program version of the treatment of the simulation particles of the two-dimensional stationary particle-in-cell code BFCPIC which has been developed at the Nuclear Research Center Karlsruhe. This parallel version has been designed for the parallel computer SUPRENUM. (orig.)

  2. Upregulation of long non-coding RNA PRNCR1 in colorectal cancer promotes cell proliferation and cell cycle progression.

    Science.gov (United States)

    Yang, Liu; Qiu, Mantang; Xu, Youtao; Wang, Jie; Zheng, Yanyan; Li, Ming; Xu, Lin; Yin, Rong

    2016-01-01

    Colorectal cancer (CRC) is one of the most common cancers worldwide. Long non-coding RNAs (lncRNAs) have been confirmed to play a critical regulatory role in various biological processes including carcinogenesis, which indicates that lncRNAs are valuable biomarkers and therapeutic targets. The novel lncRNA prostate cancer non-coding RNA 1 (PRNCR1) is located in the susceptible genomic area of CRC, however the functional role of PRNCR1 remains unknown. Thus, we aimed to investigate the clinical significance and biological function of PRNCR1 in CRC. Quantitative real-time polymerase chain reaction (qRT-PCR) was used to assess the expression profile of PRNCR1 in CRC tissues and cell lines. An antisense oligonucleotide (ASO) was designed to knock down PRNCR1. In a cohort of 63 patients, PRNCR1 was significantly overexpressed in CRC tissues compared with the expression in adjacent tissues, with an average fold increase of 10.55 (P=0.006). Additionally, a high level of PRNCR1 was associated with large tumor volume (Pline (FHC), PRNCR1 was upregulated in most CRC cell lines (HCT116, SW480, LoVo and HT-29). After knockdown of PRNCR1 by ASO, CRC cell proliferation ability was significantly inhibited. We further found that PRNCR1 knockdown induced cell cycle arrest in the G0/G1 phase and a significant decrease in the proportion of cells in the S phases. In contrast, PRNCR1 knockdown did not affect cell apoptosis or invasive ability. Hence, these data indicate that PRNCR1 promotes the proliferation of CRC cells and is a potential oncogene of CRC.

  3. Non-coding RNA regulation in pathogenic bacteria located inside eukaryotic cells

    Directory of Open Access Journals (Sweden)

    Álvaro D. Ortega

    2014-11-01

    Full Text Available Intracellular bacterial pathogens have evolved distinct lifestyles inside eukaryotic cells. Some pathogens coexist with the infected cell in an obligate intracellular state, whereas others transit between the extracellular and intracellular environment. Adaptation to these intracellular lifestyles is regulated in both space and time. Non-coding small RNAs (sRNAs are post-transcriptional regulatory molecules that fine-tune important processes in bacterial physiology including cell envelope architecture, intermediate metabolism, bacterial communication, biofilm formation and virulence. Recent studies have shown production of defined sRNA species by intracellular bacteria located inside eukaryotic cells. The molecules targeted by these sRNAs and their expression dynamics along the intracellular infection cycle remain, however, poorly characterized. Technical difficulties linked to the isolation of ‘intact’ intracellular bacteria from infected host cells might explain why sRNA regulation in these specialized pathogens is still a largely unexplored field. Transition from the extracellular to the intracellular lifestyle provides an ideal scenario in which regulatory sRNAs are intended to participate; so much work must be done in this direction. This review focuses on sRNAs expressed by intracellular bacterial pathogens during the infection of eukaryotic cells, strategies used with these pathogens to identify sRNAs required for virulence, and the experimental technical challenges associated to this type of studies. We also discuss varied techniques for their potential application to study RNA regulation in intracellular bacterial infections.

  4. CellCODE: a robust latent variable approach to differential expression analysis for heterogeneous cell populations

    OpenAIRE

    Chikina, Maria; Zaslavsky, Elena; Sealfon, Stuart C.

    2015-01-01

    Motivation: Identifying alterations in gene expression associated with different clinical states is important for the study of human biology. However, clinical samples used in gene expression studies are often derived from heterogeneous mixtures with variable cell-type composition, complicating statistical analysis. Considerable effort has been devoted to modeling sample heterogeneity, and presently, there are many methods that can estimate cell proportions or pure cell-type expression from m...

  5. Analysis of dysregulated long non-coding RNA expressions in glioblastoma cells.

    Science.gov (United States)

    Balci, Tugce; Yilmaz Susluer, Sunde; Kayabasi, Cagla; Ozmen Yelken, Besra; Biray Avci, Cigir; Gunduz, Cumhur

    2016-09-15

    Long non coding RNAs (lncRNAs) are associated with various biological roles such as embryogenesis, stem cell biology, cellular development and present specific tissue expression profiles. Aberrant expression of lncRNAs are thought to play a critical role in the progression and development of various cancer types, including gliomas. Glioblastomas (GBM) are common and malignant primary brain tumours. Brain cancer stem cells (BCSC) are isolated from both low and high-grade tumours in adults and children, by cell fraction which express neuronal stem cell surface marker CD133. The purpose of this study was to investigate the expression profiles of lncRNAs in brain tumour cells and determine its potential biological function. For this purpose, U118MG-U87MG; GBM stem cell series were used. Human parental brain cancer cells were included as the control group; the expressions of disease related human lncRNA profiles were studied by LightCycler 480 real-time PCR. Expression profiles of 83 lncRNA genes were analyzed for a significant dysregulation, compared to the control cells. Among lncRNAs, 51 lncRNA genes down-regulated, while 8 lncRNA genes were up-regulated. PCAT-1 (-2.36), MEG3 (-5.34), HOTAIR (-2.48) lncRNAs showed low expression in glioblastoma compared to the human (parental) brain cancer stem cells, indicating their role as tumour suppressor genes on gliomas. As a result, significant changes for anti-cancer gene expressions were detected with disease-related human lncRNA array plates. Identification of novel target genes may lead to promising developments in human brain cancer treatment. PMID:27306825

  6. Leap frog integrator modifications in highly collisional particle-in-cell codes

    Science.gov (United States)

    Hanzlikova, N.; Turner, M. M.

    2014-07-01

    Leap frog integration method is a standard, simple, fast, and accurate way to implement velocity and position integration in particle-in-cell codes. Due to the direct solution of kinetics of particles in phase space central to the particle-in-cell procedure, important information can be obtained on particle velocity distributions, and consequently on transport and heating processes. This approach is commonly associated with physical situations where collisional effects are weak, but can also be profitably applied in some highly collisional cases, such as occur in semiconductor devices and gaseous discharges at atmospheric pressure. In this paper, we show that the implementation of the leap frog integration method in these circumstances can violate some of the assumptions central to the accuracy of this scheme. Indeed, without adaptation, the method gives incorrect results. We show here how the method must be modified to deal correctly with highly collisional cases.

  7. A Particle In Cell code development for high current ion beam transport and plasma simulations

    CERN Document Server

    Joshi, N

    2016-01-01

    A simulation package employing a Particle in Cell (PIC) method is developed to study the high current beam transport and the dynamics of plasmas. This package includes subroutines those are suited for various planned projects at University of Frankfurt. In the framework of the storage ring project (F8SR) the code was written to describe the beam optics in toroidal magnetic fields. It is used to design an injection system for a ring with closed magnetic field lines. The generalized numerical model, in Cartesian coordinates is used to describe the intense ion beam transport through the chopper system in the low energy beam section of the FRANZ project. Especially for the chopper system, the Poisson equation is implemented with irregular geometries. The Particle In Cell model is further upgraded with a Monte Carlo Collision subroutine for simulation of plasma in the volume type ion source.

  8. Improved modeling of relativistic collisions and collisional ionization in particle-in-cell codes

    Energy Technology Data Exchange (ETDEWEB)

    Perez, F. [LULI, Ecole Polytechnique, 91128 Palaiseau Cedex (France); CEA, DAM, DIF, F-91297 Arpajon (France); Gremillet, L.; Decoster, A.; Drouin, M.; Lefebvre, E. [CEA, DAM, DIF, F-91297 Arpajon (France)

    2012-08-15

    An improved Monte Carlo collisional scheme modeling both elastic and inelastic interactions has been implemented into the particle-in-cell code CALDER[E. Lefebvre et al., Nucl. Fusion 43, 629 (2003)]. Based on the technique proposed by Nanbu and Yonemura [J. Comput. Phys. 145, 639 (1998)] allowing to handle arbitrarily weighted macro-particles, this binary collision scheme uses a more compact and accurate relativistic formulation than the algorithm recently worked out by Sentoku and Kemp [J. Comput. Phys. 227, 6846 (2008)]. Our scheme is validated through several test cases, demonstrating, in particular, its capability of modeling the electrical resistivity and stopping power of a solid-density plasma over a broad parameter range. A relativistic collisional ionization scheme is developed within the same framework, and tested in several physical scenarios. Finally, our scheme is applied in a set of integrated particle-in-cell simulations of laser-driven fast electron transport.

  9. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

    Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

  10. Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

    Science.gov (United States)

    El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

    2008-09-01

    Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

  11. FADDEEV: A fortran code for the calculation of the frequency response matrix of multiple-input, multiple-output dynamic systems

    International Nuclear Information System (INIS)

    The KDF9/EGDON programme FADDEEV has been written to investigate a technique for the calculation of the matrix of frequency responses G(jw) describing the response of the output vector y from the multivariable differential/algebraic system S to the drive of the system input vector u. S: Ex = Ax + Bu, y = Cx, G(jw) = C(jw E - A )-1 B. The programme uses an algorithm due to Faddeev and has been written with emphasis upon: (a) simplicity of programme structure and computational technique which should enable a user to find his way through the programme fairly easily, and hence facilitate its manipulation as a subroutine in a larger code; (b) rapid computational ability, particularly in systems with fairly large number of inputs and outputs and requiring the evaluation of the frequency responses at a large number of frequencies. Transport or time delays must be converted by the user to Pade or Bode approximations prior to input. Conditions under which the algorithm fails to give accurate results are identified, and methods for increasing the accuracy of the calculations are discussed. The conditions for accurate results using FADDEEV indicate that its application is specialized. (author)

  12. On Calculating the Current-Voltage Characteristic of Multi-Diode Models for Organic Solar Cells

    CERN Document Server

    Roberts, Ken

    2016-01-01

    We provide an alternative formulation of the exact calculation of the current-voltage characteristic of solar cells which have been modeled with a lumped parameters equivalent circuit with one or two diodes. Such models, for instance, are suitable for describing organic solar cells whose current-voltage characteristic curve has an inflection point, also known as an S-shaped anomaly. Our formulation avoids the risk of numerical overflow in the calculation. It is suitable for implementation in Fortran, C or on micro-controllers.

  13. Non-coding RNA as mediators in microenvironment-breast cancer cell communication.

    Science.gov (United States)

    Patel, Jimmy S; Hu, Madeleine; Sinha, Garima; Walker, Nykia D; Sherman, Lauren S; Gallagher, Ashley; Rameshwar, Pranela

    2016-09-28

    The tumor microenvironment has a critical role in the survival and decision of the cancer cells. These include support by enhanced angiogenesis, and metastasis or adaptation of dormancy. This article discusses methods by which the microenvironment sustains the tumor. This process is important as it will identify avenues of drug targets. Non-coding RNAs (ncRNAs) are evolving as key mediators in the interaction between the cancer cells and the microenvironment. Thus, the question is how to develop methods to effectively block the effects of the ncRNA and/or to introduce them to prevent metastasis, dormancy or to reverse dormancy. We focused on the advantages of using mesenchymal stem cells (MSCs) for RNA delivery. MSCs can be available as "off-the-shelf" cells. Thus far, MSCs are shown to be safe when transplanted across allogeneic barriers. We discussed the various methods by which MSCs can interact with cancer cells to deliver ncRNA or antagomirs. We also include the advances and possible confounds of using these methods. Overall, this review article provides a potential method by which MSCs can be used for effective delivery of nucleic acid to treat cancer. PMID:26582656

  14. European inter-comparison of Monte Carlo codes users for the uncertainty calculation of the kerma in air beside a caesium-137 source; Intercomparaison europeenne d'utilisateurs de codes monte carlo pour le calcul d'incertitudes sur le kerma dans l'air aupres d'une source de cesium-137

    Energy Technology Data Exchange (ETDEWEB)

    De Carlan, L.; Bordy, J.M.; Gouriou, J. [CEA Saclay, LIST, Laboratoire National Henri Becquerel, Laboratoire de Metrologie de la Dose 91 - Gif-sur-Yvette (France)

    2010-07-01

    Within the frame of the CONRAD European project (Coordination Network for Radiation Dosimetry), and more precisely within a work group paying attention to uncertainty assessment in computational dosimetry and aiming at comparing different approaches, the authors report the simulation of an irradiator containing a caesium 137 source to calculate the kerma in air as well as its uncertainty due to different parameters. They present the problem geometry, recall the studied issues (kerma uncertainty, influence of capsule source, influence of the collimator, influence of the air volume surrounding the source). They indicate the codes which have been used (MNCP, Fluka, Penelope, etc.) and discuss the obtained results for the first issue

  15. Optimising PICCANTE -- an open source particle-in-cell code for advanced simulations on Tier-0 systems

    CERN Document Server

    Sgattoni, Andrea; Sinigardi, Stefano; Marocchino, Alberto; Macchi, Andrea; Weinberg, Volker; Karmakar, Anupam

    2015-01-01

    We discuss a detailed strong and weak scaling analysis of PICCANTE, an open source, massively parallel, fully-relativistic Particle-In-Cell (PIC) code. PIC codes are widely used in plasma physics and astrophysics to study the cases where kinetic effects are relevant. PICCANTE is primarily developed to study laser-plasma interaction. Within a PRACE Preparatory Access Project, various revisions of different routines of the code have been analysed on the HPC systems JUQUEEN at J\\"ulich Supercomputing Centre (JSC), Germany, and FERMI at CINECA, Italy, to improve the parallel scalability and the I/O performance of the application. The diagnostic tool Scalasca is used to filter out suboptimal routines. Different output strategies are discussed. The detailed strong and weak scaling behaviour of the improved code is presented in comparison with the original version of the code.

  16. Non-coding RNAs as epigenetic regulator of glioma stem-like cell differentiation

    Directory of Open Access Journals (Sweden)

    Keisuke eKatsushima

    2014-02-01

    Full Text Available Glioblastomas show heterogeneous histological features. These distinct phenotypic states are thought to be associated with the presence of glioma stem cells (GSCs, which are highly tumorigenic and self-renewing sub-population of tumor cells that have different functional characteristics. Differentiation of GSCs may be regulated by multi-tiered epigenetic mechanisms that orchestrate the expression of thousands of genes. One such regulatory mechanism involves functional non-coding RNAs (ncRNAs, such as microRNAs (miRNAs; a large number of ncRNAs have been identified and shown to regulate the expression of genes associated with cell differentiation programs. Given the roles of miRNAs in cell differentiation, it is possible they are involved in the regulation of gene expression networks in GSCs that are important for the maintenance of the pluripotent state and for directing differentiation. Here, we review recent findings on ncRNAs associated with GSC differentiation and discuss how these ncRNAs contribute to the establishment of tissue heterogeneity during glioblastoma tumor formation.

  17. Mechanisms and benefits of granule cell latency coding in the mouse olfactory bulb

    Directory of Open Access Journals (Sweden)

    Sonya eGiridhar

    2012-06-01

    Full Text Available Inhibitory circuits are critical for shaping odor representations in the olfactory bulb. Individual olfactory bulb granule cells can respond to brief stimulation with extremely long (up to 1000 ms, input-specific latencies that are highly reliable. However, the mechanism and function of this long timescale activity remain unknown. We sought to elucidate the mechanism responsible for long-latency activity, and to understand the impact of widely distributed interneuron latencies on olfactory coding. We used a combination of electrophysiological, optical, and pharmacological techniques to show that long-latency inhibition is driven by late onset synaptic excitation to granule cells. We also provide evidence that tufted cells, which have intrinsic properties that favor longer latency spiking than mitral cells are the major source of this excitation. Using computational modeling, we show that widely distributed interneuron latency can increase the discriminability of similar stimuli. Thus, long-latency inhibition in the olfactory bulb requires a combination of circuit- and cellular-level mechanisms that function to improve stimulus representations.

  18. Automatic choroid cells segmentation and counting based on approximate convexity and concavity of chain code in fluorescence microscopic image

    Science.gov (United States)

    Lu, Weihua; Chen, Xinjian; Zhu, Weifang; Yang, Lei; Cao, Zhaoyuan; Chen, Haoyu

    2015-03-01

    In this paper, we proposed a method based on the Freeman chain code to segment and count rhesus choroid-retinal vascular endothelial cells (RF/6A) automatically for fluorescence microscopy images. The proposed method consists of four main steps. First, a threshold filter and morphological transform were applied to reduce the noise. Second, the boundary information was used to generate the Freeman chain codes. Third, the concave points were found based on the relationship between the difference of the chain code and the curvature. Finally, cells segmentation and counting were completed based on the characteristics of the number of the concave points, the area and shape of the cells. The proposed method was tested on 100 fluorescence microscopic cell images, and the average true positive rate (TPR) is 98.13% and the average false positive rate (FPR) is 4.47%, respectively. The preliminary results showed the feasibility and efficiency of the proposed method.

  19. A highly optimized code for calculating atomic data at neutron star magnetic field strengths using a doubly self-consistent Hartree-Fock-Roothaan method

    Science.gov (United States)

    Schimeczek, C.; Engel, D.; Wunner, G.

    2012-07-01

    Our previously published code for calculating energies and bound-bound transitions of medium-Z elements at neutron star magnetic field strengths [D. Engel, M. Klews, G. Wunner, Comput. Phys. Comm. 180 (2009) 302-311] was based on the adiabatic approximation. It assumes a complete decoupling of the (fast) gyration of the electrons under the action of the magnetic field and the (slow) bound motion along the field under the action of the Coulomb forces. For the single-particle orbitals this implied that each is a product of a Landau state and an (unknown) longitudinal wave function whose B-spline coefficients were determined self-consistently by solving the Hartree-Fock equations for the many-electron problem on a finite-element grid. In the present code we go beyond the adiabatic approximation, by allowing the transverse part of each orbital to be a superposition of Landau states, while assuming that the longitudinal part can be approximated by the same wave function in each Landau level. Inserting this ansatz into the energy variational principle leads to a system of coupled equations in which the B-spline coefficients depend on the weights of the individual Landau states, and vice versa, and which therefore has to be solved in a doubly self-consistent manner. The extended ansatz takes into account the back-reaction of the Coulomb motion of the electrons along the field direction on their motion in the plane perpendicular to the field, an effect which cannot be captured by the adiabatic approximation. The new code allows for the inclusion of up to 8 Landau levels. This reduces the relative error of energy values as compared to the adiabatic approximation results by typically a factor of three (1/3 of the original error), and yields accurate results also in regions of lower neutron star magnetic field strengths where the adiabatic approximation fails. Further improvements in the code are a more sophisticated choice of the initial wave functions, which takes into

  20. Electrochemical Impedance Spectra of Dye-Sensitized Solar Cells: Fundamentals and Spreadsheet Calculation

    Directory of Open Access Journals (Sweden)

    Subrata Sarker

    2014-01-01

    Full Text Available Electrochemical impedance spectroscopy (EIS is one of the most important tools to elucidate the charge transfer and transport processes in various electrochemical systems including dye-sensitized solar cells (DSSCs. Even though there are many books and reports on EIS, it is often very difficult to explain the EIS spectra of DSSCs. Understanding EIS through calculating EIS spectra on spreadsheet can be a powerful approach as the user, without having any programming knowledge, can go through each step of calculation on a spreadsheet and get instant feedback by visualizing the calculated results or plot on the same spreadsheet. Here, a brief account of the EIS of DSSCs is given with fundamental aspects and their spreadsheet calculation. The review should help one to develop a basic understanding about EIS of DSSCs through interacting with spreadsheet.

  1. Calculated performance of p(+)n InP solar cells with In(0.52)Al(0.48)As window layers

    Science.gov (United States)

    Jain, R. K.; Landis, G. A.

    1991-01-01

    The performance of indium phosphide solar cells with lattice matched wide band-gap In(0.52)Al(0.48)As window layers was calculated using the PC-1D computer code. The conversion efficiency of p(+)n InP solar cells is improved significantly by the window layer. No improvement is seen for n(+)p structures. The improvement in InP cell efficiency was studied as a function of In(0.52)Al(0.48)As layer thickness. The use of the window layer improves both the open circuit voltage and short circuit current.For a typical In(0.52)Al(0.48)As window layer thickness of 20 nm, the cell efficiency improves in excess of 27 percent to a value of 18.74 percent.

  2. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  3. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors; Validacion del sistema de codigos WIMS-SNAP para calculos en reactores nucleares tipo TRIGA-MARK II

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu

    2000-07-01

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  4. Numerical simulation of Ge solar cells using D-AMPS-1D code

    Energy Technology Data Exchange (ETDEWEB)

    Barrera, Marcela, E-mail: barrera@tandar.cnea.gov.ar [Comision Nacional de Energia Atomica, Avenida General Paz 1499, San Martin 1650, Buenos Aires (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina); Rubinelli, Francisco [Instituto de Desarrollo Tecnologico para la Industria Quimica (INTEC)-CONICET, Gueemes 3450, Santa Fe 3000 (Argentina); Rey-Stolle, Ignacio [Instituto de Energia Solar, Universidad Politecnica de Madrid, Avenida Complutense 30, Madrid 28040 (Spain); Pla, Juan [Comision Nacional de Energia Atomica, Avenida General Paz 1499, San Martin 1650, Buenos Aires (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina)

    2012-08-15

    A solar cell is a solid state device that converts the energy of sunlight directly into electricity by the photovoltaic effect. When light with photon energies greater than the band gap is absorbed by a semiconductor material, free electrons and free holes are generated by optical excitation in the material. The main characteristic of a photovoltaic device is the presence of internal electric field able to separate the free electrons and holes so they can pass out of the material to the external circuit before they recombine. Numerical simulation of photovoltaic devices plays a crucial role in their design, performance prediction, and comprehension of the fundamental phenomena ruling their operation. The electrical transport and the optical behavior of the solar cells discussed in this work were studied with the simulation code D-AMPS-1D. This software is an updated version of the one-dimensional (1D) simulation program Analysis of Microelectronic and Photonic Devices (AMPS) that was initially developed at The Penn State University, USA. Structures such as homojunctions, heterojunctions, multijunctions, etc., resulting from stacking layers of different materials can be studied by appropriately selecting characteristic parameters. In this work, examples of cells simulation made with D-AMPS-1D are shown. Particularly, results of Ge photovoltaic devices are presented. The role of the InGaP buffer on the device was studied. Moreover, a comparison of the simulated electrical parameters with experimental results was performed.

  5. On tentative decommissioning cost analysis with specific authentic cost calculations with the application of the Omega code on a case linked to the Intermediate storage facility for spent fuel in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Vasko, Marek; Daniska, Vladimir; Ondra, Frantisek; Bezak, Peter; Kristofova, Kristina; Tatransky, Peter; Zachar, Matej [DECOM Slovakia, spol. s.r.o., J. Bottu 2, SK-917 01 Trnava (Slovakia); Lindskog, Staffan [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2007-03-15

    The presented report is focused on tentative calculations of basic decommissioning parameters such as costs, manpower and exposure of personnel for activities of older nuclear facility decommissioning in Sweden represented by Intermediate storage facility for spent fuel in Studsvik, by means of calculation code OMEGA. This report continuously follows up two previous projects, which described methodology of cost estimates of decommissioning with an emphasis to derive cost functions for alpha contaminated material and implementation of the advanced decommissioning costing methodology for Intermediate Storage facility for Spent Fuel in Studsvik. The main purpose of the presented study is to demonstrate the trial application of the advanced costing methodology using OMEGA code for Intermediate Storage Facility for Spent Fuel in Studsvik. Basic work packages presented in report are as follows: 1. Analysis and validation input data on Intermediate Storage Facility for Spent Fuel and assemble a database suitable for standardised decommissioning cost calculations including radiological parameters, 2. Proposal of range of decommissioning calculations and define an extent of decommissioning activities, 3. Defining waste management scenarios for particular material waste streams from Intermediate Storage Facility for Spent Fuel, 4. Developing standardised cost calculation structure applied for Intermediate Storage Facility for Spent Fuel decommissioning calculation and 5. Performing tentative decommissioning calculations for Intermediate Storage Facility for Spent Fuel by OMEGA code. Calculated parameters of decommissioning are presented in structure according to Proposed Standardized List of Items for Costing Purposes. All parameters are documented and summed up in both table and graphic forms in text and Annexes. The presented report documents availability and applicability of methodology for evaluation of costs and other parameters of decommissioning in a form implemented

  6. Benchmarking of Decay Heat Measured Values of ITER Materials Induced by 14 MeV Neutron Activation with Calculated Results by ACAB Activation Code

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Ortego, P.; Rodriguez Rivada, A.

    2014-07-01

    The aim of this paper is the comparison between the calculated and measured decay heat of material samples which were irradiated at the Fusion Neutron Source of JAERI in Japan with D-T production of 14MeV neutrons. In the International Thermonuclear Experimental Reactor (ITER) neutron activation of the structural material will result in a source of heat after shutdown of the reactor. The estimation of decay heat value with qualified codes and nuclear data is an important parameter for the safety analyses of fusion reactors against lost of coolant accidents. When a loss of coolant and/or flow accident happen plasma facing components are heated up by decay heat. If the temperature of the components exceeds the allowable temperature, the accident would expand to loose the integrity of ITER. Uncertainties associated with decay prediction less than 15% are strongly requested by the ITER designers. Additionally, accurate decay heat prediction is required for making reasonable shutdown scenarios of ITER. (Author)

  7. Calculation of extrapolation curves in the 4π(LS)β-γ coincidence technique with the Monte Carlo code Geant4.

    Science.gov (United States)

    Bobin, C; Thiam, C; Bouchard, J

    2016-03-01

    At LNE-LNHB, a liquid scintillation (LS) detection setup designed for Triple to Double Coincidence Ratio (TDCR) measurements is also used in the β-channel of a 4π(LS)β-γ coincidence system. This LS counter based on 3 photomultipliers was first modeled using the Monte Carlo code Geant4 to enable the simulation of optical photons produced by scintillation and Cerenkov effects. This stochastic modeling was especially designed for the calculation of double and triple coincidences between photomultipliers in TDCR measurements. In the present paper, this TDCR-Geant4 model is extended to 4π(LS)β-γ coincidence counting to enable the simulation of the efficiency-extrapolation technique by the addition of a γ-channel. This simulation tool aims at the prediction of systematic biases in activity determination due to eventual non-linearity of efficiency-extrapolation curves. First results are described in the case of the standardization (59)Fe. The variation of the γ-efficiency in the β-channel due to the Cerenkov emission is investigated in the case of the activity measurements of (54)Mn. The problem of the non-linearity between β-efficiencies is featured in the case of the efficiency tracing technique for the activity measurements of (14)C using (60)Co as a tracer.

  8. Calculation of cell face velocity of non-staggered grid system

    KAUST Repository

    Li, Wang

    2012-07-28

    In this paper, the cell face velocities in the discretization of the continuity equation, the momentum equation, and the scalar equation of a non-staggered grid system are calculated and discussed. Both the momentum interpolation and the linear interpolation are adopted to evaluate the coefficients in the discretized momentum and scalar equations. Their performances are compared. When the linear interpolation is used to calculate the coefficients, the mass residual term in the coefficients must be dropped to maintain the accuracy and convergence rate of the solution. © Shanghai University and Springer-Verlag Berlin Heidelberg 2012.

  9. Development and implementation in the Monte Carlo code PENELOPE of a new virtual source model for radiotherapy photon beams and portal image calculation

    Science.gov (United States)

    Chabert, I.; Barat, E.; Dautremer, T.; Montagu, T.; Agelou, M.; Croc de Suray, A.; Garcia-Hernandez, J. C.; Gempp, S.; Benkreira, M.; de Carlan, L.; Lazaro, D.

    2016-07-01

    This work aims at developing a generic virtual source model (VSM) preserving all existing correlations between variables stored in a Monte Carlo pre-computed phase space (PS) file, for dose calculation and high-resolution portal image prediction. The reference PS file was calculated using the PENELOPE code, after the flattening filter (FF) of an Elekta Synergy 6 MV photon beam. Each particle was represented in a mobile coordinate system by its radial position (r s ) in the PS plane, its energy (E), and its polar and azimuthal angles (φ d and θ d ), describing the particle deviation compared to its initial direction after bremsstrahlung, and the deviation orientation. Three sub-sources were created by sorting out particles according to their last interaction location (target, primary collimator or FF). For each sub-source, 4D correlated-histograms were built by storing E, r s , φ d and θ d values. Five different adaptive binning schemes were studied to construct 4D histograms of the VSMs, to ensure histogram efficient handling as well as an accurate reproduction of E, r s , φ d and θ d distribution details. The five resulting VSMs were then implemented in PENELOPE. Their accuracy was first assessed in the PS plane, by comparing E, r s , φ d and θ d distributions with those obtained from the reference PS file. Second, dose distributions computed in water, using the VSMs and the reference PS file located below the FF, and also after collimation in both water and heterogeneous phantom, were compared using a 1.5%-0 mm and a 2%-0 mm global gamma index, respectively. Finally, portal images were calculated without and with phantoms in the beam. The model was then evaluated using a 1%-0 mm global gamma index. Performance of a mono-source VSM was also investigated and led, as with the multi-source model, to excellent results when combined with an adaptive binning scheme.

  10. Development and implementation in the Monte Carlo code PENELOPE of a new virtual source model for radiotherapy photon beams and portal image calculation.

    Science.gov (United States)

    Chabert, I; Barat, E; Dautremer, T; Montagu, T; Agelou, M; Croc de Suray, A; Garcia-Hernandez, J C; Gempp, S; Benkreira, M; de Carlan, L; Lazaro, D

    2016-07-21

    This work aims at developing a generic virtual source model (VSM) preserving all existing correlations between variables stored in a Monte Carlo pre-computed phase space (PS) file, for dose calculation and high-resolution portal image prediction. The reference PS file was calculated using the PENELOPE code, after the flattening filter (FF) of an Elekta Synergy 6 MV photon beam. Each particle was represented in a mobile coordinate system by its radial position (r s ) in the PS plane, its energy (E), and its polar and azimuthal angles (φ d and θ d ), describing the particle deviation compared to its initial direction after bremsstrahlung, and the deviation orientation. Three sub-sources were created by sorting out particles according to their last interaction location (target, primary collimator or FF). For each sub-source, 4D correlated-histograms were built by storing E, r s , φ d and θ d values. Five different adaptive binning schemes were studied to construct 4D histograms of the VSMs, to ensure histogram efficient handling as well as an accurate reproduction of E, r s , φ d and θ d distribution details. The five resulting VSMs were then implemented in PENELOPE. Their accuracy was first assessed in the PS plane, by comparing E, r s , φ d and θ d distributions with those obtained from the reference PS file. Second, dose distributions computed in water, using the VSMs and the reference PS file located below the FF, and also after collimation in both water and heterogeneous phantom, were compared using a 1.5%-0 mm and a 2%-0 mm global gamma index, respectively. Finally, portal images were calculated without and with phantoms in the beam. The model was then evaluated using a 1%-0 mm global gamma index. Performance of a mono-source VSM was also investigated and led, as with the multi-source model, to excellent results when combined with an adaptive binning scheme. PMID:27353090

  11. Open Photoacoustic Cell for Blood Sugar Measurement: Numerical Calculation of Frequency Response

    CERN Document Server

    Baumann, Bernd; Teschner, Mark

    2015-01-01

    A new approach for continuous and non-invasive monitoring of the glucose concentration in human epidermis has been suggested recently. This method is based on photoacoustic (PA) analysis of human interstitial fluid. The measurement can be performed in vitro and in vivo and, therefore, may form the basis for a non-invasive monitoring of the blood sugar level for diabetes patients. It requires a windowless PA cell with an additional opening that is pressed onto the human skin. Since signals are weak, advantage is taken of acoustic resonances of the cell. Recently, a numerical approach based on the Finite Element (FE) Method has been successfully used for the calculation of the frequency response function of closed PA cells. This method has now been adapted to obtain the frequency response of the open cell. Despite the fact that loss due to sound radiation at the opening is not included, fairly good accordance with measurement is achieved.

  12. Particle-in-cell vs straight line Gaussian calculations for an area of complex topography

    International Nuclear Information System (INIS)

    Two numerical models for the calculation of time integrated air concentraton and ground deposition of airborne radioactive effluent releases are compared. The time dependent Particle-in-Cell (PIC) model and the steady state Gaussian plume model were used for the simulation. The area selected for the comparison was the Hudson River Valley, New York. Input for the models was synthesized from meteorological data gathered in previous studies by various investigators. It was found that the PIC model more closely simulated the three-dimensional effects of the meteorology and topography. Overall, the Gaussian model calculated higher concentrations under stable conditions. In addition, because of its consideration of exposure from the returning plume after flow reversal, the PIC model calculated air concentrations over larger areas than did the Gaussian model

  13. Application of the particle-in-cell method in propagation calculations

    International Nuclear Information System (INIS)

    The Particle-in-Cell-Method that is capable of calculating the spreading of a plume in the atmosphere under instationary and inhomogeneous conditions, has a systematical advantage over the steady state Gaussian plume model usually used. Especially the fixed-point concentration time integral is calculated realistically instead of the locally integrated concentration at a constant time as is done in the plume model. Inaccuracies due to the computational techniques may be avoided in this way. On the other hand, at first the turbulent diffusion coefficients that describe the diffusion in the particle-in-cell method, must be prepared for all diffusion types. Thereby the diffusion coefficients can be seen to be mainly deduced in the steady state. This is one reason why they cannot be used in an optimal sense in a model that actually works instationary. (orig.)

  14. Inter and intra macro-cell model for point dipole–dipole energy calculations

    International Nuclear Information System (INIS)

    In the field of micromagnetics, the calculation of long-range dipole–dipole interactions in non-uniformly magnetized bodies has long posed computational problems. In this paper, we present an inter and intra macro-cell point-dipole model, which can be used to speed-up the determination of dipole–dipole energies at the atomistic level. The model can be used to accurately compute the dipole–dipole energy, using macro-cells of any shape or size. (paper)

  15. PN solutions for the slowing-down and the cell calculation problems in plane geometry

    International Nuclear Information System (INIS)

    In this work PN solutions for the slowing-down and cell problems in slab geometry are developed. To highlight the main contributions of this development, one can mention: the new particular solution developed for the PN method applied to the slowing-down problem in the multigroup model, originating a new class of polynomials denominated Chandrasekhar generalized polynomials; the treatment of a specific situation, known as a degeneracy, arising from a particularity in the group constants and the first application of the PN method, for arbitrary N, in criticality calculations at the cell level reported in literature. (author)

  16. The Non-Coding RNA Llme23 Drives the Malignant Property of Human Melanoma Cells

    Institute of Scientific and Technical Information of China (English)

    Chuan-Fang Wu; Guang-Hong Tan; Cheng-Chuan Ma; Ling Li

    2013-01-01

    Several lines of evidence support the notion that increased RNA-binding ability of polypyrimidine tract-binding (PTB) proteinassociated splicing factor (PSF) and aberrant expression of long non-coding RNAs (lncRNAs) are associated with mouse and human tumors.To identify the PSF-binding IncRNA involved in human oncogenesis,we screened a nuclear RNA repertoire of human melanoma cell line,YUSAC,through RNA-SELEX affinity chromatography.A previously unreported lncRNA,termed as Lime23,was found to bind immobilized PSF resin.The specific binding of Llme23 to both recombinant and native PSF protein was confirmed in vitro and in vivo.The expression of PSF-binding Llme23 is exclusively detected in human melanoma lines.Knocking down Lime23 remarkably suppressed the malignant property of YUSAC cells,accompanied by the repressed expression of proto-oncogene Rab23.These results may indicate that Llme23 can function as an oncogenic RNA and directly associate the PSF-binding IncRNA with human melanoma.

  17. Long non-coding RNA Loc554202 regulates proliferation and migration in breast cancer cells

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Yongguo, E-mail: 1138303166@qq.com [Department of Oncology, Second Affiliated Hospital, Nanjing Medical University, Nanjing, Jiangsu (China); Lu, Jianwei, E-mail: jianwei2010077@163.com [Cancer Hospital of Jiangsu Province, Nanjing, Jiangsu (China); Zhou, Jing, E-mail: 2310848@163.com [Department of Oncology, Taizhou People’ Hospital, Taizhou, Jiangsu (China); Tan, Xueming, E-mail: 843039795@qq.com [Department of Gastroenterology, Second Affiliated Hospital, Nanjing Medical University, Nanjing, Jiangsu (China); He, Ye, E-mail: 2825636@qq.com [Department of Oncology, Second Affiliated Hospital, Nanjing Medical University, Nanjing, Jiangsu (China); Ding, Jie, E-mail: 9111165@qq.com [Department of Oncology, Second Affiliated Hospital, Nanjing Medical University, Nanjing, Jiangsu (China); Tian, Yun, E-mail: 1815857@qq.com [Department of Oncology, Second Affiliated Hospital, Nanjing Medical University, Nanjing, Jiangsu (China); Wang, Li, E-mail: 2376737@qq.com [Department of Oncology, Second Affiliated Hospital, Nanjing Medical University, Nanjing, Jiangsu (China); Wang, Keming, E-mail: wkmys@sohu.com [Department of Oncology, Second Affiliated Hospital, Nanjing Medical University, Nanjing, Jiangsu (China)

    2014-04-04

    Highlights: • First, we have shown that upregulated of the Loc554202 in breast cancer tissues. • Second, we demonstrated the function of Loc554202 in breast cancer cell. • Finally, we demonstrated that LOC554202 knockdown could inhibit tumor growth in vivo. - Abstract: Data derived from massive cloning and traditional sequencing methods have revealed that long non-coding RNAs (lncRNA) play important roles in the development and progression of cancer. Although many studies suggest that the lncRNAs have different cellular functions, many of them are not yet to be identified and characterized for the mechanism of their functions. To address this question, we assay the expression level of lncRNAs–Loc554202 in breast cancer tissues and find that Loc554202 is significantly increased compared with normal control, and associated with advanced pathologic stage and tumor size. Moreover, knockdown of Loc554202 decreased breast cancer cell proliferation, induced apoptosis and inhibits migration/invasion in vitro and impeded tumorigenesis in vivo. These data suggest an important role of Loc554202 in breast tumorigenesis.

  18. Turbo Codes Extended with Outer BCH Code

    DEFF Research Database (Denmark)

    Andersen, Jakob Dahl

    1996-01-01

    The "error floor" observed in several simulations with the turbo codes is verified by calculation of an upper bound to the bit error rate for the ensemble of all interleavers. Also an easy way to calculate the weight enumerator used in this bound is presented. An extended coding scheme is proposed...

  19. Calculation of the Energy Band Diagram of a Photoelectrochemical Water Splitting Cell

    OpenAIRE

    Cendula, Peter; Tilley, S. David; Girnenez, Sixto; Bisquert, Juan; Schmid, Matthias; Graetzel, Michael; Schumachert, Juergen O.

    2014-01-01

    A physical model is presented for the semiconductor electrode of a photoelectrochemical cell. The model accounts for the potential drop in the Helmholtz layer and thus enables description of both band edge pinning and unpinning. The model is based on the continuity equations for charge carriers and direct charge transfer from the energy bands to the electrolyte. A quantitative calculation of the position of the energy bands and the variation of the quasi-Fermi levels in the semiconductor with...

  20. Identification of long non-coding RNA involved in osteogenic differentiation from mesenchymal stem cells using RNA-Seq data.

    Science.gov (United States)

    Song, W Q; Gu, W Q; Qian, Y B; Ma, X; Mao, Y J; Liu, W J

    2015-01-01

    The aim of this study was to identify long non-coding RNA (lncRNA) associated with osteogenic differentiation from mesenchymal stem cells (MSCs) using high-throughput RNA sequencing (RNA-Seq) data. RNA-Seq dataset was obtained from the European Bioinformatics Institute (accession No. PRJEB4496), including two replicates each for immortalized mesenchymal stem cells iMSC#3 cultured in growth medium (GM) and differentiation medium (DM) for 28 days. The clean reads were aligned to a hg19 reference genome by Tophat and assembled by Cufflinks to identify the known and novel transcripts. RPKM values were calculated to screen for differentially expressed RNA. Novel lncRNA were screened based on various filter criteria. Subsequently, the underlying function of novel lncRNAs were predicted by functional annotation by ERPIN, a co-expression network was constructed by WGCNA and the KEGG pathway enriched by KOBAS. A total of 3171 RNA differentially expressed between the DM and GM groups (2597 mRNA and 574 lncRNA) were identified. Among the 574 differentially expressed lncRNA, 357 were known and 217 were novel lncRNA. Furthermore, 32 novel lncRNA were found to be miRNA precursors (including miR-689, miR-640, miR-601, and miR-544). A total of 14,275 co-expression relationships and 217 co-expression networks were obtained between novel lncRNA and mRNA. The differentially expressed lncRNA and mRNA were enriched into 6 significant pathways, including those for cancer, ECM-receptor interaction, and focal adhesion. Therefore, novel lncRNAwere identified and their underlying function predicted, which may provide the basis for future analyses of the role of lncRNA in osteoblastic differentiation.

  1. theory and calculation of the design of nuclear reactor

    International Nuclear Information System (INIS)

    For the sake of formation of a complete general code for nuclear power reactor design, this thesis deals with a great part of this code. the code links the solution of the neutron integral transport equation by the multigroup treatment (76 energy groups) for the calculation of the reactor cell parameters by the fuel management program that solves the neutron diffusion equation inside a large number of nuclear fuel assemblies. the lattice cell code is modified to accommodate the calculation of lattice cell parameters for more than one enrichment ( one after the other). it is also modified to calculate the burn up parameters using unequal time steps. these two modifications are complicated but necessary for the link between the cell program and fuel management program. the comparison between the results of the fitted cross sections and that given by the cell calculations shows the necessity of using the cell code cross sections. this is also necessary for the sake of generality for any type of reactors. the comparison for the fuel management calculation depending on fitted data and that depending on cell calculation data insures the necessity for using the cell data i.e. insures the necessity of linking the cell calculation program by the fuel management program

  2. Long non-coding RNA regulation of liver cancer stem cell self-renewal offers new therapeutic targeting opportunities

    Science.gov (United States)

    Parasramka, Mansi A.

    2016-01-01

    Long non-coding RNAs (lncRNA) are critical regulators of gene expression, and can reprogram the transcriptome to modulate cellular processes involved in cellular growth and differentiation, and thereby contribute to tumorigenesis. In addition to effects on tumor cell growth, survival and cell signaling, lncRNA can modulate cancer stem cell (CSC) behavior, including the expression of pluripotency factors. The identification of lncRNA that are mechanistically linked to cancer stem cell self-renewal and differentiation, or aberrant signaling pathways associated with tumor growth or progression, offer new opportunities for therapeutic intervention. PMID:27358893

  3. Role of conserved non-coding DNA elements in the Foxp3 gene in regulatory T-cell fate

    OpenAIRE

    Zheng, Ye; Josefowicz, Steven; Chaudhry, Ashutosh; Peng, Xiao P.; Forbush, Katherine; Rudensky, Alexander Y.

    2010-01-01

    Immune homeostasis is dependent on tight control over the size of a population of regulatory T (Treg) cells capable of suppressing over-exuberant immune responses. The Treg cell subset is comprised of cells that commit to the Treg lineage by upregulating the transcription factor Foxp3 either in the thymus (tTreg) or in the periphery (iTreg)1,2. Considering a central role for Foxp3 in Treg cell differentiation and function3,4, we proposed that conserved non-coding DNA sequence (CNS) elements a...

  4. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  5. Monte-Carlo approach to calculate the proton stopping in warm dense matter within particle-in-cell simulations

    CERN Document Server

    Wu, D; Yu, W; Fritzsche, S

    2016-01-01

    A Monte-Carlo approach to proton stopping in warm dense matter is implemented into an existing particle-in-cell code. The model is based on multiple binary-collisions among electron-electron, electron-ion and ion-ion, taking into account contributions from both free and bound electrons, and allows to calculate particle stopping in much more natural manner. At low temperature limit, when ``all'' electron are bounded at the nucleus, the stopping power converges to the predictions of Bethe-Bloch theory, which shows good consistency with data provided by the NIST. With the rising of temperatures, more and more bound electron are ionized, thus giving rise to an increased stopping power to cold matter, which is consistent with the report of a recently experimental measurement [Phys. Rev. Lett. 114, 215002 (2015)]. When temperature is further increased, with ionizations reaching the maximum, lowered stopping power is observed, which is due to the suppression of collision frequency between projected proton beam and h...

  6. Online program ‘vipcal’ for calculating lytic viral production and lysogenic cells based on a viral reduction approach

    OpenAIRE

    Luef, Birgit; Luef, Franz; Peduzzi, Peter

    2009-01-01

    Assessing viral production (VP) requires robust methodological settings combined with precise mathematical calculations. This contribution improves and standardizes mathematical calculations of VP and the assessment of the proportion of lysogenic cells in a sample. We present an online tool ‘Viral Production Calculator’ (vipcal, http://www.univie.ac.at/nuhag-php/vipcal) that calculates lytic production and the percentage of lysogenic cells based on data obtained from a viral reduction approac...

  7. SPENVIS Implementation of End-of-Life Solar Cell Calculations Using the Displacement Damage Dose Methodology

    Science.gov (United States)

    Walters, Robert; Summers, Geoffrey P.; Warmer. Keffreu J/; Messenger, Scott; Lorentzen, Justin R.; Morton, Thomas; Taylor, Stephen J.; Evans, Hugh; Heynderickx, Daniel; Lei, Fan

    2007-01-01

    This paper presents a method for using the SPENVIS on-line computational suite to implement the displacement damage dose (D(sub d)) methodology for calculating end-of-life (EOL) solar cell performance for a specific space mission. This paper builds on our previous work that has validated the D(sub d) methodology against both measured space data [1,2] and calculations performed using the equivalent fluence methodology developed by NASA JPL [3]. For several years, the space solar community has considered general implementation of the D(sub d) method, but no computer program exists to enable this implementation. In a collaborative effort, NRL, NASA and OAI have produced the Solar Array Verification and Analysis Tool (SAVANT) under NASA funding, but this program has not progressed beyond the beta-stage [4]. The SPENVIS suite with the Multi Layered Shielding Simulation Software (MULASSIS) contains all of the necessary components to implement the Dd methodology in a format complementary to that of SAVANT [5]. NRL is currently working with ESA and BIRA to include the Dd method of solar cell EOL calculations as an integral part of SPENVIS. This paper describes how this can be accomplished.

  8. Novel methods in the Particle-In-Cell accelerator Code-Framework Warp

    International Nuclear Information System (INIS)

    The Particle-In-Cell (PIC) Code-Framework Warp is being developed by the Heavy Ion Fusion Science Virtual National Laboratory (HIFS-VNL) to guide the development of accelerators that can deliver beams suitable for high-energy density experiments and implosion of inertial fusion capsules. It is also applied in various areas outside the Heavy Ion Fusion program to the study and design of existing and next-generation high-energy accelerators, including the study of electron cloud effects and laser wakefield acceleration for example. This paper presents an overview of Warp's capabilities, summarizing recent original numerical methods that were developed by the HIFS-VNL (including PIC with adaptive mesh refinement, a large-timestep ‘drift-Lorentz’ mover for arbitrarily magnetized species, a relativistic Lorentz invariant leapfrog particle pusher, simulations in Lorentz-boosted frames, an electromagnetic solver with tunable numerical dispersion and efficient stride-based digital filtering), with special emphasis on the description of the mesh refinement capability. Selected examples of the applications of the methods to the abovementioned fields are given. (paper)

  9. Novel methods in the Particle-In-Cell accelerator Code-Framework Warp

    Energy Technology Data Exchange (ETDEWEB)

    Vay, J-L [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Grote, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Cohen, R. H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Friedman, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2012-12-26

    The Particle-In-Cell (PIC) Code-Framework Warp is being developed by the Heavy Ion Fusion Science Virtual National Laboratory (HIFS-VNL) to guide the development of accelerators that can deliver beams suitable for high-energy density experiments and implosion of inertial fusion capsules. It is also applied in various areas outside the Heavy Ion Fusion program to the study and design of existing and next-generation high-energy accelerators, including the study of electron cloud effects and laser wakefield acceleration for example. This study presents an overview of Warp's capabilities, summarizing recent original numerical methods that were developed by the HIFS-VNL (including PIC with adaptive mesh refinement, a large-timestep 'drift-Lorentz' mover for arbitrarily magnetized species, a relativistic Lorentz invariant leapfrog particle pusher, simulations in Lorentz-boosted frames, an electromagnetic solver with tunable numerical dispersion and efficient stride-based digital filtering), with special emphasis on the description of the mesh refinement capability. In addition, selected examples of the applications of the methods to the abovementioned fields are given.

  10. Non-coding RNAs change their expression profile after Retinoid induced differentiation of the promyelocytic cell line NB4

    Directory of Open Access Journals (Sweden)

    Caporaso Maria G

    2010-01-01

    Full Text Available Abstract Background The importance of non-coding RNAs (ncRNAs as fine regulators of eukaryotic gene expression has emerged by several studies focusing on microRNAs (miRNAs. miRNAs represent a newly discovered family of non coding-RNAs. They are thought to be crucial players of human hematopoiesis and related tumorigenesis and to represent a potential tool to detect the early stages of cancer. More recently, the expression regulation of numerous long ncRNAs has been linked to cell growth, differentiation and cancer although the molecular mechanism of their function is still unknown. NB4 cells are promyelocytic cells that can be induced to differentiation upon retinoic acid (ATRA treatment and represent a feasible model to study changes of non coding RNAs expression between cancer cells and their terminally differentiated counterpart. Findings we screened, by microarray analysis, the expression of 243 miRNAs and 492 human genes transcribing for putative long ncRNAs different from miRNAs in NB4 cells before and after ATRA induced differentiation. Our data show that 8 miRNAs, and 58 long ncRNAs were deregulated by ATRA induced NB4 differentiation. Conclusion our data suggest that ATRA-induced differentiation lead to deregulation of a large number of the ncRNAs that can play regulatory roles in both tumorigenesis and differentiation.

  11. Calculation of the absorbed dose for contamination in skin imparted by beta radiation through the Varskin code modified for 122 isotopes of interest for nuclear medicine, nuclear plants and research

    International Nuclear Information System (INIS)

    In this work the implementation of a modification of the Varskin code for calculation of absorbed dose by contamination in skin imparted by external radiation fields generated by beta emitting is presented. The necessary data for the execution of the code are: isotope, dose depth, isotope activity, geometry type, source radio and time of integration of the isotope, being able to execute combinations of up to five radionuclides. This program it was implemented in Fortran 5 by means of the FFSKIN source program and the executable one in binary language BFFSKIN being the maximum execution time of 5 minutes. (Author)

  12. Weekly CODE chemotherapy with recombinant human granulocyte colony-stimulating factor for relapsed or refractory small cell lung cancer.

    Science.gov (United States)

    Sato, K; Tsuchiya, S; Minato, K; Sunaga, N; Ishihara, S I; Makimoto, T; Naruse, I; Hoshino, H; Watanabe, S; Saitoh, R; Mori, M

    2000-01-01

    We used cisplatin, vincristine, doxorubicin, and etoposide (CODE) plus recombinant human granulocyte colony-stimulating factor (rhG-CSF) weekly for salvage chemotherapy in relapsed or refractory small cell lung cancer (SCLC). We reviewed the medical charts of patients between January 1993 and December 1996 at the National Nishi-Gunma Hospital. Twenty patients were treated with salvage chemotherapy. The overall response rate was 55.0%. The median survival time of extensive disease patients from the start of CODE therapy was 23 weeks and the 1-year survival rate was 21.0%. Toxicities were severe, especially in myelosuppression. CODE could be selected as a salvage therapy for chemotherapy- relapsed SCLC cases.

  13. On the structure of Lattice code WIMSD-5B

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Young; Min, Byung Joo

    2004-03-15

    The WIMS-D code is a freely available thermal reactor physics lattice code used widely for thermal research and power reactor calculation. Now the code WIMS-AECL, developed on the basis of WIMS-D, has been used as one of lattice codes for the cell calculation in Canada and also, in 1998, the latest version WIMSD-5B is released for OECD/NEA Data Bank. While WIMS-KAERI was developed and has been used, originated from WIMS-D, in Korea, it was adjusted for the cell calculation of research reactor HANARO and so it has no confirmaty to CANDU reactor. Therefore, the code development applicable to cell calculation of CANDU reactor is necessary not only for technological independence and but also for the establishment of CANDU safety analysis system. A lattice code WIMSD-5B was analyzed in order to set the system of reactor physics computer codes, to be used in the assessment of void reactivity effect. In order to improve and validate WIMSD-5B code, the analysis of the structure of WIMSD-5B lattice code was made and so its structure, algorithm and the subroutines of WIMSD-5B were presented for the cluster type and the pij method modelling the CANDU-6 fuel

  14. On the structure of Lattice code WIMSD-5B

    International Nuclear Information System (INIS)

    The WIMS-D code is a freely available thermal reactor physics lattice code used widely for thermal research and power reactor calculation. Now the code WIMS-AECL, developed on the basis of WIMS-D, has been used as one of lattice codes for the cell calculation in Canada and also, in 1998, the latest version WIMSD-5B is released for OECD/NEA Data Bank. While WIMS-KAERI was developed and has been used, originated from WIMS-D, in Korea, it was adjusted for the cell calculation of research reactor HANARO and so it has no confirmaty to CANDU reactor. Therefore, the code development applicable to cell calculation of CANDU reactor is necessary not only for technological independence and but also for the establishment of CANDU safety analysis system. A lattice code WIMSD-5B was analyzed in order to set the system of reactor physics computer codes, to be used in the assessment of void reactivity effect. In order to improve and validate WIMSD-5B code, the analysis of the structure of WIMSD-5B lattice code was made and so its structure, algorithm and the subroutines of WIMSD-5B were presented for the cluster type and the pij method modelling the CANDU-6 fuel

  15. A fast parallel code for calculating energies and oscillator strengths of many-electron atoms at neutron star magnetic field strengths in adiabatic approximation

    Science.gov (United States)

    Engel, D.; Klews, M.; Wunner, G.

    2009-02-01

    We have developed a new method for the fast computation of wavelengths and oscillator strengths for medium-Z atoms and ions, up to iron, at neutron star magnetic field strengths. The method is a parallelized Hartree-Fock approach in adiabatic approximation based on finite-element and B-spline techniques. It turns out that typically 15-20 finite elements are sufficient to calculate energies to within a relative accuracy of 10-5 in 4 or 5 iteration steps using B-splines of 6th order, with parallelization speed-ups of 20 on a 26-processor machine. Results have been obtained for the energies of the ground states and excited levels and for the transition strengths of astrophysically relevant atoms and ions in the range Z=2…26 in different ionization stages. Catalogue identifier: AECC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AECC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 3845 No. of bytes in distributed program, including test data, etc.: 27 989 Distribution format: tar.gz Programming language: MPI/Fortran 95 and Python Computer: Cluster of 1-26 HP Compaq dc5750 Operating system: Fedora 7 Has the code been vectorised or parallelized?: Yes RAM: 1 GByte Classification: 2.1 External routines: MPI/GFortran, LAPACK, PyLab/Matplotlib Nature of problem: Calculations of synthetic spectra [1] of strongly magnetized neutron stars are bedevilled by the lack of data for atoms in intense magnetic fields. While the behaviour of hydrogen and helium has been investigated in detail (see, e.g., [2]), complete and reliable data for heavier elements, in particular iron, are still missing. Since neutron stars are formed by the collapse of the iron cores of massive stars, it may be assumed that their atmospheres contain an iron plasma. Our objective is to fill the gap

  16. Micromechanics Analysis Code With Generalized Method of Cells (MAC/GMC): User Guide. Version 3

    Science.gov (United States)

    Arnold, S. M.; Bednarcyk, B. A.; Wilt, T. E.; Trowbridge, D.

    1999-01-01

    The ability to accurately predict the thermomechanical deformation response of advanced composite materials continues to play an important role in the development of these strategic materials. Analytical models that predict the effective behavior of composites are used not only by engineers performing structural analysis of large-scale composite components but also by material scientists in developing new material systems. For an analytical model to fulfill these two distinct functions it must be based on a micromechanics approach which utilizes physically based deformation and life constitutive models and allows one to generate the average (macro) response of a composite material given the properties of the individual constituents and their geometric arrangement. Here the user guide for the recently developed, computationally efficient and comprehensive micromechanics analysis code, MAC, who's predictive capability rests entirely upon the fully analytical generalized method of cells, GMC, micromechanics model is described. MAC/ GMC is a versatile form of research software that "drives" the double or triply periodic micromechanics constitutive models based upon GMC. MAC/GMC enhances the basic capabilities of GMC by providing a modular framework wherein 1) various thermal, mechanical (stress or strain control) and thermomechanical load histories can be imposed, 2) different integration algorithms may be selected, 3) a variety of material constitutive models (both deformation and life) may be utilized and/or implemented, and 4) a variety of fiber architectures (both unidirectional, laminate and woven) may be easily accessed through their corresponding representative volume elements contained within the supplied library of RVEs or input directly by the user, and 5) graphical post processing of the macro and/or micro field quantities is made available.

  17. Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Kliem, S. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)]. E-mail: S.Kliem@fzd.de; Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Rohde, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Weiss, F.-P. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)

    2007-09-15

    Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.

  18. Formulation, Implementation and Validation of a Two-Fluid model in a Fuel Cell CFD Code

    Energy Technology Data Exchange (ETDEWEB)

    Kunal Jain, Vernon Cole, Sanjiv Kumar and N. Vaidya

    2008-11-01

    Water management is one of the main challenges in PEM Fuel Cells. While water is essential for membrane electrical conductivity, excess liquid water leads to ooding of catalyst layers. Despite the fact that accurate prediction of two-phase transport is key for optimal water management, understanding of the two-phase transport in fuel cells is relatively poor. Wang et. al. [1], [2] have studied the two-phase transport in the channel and diffusion layer separately using a multiphase mixture model. The model fails to accurately predict saturation values for high humidity inlet streams. Nguyen et. al. [3] developed a two-dimensional, two-phase, isothermal, isobaric, steady state model of the catalyst and gas diffusion layers. The model neglects any liquid in the channel. Djilali et. al. [4] developed a three-dimensional two-phase multicomponent model. The model is an improvement over previous models, but neglects drag between the liquid and the gas phases in the channel. In this work, we present a comprehensive two- fluid model relevant to fuel cells. Models for two-phase transport through Channel, Gas Diffusion Layer (GDL) and Channel-GDL interface, are discussed. In the channel, the gas and liquid pressures are assumed to be same. The surface tension effects in the channel are incorporated using the continuum surface force (CSF) model. The force at the surface is expressed as a volumetric body force and added as a source to the momentum equation. In the GDL, the gas and liquid are assumed to be at different pressures. The difference in the pressures (capillary pressure) is calculated using an empirical correlations. At the Channel-GDL interface, the wall adhesion affects need to be taken into account. SIMPLE-type methods recast the continuity equation into a pressure-correction equation, the solution of which then provides corrections for velocities and pressures. However, in the two-fluid model, the presence of two phasic continuity equations gives more freedom and

  19. Combining node-centered parallel radiation transport and higher-order multi-material cell-centered hydrodynamics methods in three-temperature radiation hydrodynamics code TRHD

    Science.gov (United States)

    Sijoy, C. D.; Chaturvedi, S.

    2016-06-01

    Higher-order cell-centered multi-material hydrodynamics (HD) and parallel node-centered radiation transport (RT) schemes are combined self-consistently in three-temperature (3T) radiation hydrodynamics (RHD) code TRHD (Sijoy and Chaturvedi, 2015) developed for the simulation of intense thermal radiation or high-power laser driven RHD. For RT, a node-centered gray model implemented in a popular RHD code MULTI2D (Ramis et al., 2009) is used. This scheme, in principle, can handle RT in both optically thick and thin materials. The RT module has been parallelized using message passing interface (MPI) for parallel computation. Presently, for multi-material HD, we have used a simple and robust closure model in which common strain rates to all materials in a mixed cell is assumed. The closure model has been further generalized to allow different temperatures for the electrons and ions. In addition to this, electron and radiation temperatures are assumed to be in non-equilibrium. Therefore, the thermal relaxation between the electrons and ions and the coupling between the radiation and matter energies are required to be computed self-consistently. This has been achieved by using a node-centered symmetric-semi-implicit (SSI) integration scheme. The electron thermal conduction is calculated using a cell-centered, monotonic, non-linear finite volume scheme (NLFV) suitable for unstructured meshes. In this paper, we have described the details of the 2D, 3T, non-equilibrium, multi-material RHD code developed with a special attention to the coupling of various cell-centered and node-centered formulations along with a suite of validation test problems to demonstrate the accuracy and performance of the algorithms. We also report the parallel performance of RT module. Finally, in order to demonstrate the full capability of the code implementation, we have presented the simulation of laser driven shock propagation in a layered thin foil. The simulation results are found to be in good

  20. Expression of I-A and I-E,C region-coded Ia antigens on functional B cell subpopulations.

    Science.gov (United States)

    Frelinger, J A; Hibbler, F J; Hill, S W

    1978-12-01

    Ia antigens from specific subregions have been examined on functional B cell populations. Expression of both I-A and I-E,C region antigens was demonstrated on cells required for both lipopolysaccharide mitogenesis and polyclonal activation. Similar I-A and I-E,C subregion expression was found on cells required for response to the T-independent antigen, polyvinylpyrrolidone. TNP-specific IgM and hen egg lysozyme-specific IgG plaque-forming cells also express I-A and I-E,C region antigens. No evidence was found for an Ia- population responsive in the systems tested. Further, no evidence of preferential expression of I-A or I-E,C region antigens was observed in any system examined. Therefore, it appears that B cells express both I-A and I-E,C region-coded Ia antigens.

  1. GermlncRNA: a unique catalogue of long non-coding RNAs and associated regulations in male germ cell development.

    Science.gov (United States)

    Luk, Alfred Chun-Shui; Gao, Huayan; Xiao, Sizhe; Liao, Jinyue; Wang, Daxi; Tu, Jiajie; Rennert, Owen M; Chan, Wai-Yee; Lee, Tin-Lap

    2015-01-01

    Spermatogenic failure is a major cause of male infertility, which affects millions of couples worldwide. Recent discovery of long non-coding RNAs (lncRNAs) as critical regulators in normal and disease development provides new clues for delineating the molecular regulation in male germ cell development. However, few functional lncRNAs have been characterized to date. A major limitation in studying lncRNA in male germ cell development is the absence of germ cell-specific lncRNA annotation. Current lncRNA annotations are assembled by transcriptome data from heterogeneous tissue sources; specific germ cell transcript information of various developmental stages is therefore under-represented, which may lead to biased prediction or fail to identity important germ cell-specific lncRNAs. GermlncRNA provides the first comprehensive web-based and open-access lncRNA catalogue for three key male germ cell stages, including type A spermatogonia, pachytene spermatocytes and round spermatids. This information has been developed by integrating male germ transcriptome resources derived from RNA-Seq, tiling microarray and GermSAGE. Characterizations on lncRNA-associated regulatory features, potential coding gene and microRNA targets are also provided. Search results from GermlncRNA can be exported to Galaxy for downstream analysis or downloaded locally. Taken together, GermlncRNA offers a new avenue to better understand the role of lncRNAs and associated targets during spermatogenesis. Database URL: http://germlncrna.cbiit.cuhk.edu.hk/ PMID:25982314

  2. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author)

  3. Post-test calculations of aerosol behavior in DEMONA experiment B3 with various computer codes used in CEC member states

    International Nuclear Information System (INIS)

    Extensive research has been carried out around the world to understand the behavior of radioactive materials in the containment building of an LWR under accident conditions. Most of this material is in the form of aerosols or is attached to non-radioactive aerosol particles in the containment atmosphere. Several computer codes have been written to describe fission product aerosol behavior under accident conditions, aimed at evaluating the time dependent airborne concentration inside the reactor building and the fraction that leaks out to the environment. The objective of this Study was to perform a comparison of computer codes used in the CEC member states with a DEMONA experiment. This is a follow up study to an earlier exercise comparing codes to each other in which rigid benchmark cases of more or less artificially detailed nature had been used. In the present Study the comparison to the DEMONA experiment was to be oriented only at the experimental results without additional help provided. It should thus provide a basis for judging the practical applicability of the codes to a situation which is real but, maybe, less well defined than a theoretical benchmark case

  4. Calculation of absorbed doses in sphere volumes around the Mammosite using the Monte Carlo simulation code MCNPX; Calculo de dosis absorbida en volumenes esfericos alrededor del Mammosite utilizando el codigo de simulacion Monte Carlo MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E. L. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2008-07-01

    The objective of this study is to investigate the changes observed in the absorbed doses in mammary gland tissue when irradiated with a equipment of high dose rate known as Mammosite and introducing material resources contrary to the tissue that constitutes the mammary gland. The modeling study is performed with the code MCNPX, 2005 version, the equipment and the mammary gland and calculating the absorbed doses in tissue when introduced small volumes of air or calcium in the system. (Author)

  5. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Energy Technology Data Exchange (ETDEWEB)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  6. MCNP code

    International Nuclear Information System (INIS)

    The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids

  7. Method of Characteristics Calculations and Computer Code for Materials with Arbitrary Equations of State and Using Orthogonal Polynomial Least Square Surface Fits

    Science.gov (United States)

    Chang, T. S.

    1974-01-01

    A numerical scheme using the method of characteristics to calculate the flow properties and pressures behind decaying shock waves for materials under hypervelocity impact is developed. Time-consuming double interpolation subroutines are replaced by a technique based on orthogonal polynomial least square surface fits. Typical calculated results are given and compared with the double interpolation results. The complete computer program is included.

  8. 中欧温室规范中风荷载取值的对比%Comparative study on calculation of wind loads on greenhouse structures between codes of China and Europe

    Institute of Scientific and Technical Information of China (English)

    童乐为; 金健; 周锋

    2013-01-01

    Among all the bearing loads for greenhouse structures, wind load plays a leading role and this is especially true for a modern greenhouse structure due to its large-span and lightweight characteristics. As a result, to ensure reliability and a reasonable cost for the design work of greenhouses, wind load becomes a key factor. Since the modern greenhouse started late in China, there was no domestic load specifications for greenhouse structures until 2002 when the national code“Greenhouse Structure Design Load”was published. This code filled in the blanks of this area in China, but its shortcomings are still obvious, as it cannot fully take into consideration all the particularities of a greenhouse structure. Therefore, it is necessary to take advantage of the experiences of advanced countries in greenhouse design and have some research on greenhouse design load codes. In combination with the Chinese Load Code and Eurocode1:Part1-4, this paper conducted a comparative study on the calculation of wind loads on greenhouse structures between the codes of China and Europe. In the first part of this paper, the definitions and calculation methods of wind load were compared briefly between Chinese and European Codes for Greenhouse Structures. Then, more specific discussions were carried out on three main factors:basic wind pressure, wind profile, and wind pressure coefficient. The comparison results showed that the calculation methods of wind load for a greenhouse structure are similar between Chinese and European Codes, but the definition and selection of some parameters are too simplified to be reasonable in the Chinese Codes. Specifically, the comparison study on basic wind pressure shows that the definition of this variable is almost the same in these two codes except that when under certain circumstance, the European greenhouse code takes into consideration its importance and the design working life of the greenhouse. As a result, the basic wind pressure return

  9. BRAF activated non-coding RNA (BANCR) promoting gastric cancer cells proliferation via regulation of NF-κB1

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Zhi-Xin; Liu, Zhi-Qiang; Jiang, Biao; Lu, Xin-Yang; Ning, Xiao-Fei [Department of Gastrointestinal Surgery, Affiliated Hospital of Jining Medical University, Jining 272029 (China); Yuan, Chuan-Tao [Department of Pathology, Affiliated Hospital of Jining Medical University, Jining 272029 (China); Wang, Ai-Liang, E-mail: wang_ailiang@126.com [Department of Gastrointestinal Surgery, Affiliated Hospital of Jining Medical University, Jining 272029 (China)

    2015-09-18

    Background and objective: Long non-coding RNA, BANCR, has been demonstrated to contribute to the proliferation and migration of tumors. However, its molecular mechanism underlying gastric cancer is still unknown. In present study, we investigated whether BANCR was involved in the development of gastric cancer cells via regulation of NF-κB1. Methods: Human gastric cancer tissues were isolated as well as human gastric cell lines MGC803 and BGC823 were cultured to investigate the role of BANCR in gastric cancer. Results: BANCR expression was significantly up-regulated in gastric tumor tissues and gastric cell lines. Down-regulation of BANCR inhibited gastric cancer cell growth and promoted cell apoptosis, and it also contributed to a significant decrease of NF-κB1 (P50/105) expression and 3′UTR of NF-κB1 activity. Overexpression of NF-κB1 reversed the effect of BANCR on cancer cell growth and apoptosis. MiroRNA-9 (miR-9) targeted NF-κB1, and miR-9 inhibitor also reversed the effects of BANCR on gastric cancer cell growth and apoptosis. Conclusion: BANCR was highly expressed both in gastric tumor tissues and in cancer cells. NF-κB1 and miR-9 were involved in the role of BANCR in gastric cancer cell growth and apoptosis. - Highlights: • BANCR up-regulated in gastric cancer (GC) tissues and cell lines MGC803 and BGC823. • Down-regulation of BANCR inhibited GC cell growth and promoted cell apoptosis. • Down-regulation of BANCR contributed to decreased 3′UTR of NF-κB1 and its expression. • Overexpressed NF-κB1 reversed the effect of BANCR on GC cell growth. • miR-9 inhibitor reversed the effect of BANCR on cancer GC cell growth.

  10. Cloned s-Lap Gene Coding Area, Expression and Localizationof s-Lap/GFP Fusion Protein in Mammal Cells

    Institute of Scientific and Technical Information of China (English)

    SONG Yi-shu; SONG Zhi-yu; LI Hong-jun; Wu Yin; BAO Yong-li; TAN Da-peng; LI Yu-xin

    2005-01-01

    s-Lap is a new gene sequence from pig retinal pigment epithelial(RPE) cells, which was found and cloned in the early period of apoptosis of RPE cells damaged with visible light. We cloned the coding area sequence of the novel gene of s-Lap and constructed its recombinant eukaryotic plasmid pcDNA3.1-GFP/s-lap with the recombinant DNA technique. The expression and localization of s-lap/GFP fusion protein in CHO and B16 cell lines were studied with the instantaneously transfected pcDNA3.1-GFP/s-lap recombinant plasmid. s-Lap/GFP fusion protein can be expressed in CHO and B16 cells with a high rate expression in the nuclei.

  11. SFACTOR: a computer code for calculating dose equivalent to a target organ per microcurie-day residence of a radionuclide in a source organ

    International Nuclear Information System (INIS)

    A computer code SFACTOR was developed to estimate the average dose equivalent S (rem/μCi-day) to each of a specified list of target organs per microcurie-day residence of a radionuclide in source organs in man. Source and target organs of interest are specified in the input data stream, along with the nuclear decay information. The SFACTOR code computes components of the dose equivalent rate from each type of decay present for a particular radionuclide, including alpha, electron, and gamma radiation. For those transuranic isotopes which also decay by spontaneous fission, components of S from the resulting fission fragments, neutrons, betas, and gammas are included in the tabulation. Tabulations of all components of S are provided for an array of 22 source organs and 24 target organs for 52 radionuclides in an adult

  12. HYDRA-I: a three-dimensional finite difference code for calculating the thermohydraulic performance of a fuel assembly contained within a canister

    Energy Technology Data Exchange (ETDEWEB)

    McCann, R.A.

    1980-12-01

    A finite difference computer code, named HYDRA-I, has been developed to simulate the three-dimensional performance of a spent fuel assembly contained within a cylindrical canister. The code accounts for the coupled heat transfer modes of conduction, convection, and radiation and permits spatially varying boundary conditions, thermophysical properties, and power generation rates. This document is intended as a manual for potential users of HYDRA-I. A brief discussion of the governing equations, the solution technique, and a detailed description of how to set up and execute a problem are presented. HYDRA-I is designed for operation on a CDC 7600 computer. An appendix is included that summarizes approximately two dozen different cases that have been examined. The cases encompass variations in fuel assembly and canister configurations, power generation rates, filler materials, and gases. The results presented show maximum and various local temperatures and heat fluxes illustrating the changing importance of the three heat transfer modes. Finally, the need for comparison with experimental data is emphasized as an aid in code verification although the limited data available indicate excellent agreement.

  13. Numerical Prediction of the Performance of Integrated Planar Solid-Oxide Fuel Cells, with Comparisons of Results from Several Codes

    Energy Technology Data Exchange (ETDEWEB)

    G. L. Hawkes; J. E. O' Brien; B. A. Haberman; A. J. Marquis; C. M. Baca; D. Tripepi; P. Costamagna

    2008-06-01

    A numerical study of the thermal and electrochemical performance of a single-tube Integrated Planar Solid Oxide Fuel Cell (IP-SOFC) has been performed. Results obtained from two finite-volume computational fluid dynamics (CFD) codes FLUENT and SOHAB and from a two-dimensional inhouse developed finite-volume GENOA model are presented and compared. Each tool uses physical and geometric models of differing complexity and comparisons are made to assess their relative merits. Several single-tube simulations were run using each code over a range of operating conditions. The results include polarization curves, distributions of local current density, composition and temperature. Comparisons of these results are discussed, along with their relationship to the respective imbedded phenomenological models for activation losses, fluid flow and mass transport in porous media. In general, agreement between the codes was within 15% for overall parameters such as operating voltage and maximum temperature. The CFD results clearly show the effects of internal structure on the distributions of gas flows and related quantities within the electrochemical cells.

  14. Continuous-Energy Adjoint Flux and Perturbation Calculation using the Iterated Fission Probability Method in Monte Carlo Code TRIPOLI-4® and Underlying Applications

    Science.gov (United States)

    Truchet, G.; Leconte, P.; Peneliau, Y.; Santamarina, A.; Malvagi, F.

    2014-06-01

    Pile-oscillation experiments are performed in the MINERVE reactor at the CEA Cadarache to improve nuclear data accuracy. In order to precisely calculate small reactivity variations (kinetic parameters (βeff, Λeff) or sensitivity parameters.

  15. Macrophage migration inhibition factor against cell-surface antigens coded by the major histocompatibility complex and other genes in mice.

    Directory of Open Access Journals (Sweden)

    Ohashi,Katsuhide

    1983-02-01

    Full Text Available We developed an indirect capillary tube method to improve reproducibility of macrophage migration inhibition (MI tests using a one-way mixed lymphocyte culture. MI response could be induced to cell-surface antigens coded by either H-2 or non-H-2 (background genes. The sensitivity was more readily induced across H-2 + background differences. The presence of only background difference did not induce the MI response to much extent. High MI activities were obtained to antigens coded by either K end or D end of the H-2 complex + background difference. Moderate activities were induced across the H-2D difference + background. These results suggest that the D region of the H-2 complex may direct a MI response when an H-2I difference is present during sensitization.

  16. Mutation rates of TGFBR2 and ACVR2 coding microsatellites in human cells with defective DNA mismatch repair.

    Directory of Open Access Journals (Sweden)

    Heekyung Chung

    Full Text Available Microsatellite instability promotes colonic tumorigenesis through generating frameshift mutations at coding microsatellites of tumor suppressor genes, such as TGFBR2 and ACVR2. As a consequence, signaling through these TGFbeta family receptors is abrogated in DNA Mismatch repair (MMR-deficient tumors. How these mutations occur in real time and mutational rates of these human coding sequences have not previously been studied. We utilized cell lines with different MMR deficiencies (hMLH1-/-, hMSH6-/-, hMSH3-/-, and MMR-proficient to determine mutation rates. Plasmids were constructed in which exon 3 of TGFBR2 and exon 10 of ACVR2 were cloned +1 bp out of frame, immediately after the translation initiation codon of an enhanced GFP (EGFP gene, allowing a -1 bp frameshift mutation to drive EGFP expression. Mutation-resistant plasmids were constructed by interrupting the coding microsatellite sequences, preventing frameshift mutation. Stable cell lines were established containing portions of TGFBR2 and ACVR2, and nonfluorescent cells were sorted, cultured for 7-35 days, and harvested for flow cytometric mutation detection and DNA sequencing at specific time points. DNA sequencing revealed a -1 bp frameshift mutation (A9 in TGFBR2 and A7 in ACVR2 in the fluorescent cells. Two distinct fluorescent populations, M1 (dim, representing heteroduplexes and M2 (bright, representing full mutants were identified, with the M2 fraction accumulating over time. hMLH1 deficiency revealed 11 (5.91 x 10(-4 and 15 (2.18 x 10(-4 times higher mutation rates for the TGFBR2 and ACVR2 microsatellites compared to hMSH6 deficiency, respectively. The mutation rate of the TGFBR2 microsatellite was approximately 3 times higher in both hMLH1 and hMSH6 deficiencies than the ACVR2 microsatellite. The -1 bp frameshift mutation rates of TGFBR2 and ACVR2 microsatellite sequences are dependent upon the human MMR background.

  17. HASEonGPU-An adaptive, load-balanced MPI/GPU-code for calculating the amplified spontaneous emission in high power laser media

    Science.gov (United States)

    Eckert, C. H. J.; Zenker, E.; Bussmann, M.; Albach, D.

    2016-10-01

    We present an adaptive Monte Carlo algorithm for computing the amplified spontaneous emission (ASE) flux in laser gain media pumped by pulsed lasers. With the design of high power lasers in mind, which require large size gain media, we have developed the open source code HASEonGPU that is capable of utilizing multiple graphic processing units (GPUs). With HASEonGPU, time to solution is reduced to minutes on a medium size GPU cluster of 64 NVIDIA Tesla K20m GPUs and excellent speedup is achieved when scaling to multiple GPUs. Comparison of simulation results to measurements of ASE in Y b 3 + : Y AG ceramics show perfect agreement.

  18. SCRIC: a code dedicated to the detailed emission and absorption of heterogeneous NLTE plasmas; application to xenon EUV sources; SCRIC: un code pour calculer l'absorption et l'emission detaillees de plasmas hors equilibre, inhomogenes et etendus; application aux sources EUV a base de xenon

    Energy Technology Data Exchange (ETDEWEB)

    Gaufridy de Dortan, F. de

    2006-07-01

    Nearly all spectral opacity codes for LTE and NLTE plasmas rely on configurations approximate modelling or even supra-configurations modelling for mid Z plasmas. But in some cases, configurations interaction (either relativistic and non relativistic) induces dramatic changes in spectral shapes. We propose here a new detailed emissivity code with configuration mixing to allow for a realistic description of complex mid Z plasmas. A collisional radiative calculation. based on HULLAC precise energies and cross sections. determines the populations. Detailed emissivities and opacities are then calculated and radiative transfer equation is resolved for wide inhomogeneous plasmas. This code is able to cope rapidly with very large amount of atomic data. It is therefore possible to use complex hydrodynamic files even on personal computers in a very limited time. We used this code for comparison with Xenon EUV sources within the framework of nano-lithography developments. It appears that configurations mixing strongly shifts satellite lines and must be included in the description of these sources to enhance their efficiency. (author)

  19. Application of TEMPPC code to the IEA-R1 nuclear reactor core hydrothermal calculations operating at 2 MW for determining the minimal coolant flow

    International Nuclear Information System (INIS)

    A thermohydraulic study of the IEA-R1 nuclear reactor core on steady-state operating condition and forced convection, is presented. The objective of this calculation is to obtain the minimal flow rate of coolant necessary at the reactor core, limited by the temperature associated to the beginning of nucleate boiling over the fuel plates at a normal operating power (2MW) for a certain inlet coolant temperature. The coolant system safety level is also calculated in this paper, which is divided in three steps: thermohydraulic calculation, without using the uncertainty factors and, after that, considering these factor by two methods: the statistical and the conventional ones. Whichever the method accepted, the results obtained by the program TEMPPC show a great safety margin with respect to the termohydraulic parameters from the IEA-R1 nuclear reactor. (Author)

  20. Dynamic transcription of long non-coding RNA genes during CD4+ T cell development and activation.

    Directory of Open Access Journals (Sweden)

    Fei Xia

    Full Text Available BACKGROUND: Recent evidence shows that long non-coding RNA (LncRNA play important regulatory roles in many biology process, including cell development, activation and oncogenesis. However, the roles of these LncRNAs in the development and activation of CD4+ T cells, which is an important component of immune response, remain unknown. RESULTS: To predict the function of LncRNA in the development and activation of CD4+ T cells, first, we examined the expression profiles of LncRNAs and mRNAs in CD4-CD8- (DN, CD4+CD8+ (DP, CD4+CD8-, and activated CD4+CD8- T cells in a microarray analysis and verified these results by real time PCRs (qPCR. We found that the expression of hundreds of LncRNAs significantly changed in each process of developmental transition, including DN into DP, DP into CD4+CD8-, and CD4+CD8- into activated CD4+ T cells. A Kendall distance analysis suggested that the expression of LncRNAs in DN, DP, CD4+CD8- T cells and activated CD4+ T cells were correlated with the expression of mRNAs in these T cells. The Blat algorithm and GO analysis suggested that LncRNAs may exert important roles in the development and activation of CD4+ T cells. These roles included proliferation, homeostasis, maturation, activation, migration, apoptosis and calcium ion transportation. CONCLUSION: The present study found that the expression profiles of LncRNAs in different stages of CD4+ T cells are distinguishable. LncRNAs are involved in the key biological process in CD4+ T cell development and activation.

  1. Coupling CFD code with system code and neutron kinetic code

    International Nuclear Information System (INIS)

    Highlights: • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case. - Abstract: The aim of this work was to develop the coupling interface between CFD code Fluent and system code Athlet internally coupled with neutron kinetic code Dyn3D. The coupling interface is intended for simulation of complex transients such as Main Steam Line Break scenarios, which cannot be modeled separately first by system and neutron kinetic code and then by CFD code, because of the feedback between the codes. In the first part of this article, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. The second part of the article presents a demonstration simulation performed by the coupled system of Athlet/Dyn3D and Fluent. The “Opening a Steam Dump to the Atmosphere” test carried out at the Temelin NPP (VVER-1000) was simulated by the coupled system. In this simulation, the primary and secondary circuits were modeled by Athlet, mixing in downcomer and lower plenum was simulated by Fluent and heat generation in the core was calculated by Dyn3D. The results of the simulation with Athlet/Dyn3D+Fluent were compared with the experimental data and the results from a calculation performed with Athlet/Dyn3D without Fluent

  2. Comparing Spray Characteristics from Reynolds Averaged Navier-Stokes (RANS) National Combustion Code (NCC) Calculations Against Experimental Data for a Turbulent Reacting Flow

    Science.gov (United States)

    Iannetti, Anthony C.; Moder, Jeffery P.

    2010-01-01

    Developing physics-based tools to aid in reducing harmful combustion emissions, like Nitrogen Oxides (NOx), Carbon Monoxide (CO), Unburnt Hydrocarbons (UHC s), and Sulfur Dioxides (SOx), is an important goal of aeronautics research at NASA. As part of that effort, NASA Glenn Research Center is performing a detailed assessment and validation of an in-house combustion CFD code known as the National Combustion Code (NCC) for turbulent reacting flows. To assess the current capabilities of NCC for simulating turbulent reacting flows with liquid jet fuel injection, a set of Single Swirler Lean Direct Injection (LDI) experiments performed at the University of Cincinnati was chosen as an initial validation data set. This Jet-A/air combustion experiment operates at a lean equivalence ratio of 0.75 at atmospheric pressure and has a 4 percent static pressure drop across the swirler. Detailed comparisons of NCC predictions for gas temperature and gaseous emissions (CO and NOx) against this experiment are considered in a previous work. The current paper is focused on detailed comparisons of the spray characteristics (radial profiles of drop size distribution and at several radial rakes) from NCC simulations against the experimental data. Comparisons against experimental data show that the use of the correlation for primary spray break-up implemented by Raju in the NCC produces most realistic results, but this result needs to be improved. Given the single or ten step chemical kinetics models, use of a spray size correlation gives similar, acceptable results

  3. Non-coding RNA regulation in pathogenic bacteria located inside eukaryotic cells

    NARCIS (Netherlands)

    Ortega, Alvaro D.; Quereda, Juan J; Pucciarelli, M Graciela; García-del Portillo, Francisco

    2014-01-01

    Intracellular bacterial pathogens have evolved distinct lifestyles inside eukaryotic cells. Some pathogens coexist with the infected cell in an obligate intracellular state, whereas others transit between the extracellular and intracellular environment. Adaptation to these intracellular lifestyles i

  4. The computer code SEURBNUK-2

    International Nuclear Information System (INIS)

    SEURBNUK-2 has been designed to model the hydrodynamic development in time of a hypothetical core disrupture accident in a fast breeder reactor. SEURBNUK-2 is a two-dimensional, axisymmetric, eulerian, finite difference containment code. The numerical procedure adopted in SEURBNUK to solve the hydrodynamic equations is based on the semi-implicit ICE method. SEURBNUK has a full thin shell treatment for tanks of arbitrary shape and includes the effects of the compressibility of the fluid. Fluid flow through porous media and porous structures can also be accommodated. An important feature of SEURBNUK is that the thin shell equations are solved quite separately from those of the fluid, and the time step for the fluid flow calculation can be an integer multiple of that for calculating the shell motion. The interaction of the shell with the fluid is then considered as a modification to the coefficients in the implicit pressure equations, the modifications naturally depending on the behaviour of the thin shell section within the fluid cell. The code is limited to dealing with a single fluid, the coolant, whereas the bubble and the cover gas are treated as cavities of uniform pressure calculated via appropriate pressure-volume-energy relationships. This manual describes the input data specifications needed for the execution of SEURBNUK-2 calculations and nine sample problems of varying degrees of complexity highlight the code capabilities. After explaining the output facilities information is included to aid those unfamiliar with SEURBNUK-2 to avoid the common pit-falls experienced by novices

  5. Long non-coding RNAs as surrogate indicators for chemical stress responses in human-induced pluripotent stem cells.

    Science.gov (United States)

    Tani, Hidenori; Onuma, Yasuko; Ito, Yuzuru; Torimura, Masaki

    2014-01-01

    In this study, we focused on two biological products as ideal tools for toxicological assessment: long non-coding RNAs (lncRNAs) and human-induced pluripotent stem cells (hiPSCs). lncRNAs are an important class of pervasive non-protein-coding transcripts involved in the molecular mechanisms associated with responses to cellular stresses. hiPSCs possess the capabilities of self-renewal and differentiation into multiple cell types, and they are free of the ethical issues associated with human embryonic stem cells. Here, we identified six novel lncRNAs (CDKN2B-AS1, MIR22HG, GABPB1-AS1, FLJ33630, LINC00152, and LINC0541471_v2) that respond to model chemical stresses (cycloheximide, hydrogen peroxide, cadmium, or arsenic) in hiPSCs. Our results indicated that the lncRNAs responded to general and specific chemical stresses. Compared with typical mRNAs such as p53-related mRNAs, the lncRNAs highly and rapidly responded to chemical stresses. We propose that these lncRNAs have the potential to be surrogate indicators of chemical stress responses in hiPSCs.

  6. Recent advances in the involvement of long non-coding RNAs in neural stem cell biology and brain pathophysiology

    Directory of Open Access Journals (Sweden)

    Daphne eAntoniou

    2014-04-01

    Full Text Available Exploration of non-coding genome has recently uncovered a growing list of formerly unknown regulatory long non-coding RNAs (lncRNAs with important functions in stem cell pluripotency, development and homeostasis of several tissues. Although thousands of lncRNAs are expressed in mammalian brain in a highly patterned manner, their roles in brain development have just begun to emerge. Recent data suggest key roles for these molecules in gene regulatory networks controlling neuronal and glial cell differentiation. Analysis of the genomic distribution of genes encoding for lncRNAs indicates a physical association of these regulatory RNAs with transcription factors (TFs with well-established roles in neural differentiation, suggesting that lncRNAs and TFs may form coherent regulatory networks with important functions in neural stem cells (NSCs. Additionally, many studies show that lncRNAs are involved in the pathophysiology of brain-related diseases/disorders. Here we discuss these observations and investigate the links between lncRNAs, brain development and brain-related diseases. Understanding the functions of lncRNAs in NSCs and brain organogenesis could revolutionize the basic principles of developmental biology and neuroscience.

  7. Long non-coding RNAs as surrogate indicators for chemical stress responses in human-induced pluripotent stem cells.

    Directory of Open Access Journals (Sweden)

    Hidenori Tani

    Full Text Available In this study, we focused on two biological products as ideal tools for toxicological assessment: long non-coding RNAs (lncRNAs and human-induced pluripotent stem cells (hiPSCs. lncRNAs are an important class of pervasive non-protein-coding transcripts involved in the molecular mechanisms associated with responses to cellular stresses. hiPSCs possess the capabilities of self-renewal and differentiation into multiple cell types, and they are free of the ethical issues associated with human embryonic stem cells. Here, we identified six novel lncRNAs (CDKN2B-AS1, MIR22HG, GABPB1-AS1, FLJ33630, LINC00152, and LINC0541471_v2 that respond to model chemical stresses (cycloheximide, hydrogen peroxide, cadmium, or arsenic in hiPSCs. Our results indicated that the lncRNAs responded to general and specific chemical stresses. Compared with typical mRNAs such as p53-related mRNAs, the lncRNAs highly and rapidly responded to chemical stresses. We propose that these lncRNAs have the potential to be surrogate indicators of chemical stress responses in hiPSCs.

  8. Development of a computational code for calculations of shielding in dental facilities; Desenvolvimento de um codigo computacional para calculos de blindagem em instalacoes odontologicas

    Energy Technology Data Exchange (ETDEWEB)

    Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L., E-mail: deise_dy@hotmail.com, E-mail: diogosb@outlook.com, E-mail: raoniwa@yahoo.com.br, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper is prepared in order to address calculations of shielding to minimize the interaction of patients with ionizing radiation and / or personnel. The work includes the use of protection report Radiation in Dental Medicine (NCRP-145 or Radiation Protection in Dentistry), which establishes calculations and standards to be adopted to ensure safety to those who may be exposed to ionizing radiation in dental facilities, according to the dose limits established by CNEN-NN-3.1 standard published in September / 2011. The methodology comprises the use of computer language for processing data provided by that report, and a commercial application used for creating residential projects and decoration. The FORTRAN language was adopted as a method for application to a real case. The result is a programming capable of returning data related to the thickness of material, such as steel, lead, wood, glass, plaster, acrylic, acrylic and leaded glass, which can be used for effective shielding against single or continuous pulse beams. Several variables are used to calculate the thickness of the shield, as: number of films used in the week, film load, use factor, occupational factor, distance between the wall and the source, transmission factor, workload, area definition, beam intensity, intraoral and panoramic exam. Before the application of the methodology is made a validation of results with examples provided by NCRP-145. The calculations redone from the examples provide answers consistent with the report.

  9. The effects of some parameters on the calculated 1H NMR relaxation times of cell water

    International Nuclear Information System (INIS)

    The effect of some parameters on the longitudinal and transverse relaxation times is calculated and a comparison between the calculated relaxation times with the results of different measurements is made. (M.S.)

  10. The non-coding RNA MALAT1 is a critical regulator of the metastasis phenotype of lung cancer cells

    OpenAIRE

    Gutschner, Tony; Hämmerle, Monika; Eißmann, Moritz; Hsu, Jeff; Kim, Youngsoo; Hung, Gene; Revenko, Alexey; Arun, Gayatri; Stentrup, Marion; Groß, Matthias; Zörnig, Martin; MacLeod, A. Robert; Spector, David L.; Diederichs, Sven

    2012-01-01

    The long non-coding RNA MALAT1, also known as MALAT-1 or NEAT2, is a highly conserved nuclear ncRNA and a predictive marker for metastasis development in lung cancer. To uncover its functional importance, we developed a MALAT1 knockout model in human lung tumor cells by genomically integrating RNA destabilizing elements using Zinc Finger Nucleases. The achieved 1000-fold MALAT1 silencing provides a unique loss-of-function model. Proposed mechanisms of action include regulation of splicing or ...

  11. Benchmark calculations of power distribution within fuel assemblies. Phase 2: comparison of data reduction and power reconstruction methods in production codes

    International Nuclear Information System (INIS)

    Systems loaded with plutonium in the form of mixed-oxide (MOX) fuel show somewhat different neutronic characteristics compared with those using conventional uranium fuels. In order to maintain adequate safety standards, it is essential to accurately predict the characteristics of MOX-fuelled systems and to further validate both the nuclear data and the computation methods used. A computation benchmark on power distribution within fuel assemblies to compare different techniques used in production codes for fine flux prediction in systems partially loaded with MOX fuel was carried out at an international level. It addressed first the numerical schemes for pin power reconstruction, then investigated the global performance including cross-section data reduction methods. This report provides the detailed results of this second phase of the benchmark. The analysis of the results revealed that basic data still need to be improved, primarily for higher plutonium isotopes and minor actinides. (author)

  12. Construction of a plasmid coding for green fluorescent protein tagged cathepsin L and data on expression in colorectal carcinoma cells

    Directory of Open Access Journals (Sweden)

    Tripti Tamhane

    2015-12-01

    Full Text Available The endo-lysosomal cysteine cathepsin L has recently been shown to have moonlighting activities in that its unexpected nuclear localization in colorectal carcinoma cells is involved in cell cycle progression (Tamhane et al., 2015 [1]. Here, we show data on the construction and sequence of a plasmid coding for human cathepsin L tagged with an enhanced green fluorescent protein (phCL-EGFP in which the fluorescent protein is covalently attached to the C-terminus of the protease. The plasmid was used for transfection of HCT116 colorectal carcinoma cells, while data from non-transfected and pEGFP-N1-transfected cells is also shown. Immunoblotting data of lysates from non-transfected controls and HCT116 cells transfected with pEGFP-N1 and phCL-EGFP, showed stable expression of cathepsin L-enhanced green fluorescent protein chimeras, while endogenous cathepsin L protein amounts exceed those of hCL-EGFP chimeras. An effect of phCL-EGFP expression on proliferation and metabolic states of HCT116 cells at 24 h post-transfection was observed.

  13. Accurate measurement of sample conductivity in a diamond anvil cell with axis symmetrical electrodes and finite difference calculation

    Directory of Open Access Journals (Sweden)

    Jie Yang

    2011-09-01

    Full Text Available We report a relatively precise method of conductivity measurement in a diamond anvil cell with axis symmetrical electrodes and finite difference calculation. The axis symmetrical electrodes are composed of two parts: one is a round thin-film electrode deposited on diamond facet and the other is the inside wall of metal gasket. Due to the asymmetrical configuration of the two electrodes, finite difference method can be applied to calculate the conductivity of sample, which can reduce the measurement error.

  14. A comparison of the FA's models with the detailed and simplified description of the design elements in calculations by MCU code

    International Nuclear Information System (INIS)

    The full-scale three-dimensional computer model describing the physical processes that occur in the reactor core of the WWER-1200 assumes the detailed description of the geometry of the fuel assembly. However, detailed description of the geometry essentially complicates the tasking of the entrance data, increases the time of calculation and probability of errors. As a consequence, creating the simplified models of the design elements becomes necessary. In the present work the simplified geometrical models of fuel assemblies design elements are offered. Comparative calculations of the simplified models and the exact models completely corresponding to the working drawings of a general view OKD 'Hydropress' are carried out. It is shown that in many problems using of simplified description of the fuel assembly elements geometry is possible. (Authors)

  15. Dose calculations for a simplified Mammosite system with the Monte Carlo Penelope and MCNPX simulation codes; Calculos de dosis para un sistema Mammosite simplificado con los codigos de simulacion Monte Carlo PENELOPE y MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L.; Varon T, C.F.; Pedraza N, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: elrc@nuclear.inin.mx

    2007-07-01

    The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)

  16. Parallelizing Particle-In-Cell Codes with OpenMP and MPI

    OpenAIRE

    Larsgård, Nils Magnus

    2007-01-01

    Today's supercomputers often consists of clusters of SMP nodes. Both OpenMP and MPI are programming paradigms that can be used for parallelization of codes for such architectures. OpenMP uses shared memory, and hence is viewed as a simpler programming paradigm than MPI that is primarily a distributed memory paradigm. However, the Open MP applications may not scale beyond one SMP node. On the other hand, if we only use MPI, we might introduce overhead in intra-node communication. In this the...

  17. Large scale tracking of stem cells using sparse coding and coupled graphs

    DEFF Research Database (Denmark)

    Vestergaard, Jacob Schack; Dahl, Anders Lindbjerg; Holm, Peter;

    Stem cell tracking is an inherently large scale problem. The challenge is to identify and track hundreds or thousands of cells over a time period of several weeks. This requires robust methods that can leverage the knowledge of specialists on the field. The tracking pipeline presented here consists...... of a dictionary learning method for segmentation of phase contrast microscopy images. Linking of the cells between two images is solved by a graph formulation of the tracking problem....

  18. Tritons at energies of 10 MeV to 1 TeV: Conversion coefficients for fluence-to-absorbed dose, equivalent dose, effective dose, and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.C

    International Nuclear Information System (INIS)

    Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult female and an adult male to tritons (3H+) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). Coefficients were calculated using Monte Carlo transport code MCNPX 2.7.C and BodyBuilderTM 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and calculation of gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 3%. The greatest difference, 43%, occurred at 30 MeV. Published by Oxford Univ. Press on behalf of the US Government 2010. (authors)

  19. Long Non-coding RNAs Expression Profile in HepG2 Cells Reveals the Potential Role of Long Non-coding RNAs in the Cholesterol Metabolism

    Institute of Scientific and Technical Information of China (English)

    Gang Liu; Xinxin Zheng; Yanlu Xu; Jie Lu; Jingzhou Chen; Xiaohong Huang

    2015-01-01

    Background:Green tea has been shown to improve cholesterol metabolism in animal studies,but the molecular mechanisms underlying this function have not been fully understood.Long non-coding RNAs (lncRNAs) have recently emerged as a major class of regulatory molecules involved in a broad range of biological processes and complex diseases.Our aim was to identify important lncRNAs that might play an important role in contributing to the benefits of epigallocatechin-3-gallate (EGCG) on cholesterol metabolism.Methods:Microarrays was used to reveal the lncRNA and mRNA profiles in green tea polyphenol(-)-epigallocatechin gallate in cultured human liver (HepG2) hepatocytes treated with EGCG and bioinformatic analyses of the predicted target genes were performed to identify lncRNA-mRNA targeting relationships.RNA interference was used to investigate the role of lncRNAs in cholesterol metabolism.Results:The expression levels of 15 genes related to cholesterol metabolism and 285 lncRNAs were changed by EGCG treatment.Bioinformatic analysis found five matched lncRNA-mRNA pairs for five differentially expressed lncRNAs and four differentially expressed mRNA.In particular,the lncRNA4 T102202 and its potential targets mRNA-3-hydroxy-3-methylglutaryl coenzyme A reductase (HMGCR) were identified.Using a real-time polymerase chain reaction technique,we confirmed that EGCG down-regulated mRNA expression level of the HMGCR and up-regulated expression ofAT102202.After AT102202 knockdown in HepG2,we observed that the level of HMGCR expression was significantly increased relative to the scrambled small interfering RNA control (P < 0.05).Conclusions:Our results indicated that EGCG improved cholesterol metabolism and meanwhile changed the lncRNAs expression profile in HepG2 cells.LncRNAs may play an important role in the cholesterol metabolism.

  20. Long non-coding RNAs: An emerging powerhouse in the battle between life and death of tumor cells.

    Science.gov (United States)

    Xiong, Xing-Dong; Ren, Xingcong; Cai, Meng-Yun; Yang, Jay W; Liu, Xinguang; Yang, Jin-Ming

    2016-05-01

    Long non-coding RNAs (lncRNAs) represent a class of non-protein coding transcripts longer than 200 nucleotides that have aptitude for regulating gene expression at the transcriptional, post-transcriptional or epigenetic levels. In recent years, lncRNAs, which are believed to be the largest transcript class in the transcriptomes, have emerged as important players in a variety of biological processes. Notably, the identification and characterization of numerous lncRNAs in the past decade has revealed a role for these molecules in the regulation of cancer cell survival and death. It is likely that this class of non-coding RNA constitutes a critical contributor to the assorted known or/and unknown mechanisms of intrinsic or acquired drug resistance. Moreover, the expression of lncRNAs is altered in various patho-physiological conditions, including cancer. Therefore, lncRNAs represent potentially important targets in predicting or altering the sensitivity or resistance of cancer cells to various therapies. Here, we provide an overview on the molecular functions of lncRNAs, and discuss their impact and importance in cancer development, progression, and therapeutic outcome. We also provide a perspective on how lncRNAs may alter the efficacy of cancer therapy and the promise of lncRNAs as novel therapeutic targets for overcoming chemoresistance. A better understanding of the functional roles of lncRNA in cancer can ultimately translate to the development of novel, lncRNA-based intervention strategies for the treatment or prevention of drug-resistant cancer. PMID:27180308

  1. Ab initio calculations of X-ray magnetic circular dichroism spectra within the projector augmented wave method: An implementation into the VASP code

    Science.gov (United States)

    Dixit, Anant; Alouani, M.

    2016-10-01

    X-ray absorption and X-ray magnetic circular dichroism (XMCD) are very powerful tools for probing the orbital and spin moments of each atomic species orbital of magnetic materials. In this work, we present the implementation of a module for computing the X-ray absorption and XMCD spectra into the VASP code. We provide a derivation of the absorption cross-section in the electric dipole approximation. The matrix elements, which make up the X-ray absorption cross-section for a given polarization of light, are then computed using either the momentum operator p or the position operator r, within the projector augmented wave method. The core electrons are described using the relativistic basis-set whereas for the valence electrons, the spin-orbit coupling is added perturbatively to the semi-relativistic Hamiltonian. We show that both the p and the r implementations lead to the same results. The results for the K-edge and L23-edges of bcc-iron are then computed and compared to experiment.

  2. Calculation of the thermal disadvantage factor for a reactor cell with anisotropic scattering by the Fn method

    International Nuclear Information System (INIS)

    The F sub(N) method is used for the calculation of the thermal disadvantage factor in reactor cells with anisotropic scattering in the moderator. Numerical results were obtained for several reactor cells and compared with the results obtained by other methods. The results confirmed the physical conclusion, that the higher order terms in the expansion of the scattering law have an insignificant effect on the thermal disadvantage factor. (E.G.)

  3. Time-dependent distribution functions and resulting synthetic NPA spectra in C-Mod calculated with the CQL3D-Hybrid-FOW, AORSA full-wave, and DC Lorentz codes

    Science.gov (United States)

    Harvey, R. W.; Petrov, Yu.; Jaeger, E. F.; Berry, L. A.; Bonoli, P. T.; Bader, A.

    2015-12-01

    A time-dependent simulation of C-Mod pulsed TCRF power is made obtaining minority hydrogen ion distributions with the CQL3D-Hybrid-FOW finite-orbit-width Fokker-Planck code. Cyclotron-resonant TCRF fields are calculated with the AORSA full wave code. The RF diffusion coefficients used in CQL3D are obtained with the DC Lorentz gyro-orbit code for perturbed particle trajectories in the combined equilibrium and TCRF electromagnetic fields. Prior results with a zero-banana-width simulation using the CQL3D/AORSA/DC time-cycles showed a pronounced enhancement of the H distribution in the perpendicular velocity direction compared to results obtained from Stix's quasilinear theory, and this substantially increased the rampup rate of the observed vertically-viewed neutral particle analyzer (NPA) flux, in general agreement with experiment. However, ramp down of the NPA flux after the pulse, remained long compared to the experiment. The present study compares the new FOW results, including relevant gyro-radius effects, to determine the importance of these new effects on the the NPA time-dependence.

  4. Development of a computer code for shielding calculation in X-ray facilities; Desenvolvimento de um codigo computacional para calculos de blindagem em salas radiograficas

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F., E-mail: diogosb@outlook.com, E-mail: deise_dy@hotmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: malu@ien.gov.br, E-mail: tony@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    The construction of an effective barrier against the interaction of ionizing radiation present in X-ray rooms requires consideration of many variables. The methodology used for specifying the thickness of primary and secondary shielding of an traditional X-ray room considers the following factors: factor of use, occupational factor, distance between the source and the wall, workload, Kerma in the air and distance between the patient and the receptor. With these data it was possible the development of a computer program in order to identify and use variables in functions obtained through graphics regressions offered by NCRP Report-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) for the calculation of shielding of the room walls as well as the wall of the darkroom and adjacent areas. With the built methodology, a program validation is done through comparing results with a base case provided by that report. The thickness of the obtained values comprise various materials such as steel, wood and concrete. After validation is made an application in a real case of radiographic room. His visual construction is done with the help of software used in modeling of indoor and outdoor. The construction of barriers for calculating program resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in September / 2011.

  5. Adaptation to background light enables contrast coding at rod bipolar cell synapses

    OpenAIRE

    Ke, Jiang-Bin; Wang, Yanbin V.; Borghuis, Bart G.; Cembrowski, Mark S.; Riecke, Hermann; Kath, William L.; Demb, Jonathan B; Joshua H Singer

    2013-01-01

    Rod photoreceptors contribute to vision over a ~6 log-unit range of light intensities. The wide dynamic range of rod vision is thought to depend upon light intensity-dependent switching between two parallel pathways linking rods to ganglion cells: a rod→rod bipolar (RB) cell pathway that operates at dim backgrounds and a rod→cone→cone bipolar cell pathway that operates at brighter backgrounds. We evaluated this conventional model of rod vision by recording rod-mediated light responses from ga...

  6. Numerical Calculations of Single-Cell Electroporation with an Electrolyte-Filled Capillary

    OpenAIRE

    Zudans, Imants; Agarwal, Aparna; Orwar, Owe; Weber, Stephen G.

    2007-01-01

    An electric field is focused on one cell in single-cell electroporation. This enables selective electroporation treatment of the targeted cell without affecting its neighbors. While factors that lead to membrane permeation are the same as in bulk electroporation, quantitative description of the single-cell experiments is more complicated. This is due to the fact that the potential distribution cannot be solved analytically. We present single-cell electroporation with an electrolyte-filled cap...

  7. Deuterons at energies of 10 MeV to 1 TeV: Conversion coefficients for fluence-to-absorbed dose, equivalent dose, effective dose and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.C

    International Nuclear Information System (INIS)

    Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult female and an adult male to deuterons (2H+) in the energy range 10 MeV-1 TeV (0.01-1000 GeV). Coefficients were calculated using the Monte Carlo transport code MCNPX 2.7.C and BodyBuilderTM 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of the effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. Coefficients for the equivalent and effective dose incorporated a radiation weighting factor of 2. At 15 of 19 energies for which coefficients for the effective dose were calculated, coefficients based on ICRP 1990 and 2007 recommendations differed by < 3 %. The greatest difference, 47 %, occurred at 30 MeV. (authors)

  8. Helions at energies of 10 MeV to 1 TeV: Conversion coefficients for fluence-to-absorbed dose, equivalent dose, effective dose and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.C

    International Nuclear Information System (INIS)

    Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent, for isotropic exposure of an adult male and an adult female to helions (3He2+) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). Calculations were performed using Monte Carlo transport code MCNPX 2.7.C and BodyBuilderTM 1.3 anthropomorphic phantoms modified to allow calculation of effective dose using tissues and tissue weighting factors from either the 1990 or 2007 recommendations of the International Commission on Radiological Protection (ICRP), and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 2%. The greatest difference, 62%, occurred at 100 MeV. Published by Oxford Univ. Press on behalf of the U.S. Government 2010. (authors)

  9. Polypeptide composition and gag gene-coded products of type-D oncovirus from HEp-2 cells.

    Science.gov (United States)

    Morozov, V A

    1982-01-01

    The protein composition of type-D oncovirus HEp-2, isolated from cell-free medium of continuous human HEp-2 cell line, has been investigated using electrophoresis on gradient polyacrylamide gels with sodium dodecyl sulfate (SDS). Labeling with 14C-amino acids revealed five viral polypeptides with molecular weights of 70 000 (gp70), 27 000 (p27), 19 000 (p19), 15 000 (p15), 12 000-10 000 (p12-10). The 70 000 dalton protein was shown to be the only glycoprotein by incorporation of radioactive glucosamine. A polypeptide with molecular weight of 78 000 has been specifically precipitated from pulse-labeled type-D oncovirus producing HEp-2 cells with goat anti Mason-Pfizer p27 serum. This protein was shown to be gag gene-coded polyprotein precursor (Pr78gag) of the major virus polypeptide p27. Pulse-labeled HEp-2 and Mason-Pfizer infected Tu 197 cells were rinsed, lysed, clarified and precipitated with goat anti Mason-Pfizer p27 serum. In both cases Pr78gag was detected.

  10. Circadian plasticity in photoreceptor cells controls visual coding efficiency in Drosophila melanogaster.

    Directory of Open Access Journals (Sweden)

    Martin Barth

    Full Text Available In the fly Drosophila melanogaster, neuronal plasticity of synaptic terminals in the first optic neuropil, or lamina, depends on early visual experience within a critical period after eclosion. The current study revealed two additional and parallel mechanisms involved in this type of synaptic terminal plasticity. First, an endogenous circadian rhythm causes daily oscillations in the volume of photoreceptor cell terminals. Second, daily visual experience precisely modulates the circadian time course and amplitude of the volume oscillations that the photoreceptor-cell terminals undergo. Both mechanisms are separable in their molecular basis. We suggest that the described neuronal plasticity in Drosophila ensures continuous optimal performance of the visual system over the course of a 24 h-day. Moreover, the sensory system of Drosophila cannot only account for predictable, but also for acute, environmental changes. The volumetric changes in the synaptic terminals of photoreceptor cells are accompanied by circadian and light-induced changes of presynaptic ribbons as well as extensions of epithelial glial cells into the photoreceptor terminals, suggesting that the architecture of the lamina is altered by both visual exposure and the circadian clock. Clock-mutant analysis and the rescue of PER protein rhythmicity exclusively in all R1-6 cells revealed that photoreceptor-cell plasticity is autonomous and sufficient to control visual behavior. The strength of a visually guided behavior, the optomotor turning response, co-varies with synaptic-terminal volume oscillations of photoreceptor cells when elicited at low light levels. Our results show that behaviorally relevant adaptive processing of visual information is performed, in part, at the level of visual input level.

  11. Dose Calculations for Lung Inhomogeneity in High-Energy Photon Beams and Small Beamlets: A Comparison between XiO and TiGRT Treatment Planning Systems and MCNPX Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    Asghar Mesbahi

    2015-09-01

    Full Text Available Introduction Radiotherapy with small fields is used widely in newly developed techniques. Additionally, dose calculation accuracy of treatment planning systems in small fields plays a crucial role in treatment outcome. In the present study, dose calculation accuracy of two commercial treatment planning systems was evaluated against Monte Carlo method. Materials and Methods Siemens Once or linear accelerator was simulated, using MCNPX Monte Carlo code, according to manufacturer’s instructions. Three analytical algorithms for dose calculation including full scatter convolution (FSC in TiGRT, along with convolution and superposition in XiO system were evaluated for a small solid liver tumor. This solid tumor with a diameter of 1.8 cm was evaluated in a thorax phantom, and calculations were performed for different field sizes (1×1, 2×2, 3×3 and4×4 cm2. The results obtained in these treatment planning systems were compared with calculations by MC method (regarded as the most reliable method. Results For FSC and convolution algorithm, comparison with MC calculations indicated dose overestimations of up to 120%and 25% inside the lung and tumor, respectively in 1×1 cm2field size, using an 18 MV photon beam. Regarding superposition, a close agreement was seen with MC simulation in all studied field sizes. Conclusion The obtained results showed that FSC and convolution algorithm significantly overestimated doses of the lung and solid tumor; therefore, significant errors could arise in treatment plans of lung region, thus affecting the treatment outcomes. Therefore, use of MC-based methods and super position is recommended for lung treatments, using small fields and beamlets.

  12. SP-FISPACT2001. A computer code for activation and decay calculations for intermediate energies. A connection of FISPACT with MCNPX; SP-FISPACT2001. Una connessione di FISPACT con MCNPX per la codifica computerizzata delle energie intermedie

    Energy Technology Data Exchange (ETDEWEB)

    Petrovich, C. [ENEA, Divisione Sistemi Energetici Ecosostenibili, Centro Ricerche Ezio Clementel, Bologna (Italy)

    2001-07-01

    The calculation of the number of atoms and the activity of materials following nuclear interactions at incident energies up to several GeV is necessary in the design of Accelerator Driven Systems, Radioactive Ion Beam and proton accelerator facilities such as spallation neutron sources. As well as the radioactivity of the materials, this allows the evaluation of the formation of active gaseous elements and the assessment of possible corrosion problems The particle energies involved here are higher than those used in typical nuclear reactors and fusion devices for which many codes already exist. These calculations can be performed by coupling two different computer codes: MCNPX and SP-FISPACT. MCNPX performs Monte Carlo particle transport up to energies of several GeV. SP-FISPACT is a modification of FISPACT, a code designed for fusion applications and able to calculate neutron activation for energies <20 MeV. In such a way it is possible to perform a hybrid calculation in which neutron activation data are used for neutron interactions at energies <20 MeV and intermediate energy physics models for all the other nuclear interactions. [Italian] In fase di design di sistemi ADS (Accelerator Driven Systems), di strutture con acceleratori quali quelli finalizzate alla produzione di fasci di ioni radioattivi o a sorgenti neutroniche di spallazione e' necessario calcolare la composizione e l'attivita' di materiali a seguito di interazioni nucleari con energie fino a qualche GeV. Oltre la radioattivita' dei materiali, questi calcoli permettono di prevedere la formazione di elementi gassosi attivi e possibili problemi di corrosione. Le energie delle particelle qui coinvolte sono piu' alte di quelle usate in tipici reattori nucleari ed in dispositivi finalizzati alla fusione, per i quali sono gia' disponibili diversi codici. Questi tipi di calcoli possono essere eseguiti accoppiando due codici differenti: MCNPX e SP-FISPACT. MCNPX trasporta

  13. Approach to the calculation of energy deposition in a container of fuel irradiated by the neutronic codes coupling fluid-dynamics; Aprpoximacion al calculo de la deposicion energetica en un contenedor de combustible irradiado mediante el acoplamiento de codigos neutronico fluido-dinamicos

    Energy Technology Data Exchange (ETDEWEB)

    Hueso, C.; Aleman, A.; Colomer, C.; Fabbri, M.; Martin, M.; Saellas, J.

    2013-07-01

    In this work identifies a possible area of improvement through the creation of a code of coupling between deposition energy codes which calculate neutron (MCNP), and data from heading into fluid dynamics (ANSYS-Fluent) or codes thermomechanical, called MAFACS (Monte Carlo ANSYS Fluent Automatic Coupling Software), being possible to so summarize the process by shortening the needs of computing time, increasing the precision of the results and therefore improving the design of the components.

  14. Latest development in codes for electromagnetic fields

    International Nuclear Information System (INIS)

    The latest development in codes for calculating electromagnetic field has concentrated on fully three-dimensional programs. Two-dimensional codes are well established and in routine use for designing accelerator equipment. However, some 2D problems are not yet solved and are still being worked on, such as determining the impedance of cylindrical objects above the cut-off frequency of the adjacent beam pipe. Three-dimensional codes are still being developed but only a few are being used. The most advanced package, the MAFIA system, has been extended recently to allow solution of electrostatic and magnetostatic fields with quasi-open boundary conditions. The modules solving for resonant frequencies and modes have been improved in their accuracy and speed. A new major revision of the whole family of codes will provide a consistent and more comfortable user interface, which is fully menu-driven with built in on-line help for all commands. This new release will also include new modules for 3D particle-in-cell simulation. Many new modules are under development and will make this system a universal tool in the design of electromagnetic devices. A user guide to the MAFIA codes with easy-to-follow instructions and examples is now available to noncommercial institutions, as is also the source code. Concurrent codes, which are not generally available, such as ARGUS or SOS, have also been improved. One can observe that the different codes become more and more alike with time. 43 refs., 6 figs., 1 tab

  15. Calculation of the Performance of Solar Cells With Spectral Down Shifters Using Realistic Outdoor Solar Spectra

    NARCIS (Netherlands)

    van Sark, W.G.J.H.M.

    2007-01-01

    Spectral down converters and shifters have been proposed as a good means to enhance the efficiency of underlying solar cells. In this paper, we focus on the simulation of the outdoor performance of solar cells with spectral down shifters, i.e., multicrystalline silicon solar cells with semiconductor

  16. Development of 2D particle-in-cell code to simulate high current, low energy beam in a beam transport system

    Indian Academy of Sciences (India)

    S C L Srivastava; S V L S Rao; P Singh

    2007-10-01

    A code for 2D space-charge dominated beam dynamics study in beam transport lines is developed. The code is used for particle-in-cell (PIC) simulation of -uniform beam in a channel containing solenoids and drift space. It can also simulate a transport line where quadrupoles are used for focusing the beam. Numerical techniques as well as the results of beam dynamics studies are presented in the paper.

  17. Tokamak Systems Code

    International Nuclear Information System (INIS)

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  18. INDAR: a computer code for the calculation of critical group radiation exposure from routine discharges of radioactivity to seas and estuaries - description and users' guide

    International Nuclear Information System (INIS)

    The computer program INDAR enables detailed estimates to be made of critical group radiation exposure arising from routine discharges of radioactivity for coastal sites where the discharge is close to the shore and the shoreline is reasonably straight, and for estuarine sites where radioactivity is rapidly mixed across the width of the estuary. Important processes which can be taken into account include the turbulence generated by the discharge, the effects of a sloping sea bed and the variation with time of the lateral dispersion coefficient. The significance of the timing of discharges can also be assessed. INDAR uses physically meaningful hydrographic parameters directly. For most sites the most important exposure pathways are seafood consumption, external exposure over estuarine sediments and beaches, and the handling of fishing gear. As well as for these primary pathways, INDAR enables direct calculations to be made for some additional exposure pathways. The secondary pathways considered are seaweed consumption, swimming, the handling of materials other than fishing gear and the inhalation of activity. (author)

  19. SHAPEMOL: a 3-D code to calculate CO line emission in planetary and protoplanetary nebulae. Detailed model fitting of the complex nebula NGC 6302

    CERN Document Server

    Santander-Garcia, M; Koning, N; Steffen, W

    2014-01-01

    Modern instrumentation in radioastronomy constitutes a valuable tool for studying the Universe: ALMA has reached unprecedented sensitivities and spatial resolution, while Herschel/HIFI has opened a new window for probing molecular warm gas (~50-1000 K). On the other hand, the SHAPE software has emerged in the last few years as a standard tool for determining the morphology and velocity field of different kinds of gaseous emission nebulae via spatio-kinematical modelling. SHAPE implements radiative transfer solving, but it is only available for atomic species and not for molecules. Being aware of the growing importance of the development of tools for simplifying the analyses of molecular data, we introduce shapemol, a complement to SHAPE with which we intend to fill the so far under-developed molecular niche. shapemol enables user-friendly, spatio-kinematic modeling with accurate non-LTE calculations of excitation and radiative transfer in CO lines. It allows radiative transfer solving in the 12CO and 13CO J=1...

  20. Comparison in the design and thermodynamic calculation of BNPP steam plant with the previous design (PWR-1300) and the new design (VVER-1000) using CXEMAcomputer code

    International Nuclear Information System (INIS)

    In a general view, the steam cycle of the Bushehr nuclear power plant is similar to that of thermal power plants using fossil fuel. However, dissimilarities appear in design and fabrication of the turbin unit components, due to the necessity of using the saturated steam through the cycle. At the same time and considering the modifications implemented in the new design of the Bushehr nuclear power plant (to say, adoption of the Russia VVER-1000 instead of the previous german PWR-1300 plants), the turbine unit has also undergone some modification. The major changes include the employment of 3 low pressure turbine cylinders instead of the ex-2 ones of the german design and RPM of 3000 as the turbine rotation rate instead of the 1500 one for the german design. In this paper in addition to the comparison of the two designs, the turbine parameters of both designs are also calculated with the CXEMAcode and the achieved results are brought into comparison and analysis. This paper has been prepared on the basis of the results represented in the author's (D.Golmoradi) Master of Sciences thesis