Performance evaluation for compressible flow calculations on five parallel computers of different architectures

International Nuclear Information System (INIS)

Kimura, Toshiya.

1997-03-01

A two-dimensional explicit Euler solver has been implemented for five MIMD parallel computers of different machine architectures in Center for Promotion of Computational Science and Engineering of Japan Atomic Energy Research Institute. These parallel computers are Fujitsu VPP300, NEC SX-4, CRAY T94, IBM SP2, and Hitachi SR2201. The code was parallelized by several parallelization methods, and a typical compressible flow problem has been calculated for different grid sizes changing the number of processors. Their effective performances for parallel calculations, such as calculation speed, speed-up ratio and parallel efficiency, have been investigated and evaluated. The communication time among processors has been also measured and evaluated. As a result, the differences on the performance and the characteristics between vector-parallel and scalar-parallel computers can be pointed, and it will present the basic data for efficient use of parallel computers and for large scale CFD simulations on parallel computers. (author)

Computer calculation of dose distributions in radiotherapy. Report of a panel

International Nuclear Information System (INIS)

1966-01-01

As in most areas of scientific endeavour, the advent of electronic computers has made a significant impact on the investigation of the physical aspects of radiotherapy. Since the first paper on the subject was published in 1955 the literature has rapidly expanded to include the application of computer techniques to problems of external beam, and intracavitary and interstitial dosimetry. By removing the tedium of lengthy repetitive calculations, the availability of automatic computers has encouraged physicists and radiotherapists to take a fresh look at many fundamental physical problems of radiotherapy. The most important result of the automation of dosage calculations is not simply an increase in the quantity of data but an improvement in the quality of data available as a treatment guide for the therapist. In October 1965 the International Atomic Energy Agency convened a panel in Vienna on the 'Use of Computers for Calculation of Dose Distributions in Radiotherapy' to assess the current status of work, provide guidelines for future research, explore the possibility of international cooperation and make recommendations to the Agency. The panel meeting was attended by 15 participants from seven countries, one observer, and two representatives of the World Health Organization. Participants contributed 20 working papers which served as the bases of discussion. By the nature of the work, computer techniques have been developed by a few advanced centres with access to large computer installations. However, several computer methods are now becoming 'routine' and can be used by institutions without facilities for research. It is hoped that the report of the Panel will provide a comprehensive view of the automatic computation of radiotherapeutic dose distributions and serve as a means of communication between present and potential users of computers

Calculation of the D-COM blind problem with computer codes PIN and RELA

International Nuclear Information System (INIS)

Pazdera, F.; Barta, O.; Smid, J.

1985-01-01

The results of the blind and post-experimental calculations of the 'D-COM Blind Problem on Fission Gas Release', performed within the framework of the IAEA coordinated research programme for 'The Development of Computer Models for Fuel Element Behaviour in Water Reactors', are presented. The results are compared with experimental data. A sensitivity study shows a possible explanation of some discrepancies between calculated and experimental results during the bump test performed after base irradiation. The calculations were performed with the computer codes PIN and RELA. Some submodels used in the calculations are also described. (author)

Computer and engineering calculations of Brazilian Tokamak-II

International Nuclear Information System (INIS)

Wang, S.; Chen, Y.; Sa, W.P. de; Nascimento, I.C.; Tuszel, A.G.; Galvao, R.M.O.; Machida, M.

1990-01-01

Analytical and computer calculations carried out by researches of Physics Institute - University of Sao Paulo (IFUSP), for defining the engineering project and constructing the TBR-II tokamak are presented. The hydrodynamics behavioue and determined parameters for magnetic confinement of the plasma were analysed. The computer code was developed using magnetohydrodynamics (MHD) equations which involve plasma interactions, magnetic field and electrical current circulating in more than 20 coils distributed around toroidal vase of the plasma. The electromagnetic, thermal and mechanical couplings are also presented. The TBR-II will be feed by two turbo-generators with 15 MW each one. (M.C.K.) [pt

Geochemical modelling. Whatif-AQ: a computer programme for speciation calculations

International Nuclear Information System (INIS)

Skytte-Jensen, B.; Jensen, H.; Nielsen, O.J.

1985-01-01

WHATIF-AQ is part of a family of programmes for calculations of geochemistry in the near-field of radioactive waste with temperature gradients. The computer programme presented in this report calculates speciation and saturation indices for an aqueous solution at temperatures between 0-125 0 C. This report also serves as a manual to the programme

RAFT: a computer program for fault tree risk calculations

International Nuclear Information System (INIS)

Seybold, G.D.

1977-11-01

A description and user instructions are presented for RAFT, a FORTRAN computer code for calculation of a risk measure for fault tree cut sets. RAFT calculates release quantities and a risk measure based on the product of probability and release quantity for cut sets of fault trees modeling the accidental release of radioactive material from a nuclear fuel cycle facility. Cut sets and their probabilities are supplied as input to RAFT from an external fault tree analysis code. Using the total inventory available of radioactive material, along with release fractions for each event in a cut set, the release terms are calculated for each cut set. Each release term is multiplied by the cut set probability to yield the cut set risk measure. RAFT orders the dominant cut sets on the risk measure. The total risk measure of processed cut sets and their fractional contributions are supplied as output. Input options are available to eliminate redundant cut sets, apply threshold values on cut set probability and risk, and control the total number of cut sets output. Hash addressing is used to remove redundant cut sets from the analysis. Computer hardware and software restrictions are given along with a sample problem and cross-reference table of the code. Except for the use of file management utilities, RAFT is written exclusively in FORTRAN language and is operational on a Control Data, CYBER 74-18--series computer system. 4 figures

Parallel diffusion calculation for the PHAETON on-line multiprocessor computer

International Nuclear Information System (INIS)

Collart, J.M.; Fedon-Magnaud, C.; Lautard, J.J.

1987-04-01

The aim of the PHAETON project is the design of an on-line computer in order to increase the immediate knowledge of the main operating and safety parameters in power plants. A significant stage is the computation of the three dimensional flux distribution. For cost and safety reason a computer based on a parallel microprocessor architecture has been studied. This paper presents a first approach to parallelized three dimensional diffusion calculation. A computing software has been written and built in a four processors demonstrator. We present the realization in progress, concerning the final equipment. 8 refs

Computer program for calculation of ideal gas thermodynamic data

Science.gov (United States)

Gordon, S.; Mc Bride, B. J.

1968-01-01

Computer program calculates ideal gas thermodynamic properties for any species for which molecular constant data is available. Partial functions and derivatives from formulas based on statistical mechanics are provided by the program which is written in FORTRAN 4 and MAP.

CRYOCOL a computer program to calculate the cryogenic distillation of hydrogen isotopes

International Nuclear Information System (INIS)

Douglas, S.R.

1993-02-01

This report describes the computer model and mathematical method coded into the AECL Research computer program CRYOCOL. The purpose of CRYOCOL is to calculate the separation of hydrogen isotopes by cryogenic distillation. (Author)

Computer calculations of air and steam blowdown suppression

International Nuclear Information System (INIS)

Norris, D.M. Jr.; McMaster, W.H.; Landram, C.S.; Quinones, D.F.; Gong, E.Y.; Macken, N.A.

1980-01-01

We describe a computer code that combines an Eulerian incompressible-fluid algorithm (SOLA) with a Lagrangian finite-element shell algorithm. The former models the fluid and the latter models the containing structure in an analysis of pressure suppression in boiling-water reactors. The code (PELE-IC) calculates loads and structural response from air blow-down and from the oscillatory condensation of steam bubbles in a water pool. The fluid, structure, and coupling algorithms are tested by recalculating problems that have known analytical solutions, including tank drainage, spherical bubble growth, coupling for circular plates, and submerged cylinder vibration. Code calculations are also compared with the results of small-scale blowdown experiments. (orig.)

Some questions of using coding theory and analytical calculation methods on computers

International Nuclear Information System (INIS)

Nikityuk, N.M.

1987-01-01

Main results of investigations devoted to the application of theory and practice of correcting codes are presented. These results are used to create very fast units for the selection of events registered in multichannel detectors of nuclear particles. Using this theory and analytical computing calculations, practically new combination devices, for example, parallel encoders, have been developed. Questions concerning the creation of a new algorithm for the calculation of digital functions by computers and problems of devising universal, dynamically reprogrammable logic modules are discussed

Traction calculation of band conveyors using the ''Nairi-2'' computer

Energy Technology Data Exchange (ETDEWEB)

Kutlunin, V A; Livshis, A V; Pod' yemshchikov, A N

1982-01-01

An algorithm is described and a program is introduced for traction calculation of band conveyors using the ''Nairi-2'' computer. The calculation system of the conveyor is derived by means of dividing the closed loop of the band into straight sections, which are separated by drums. The number of sections depends on the number of drums, the bypass system of them by the band, the shape of the conveyor route and in each specific case, a specific number is chosen. The initial information for the computer is assigned by a matrix, each row of which consists of parameters of the corresponding section. As a result, the forces of the beginning and end of the sections are found, and the required output of the drive motors with observance of the conditions of non-slippage of the drive drums and limiting of the maximum slack of the band between rollers are also found. The program allows one to make traction calculations of band conveyors with any routing shape for any number and position of the drive drums.

FLATT - a computer programme for calculating flow and temperature transients in nuclear fuels

International Nuclear Information System (INIS)

Venkat Raj, V.; Koranne, S.M.

1976-01-01

FLATT is a computer code written in Fortran language for BESM-6 computer. The code calculates the flow transients in the coolant circuit of a nuclear reactor, caused by pump failure, and the consequent temperature transients in the fuel, clad, and the coolant. In addition any desired flow transient can be fed into the programme and the resulting temperature transients can be calculated. A case study is also presented. (author)

Calculating computer-generated optical elements to produce arbitrary intensity distributions

International Nuclear Information System (INIS)

Findlay, S.; Nugent, K.A.; Scholten, R.E.

2000-01-01

Full text: We describe preliminary investigation into using a computer to generate optical elements (CGOEs) with phase-only variation, that will produce an arbitrary intensity distribution in a given image plane. An iterative calculation cycles between the CGOE and the image plane and modifies each according to the appropriate constraints. We extend this to the calculation of defined intensity distributions in two separated planes by modifying both phase and intensity at the CGOE

COMPUTER PROGRAM FOR CALCULATION MICROCHANNEL HEAT EXCHANGERS FOR AIR CONDITIONING SYSTEMS

Directory of Open Access Journals (Sweden)

Olga V. Olshevska

2016-08-01

Full Text Available Creating a computer program to calculate microchannel air condensers to reduce design time and carrying out variant calculations. Software packages for thermophysical properties of the working substance and the coolant, the correlation equation for calculating heat transfer, aerodynamics and hydrodynamics, the thermodynamic equations for the irreversible losses and their minimization in the heat exchanger were used in the process of creating. Borland Delphi 7 is used for creating software package.

GENGTC-JB: a computer program to calculate temperature distribution for cylindrical geometry capsule

International Nuclear Information System (INIS)

Someya, Hiroyuki; Kobayashi, Toshiki; Niimi, Motoji; Hoshiya, Taiji; Harayama, Yasuo

1987-09-01

In design of JMTR irradiation capsules contained specimens, a program (named GENGTC) has been generally used to evaluate temperature distributions in the capsules. The program was originally compiled by ORNL(U.S.A.) and consisted of very simple calculation methods. From the incorporated calculation methods, the program is easy to use, and has many applications to the capsule design. However, it was considered to replace original computing methods with advanced ones, when the program was checked from a standpoint of the recent computer abilities, and also to be complicated in data input. Therefore, the program was versioned up as aim to make better calculations and improve input method. The present report describes revised calculation methods and input/output guide of the version-up program. (author)

Hauser*5, a computer code to calculate nuclear cross sections

International Nuclear Information System (INIS)

Mann, F.M.

1979-07-01

HAUSER*5 is a computer code that uses the statistical (Hauser-Feshbach) model, the pre-equilibrium model, and a statistical model of direct reactions to predict nuclear cross sections. The code is unrestricted as to particle type, includes fission and capture, makes width-fluctuation corrections, and performs three-body calculations - all in minimum computer time. Transmission coefficients can be generated internally or supplied externally. This report describes equations used, necessary input, and resulting output. 2 figures, 4 tables

A computational framework for automation of point defect calculations

International Nuclear Information System (INIS)

Goyal, Anuj; Gorai, Prashun; Peng, Haowei

2017-01-01

We have developed a complete and rigorously validated open-source Python framework to automate point defect calculations using density functional theory. Furthermore, the framework provides an effective and efficient method for defect structure generation, and creation of simple yet customizable workflows to analyze defect calculations. This package provides the capability to compute widely-accepted correction schemes to overcome finite-size effects, including (1) potential alignment, (2) image-charge correction, and (3) band filling correction to shallow defects. Using Si, ZnO and In2O3 as test examples, we demonstrate the package capabilities and validate the methodology.

PULSTRI-1 computer program for mixed core pulse calculation

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Dimic, V.

1990-01-01

PUISTRI-1 is a computer code designed for calculations of the pulse parameters of TRIGA Mark II reactor with mixed core. The code is provided with data for four types of fuel elements: standard 8.5 and 12 w/o, LEU and FLIP. The pulse parameters, such as maximum power, prompt pulse energy and average fuel temperatures are calculated in adiabatic point kinetics, approximation, modified by taking into account temperature dependence of fuel temperature reactivity coefficient and thermal capacity factor averaged over all elements in the core. Maximal fuel temperature at power peaking location is calculated from total released energy using total power peaking factor and heat capacity of the element at the location of the power peaking. Results of the code were compared to data found in references (mainly General Atomics safety analysis reports) showing good agreement for all main pulse parameters. The most important parameters, average and maximal fuel temperature, are found to be systematically slightly overpredicted (20 C and 50 C, respectively). Other parameters (energy, peak power, width) agree within ± 10 % to the reference values. The code is written in FORTRAN for IBM PC computer. The input is user friendly. running time of IBM PC AT is a few seconds. It is designed for practical applications in pulse experiments as an analytical tool for predicting pulse parameters. (orig.)

Computer code PRECIP-II for the calculation of Zr-steam reaction

International Nuclear Information System (INIS)

Suzuki, Motoye; Kawasaki, Satoru; Furuta, Teruo

1978-06-01

The computer code PRECIP-II developed, a modification of S.Malang's SIMTRAN-I, is to calculate Zr-Steam reaction under LOCA conditions. Improved are the following: 1. treatment of boundary conditions at alpha/beta phase interface during temperature decrease. 2. method of time-mesh control. 3. number of input-controllable parameters, and output format. These improvements made possible physically reasonable calculations for an increased number of temperature history patterns, including the cladding temperature excursion assumed during LOCA. Calculations were made along various transient temperature histories, with the parameters so modified as to enable fitting of numerical results of weight gain, oxide thickness and alpha phase thickness in isothermal reactions to the experimental data. Then the computed results were compared with the corresponding experimental values, which revealed that most of the differences lie within +-10%. Slow cooling effect on ductility change of Zircaloy-4 was investigated with some of the oxidized specimens by a ring compression test; the effect is only slight. (auth.)

Review of calculational models and computer codes for environmental dose assessment of radioactive releases

International Nuclear Information System (INIS)

Strenge, D.L.; Watson, E.C.; Droppo, J.G.

1976-06-01

The development of technological bases for siting nuclear fuel cycle facilities requires calculational models and computer codes for the evaluation of risks and the assessment of environmental impact of radioactive effluents. A literature search and review of available computer programs revealed that no one program was capable of performing all of the great variety of calculations (i.e., external dose, internal dose, population dose, chronic release, accidental release, etc.). Available literature on existing computer programs has been reviewed and a description of each program reviewed is given

Review of calculational models and computer codes for environmental dose assessment of radioactive releases

Energy Technology Data Exchange (ETDEWEB)

Strenge, D.L.; Watson, E.C.; Droppo, J.G.

1976-06-01

The development of technological bases for siting nuclear fuel cycle facilities requires calculational models and computer codes for the evaluation of risks and the assessment of environmental impact of radioactive effluents. A literature search and review of available computer programs revealed that no one program was capable of performing all of the great variety of calculations (i.e., external dose, internal dose, population dose, chronic release, accidental release, etc.). Available literature on existing computer programs has been reviewed and a description of each program reviewed is given.

Computer program FPIP-REV calculates fission product inventory for U-235 fission

Science.gov (United States)

Brown, W. S.; Call, D. W.

1967-01-01

Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

Massive calculations of electrostatic potentials and structure maps of biopolymers in a distributed computing environment

International Nuclear Information System (INIS)

Akishina, T.P.; Ivanov, V.V.; Stepanenko, V.A.

2013-01-01

Among the key factors determining the processes of transcription and translation are the distributions of the electrostatic potentials of DNA, RNA and proteins. Calculations of electrostatic distributions and structure maps of biopolymers on computers are time consuming and require large computational resources. We developed the procedures for organization of massive calculations of electrostatic potentials and structure maps for biopolymers in a distributed computing environment (several thousands of cores).

Reactor physics simulations with coupled Monte Carlo calculation and computational fluid dynamics

International Nuclear Information System (INIS)

Seker, V.; Thomas, J.W.; Downar, T.J.

2007-01-01

A computational code system based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program 'McSTAR'. McSTAR is written in FORTRAN90 programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STARCD for every region. Three different methods were investigated and implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping between MCNP and STAR-CD cells. Preliminary testing of the code was performed using a 3x3 PWR pin-cell problem. The preliminary results are compared with those obtained from a STAR-CD coupled calculation with the deterministic transport code DeCART. Good agreement in the k eff and the power profile was observed. Increased computational capabilities and improvements in computational methods have accelerated interest in high fidelity modeling of nuclear reactor cores during the last several years. High-fidelity has been achieved by utilizing full core neutron transport solutions for the neutronics calculation and computational fluid dynamics solutions for the thermal-hydraulics calculation. Previous researchers have reported the coupling of 3D deterministic neutron transport method to CFD and their application to practical reactor analysis problems. One of the principal motivations of the work here was to utilize Monte Carlo methods to validate the coupled deterministic neutron transport

Electron correlation in molecules: concurrent computation Many-Body Perturbation Theory (ccMBPT) calculations using macrotasking on the NEC SX-3/44 computer

International Nuclear Information System (INIS)

Moncrieff, D.; Wilson, S.

1992-06-01

The ab initio determination of the electronic structure of molecules is a many-fermion problem involving the approximate description of the motion of the electrons in the field of fixed nuclei. It is an area of research which demands considerable computational resources but having enormous potential in fields as diverse as interstellar chemistry and drug design, catalysis and solid state chemistry, molecular biology and environmental chemistry. Electronic structure calculations almost invariably divide into two main stages: the approximate solution of an independent electron model, in which each electron moves in the average field created by the other electrons in the system, and then, the more computationally demanding determination of a series of corrections to this model, the electron correlation effects. The many-body perturbation theory expansion affords a systematic description of correlation effects, which leads directly to algorithms which are suitable for concurrent computation. We term this concurrent computation Many-Body Perturbation Theory (ccMBPT). The use of a dynamic load balancing technique on the NEC SX-3/44 computer in electron correlation calculations is investigated for the calculation of the most demanding energy component in the most accurate of contemporary ab initio studies. An application to the ground state of the nitrogen molecule is described. We also briefly discuss the extent to which the calculation of the dominant corrections to such studies can be rendered computationally tractable by exploiting both the vector processing and parallel processor capabilities of the NEC SX-3/44 computer. (author)

Calculating the mounting parameters for Taylor Spatial Frame correction using computed tomography.

Science.gov (United States)

Kucukkaya, Metin; Karakoyun, Ozgur; Armagan, Raffi; Kuzgun, Unal

2011-07-01

The Taylor Spatial Frame uses a computer program-based six-axis deformity analysis. However, there is often a residual deformity after the initial correction, especially in deformities with a rotational component. This problem can be resolved by recalculating the parameters and inputting all new deformity and mounting parameters. However, this may necessitate repeated x-rays and delay treatment. We believe that error in the mounting parameters is the main reason for most residual deformities. To prevent these problems, we describe a new calculation technique for determining the mounting parameters that uses computed tomography. This technique is especially advantageous for deformities with a rotational component. Using this technique, exact calculation of the mounting parameters is possible and the residual deformity and number of repeated x-rays can be minimized. This new technique is an alternative method to accurately calculating the mounting parameters.

A model for calculating the optimal replacement interval of computer systems

International Nuclear Information System (INIS)

Fujii, Minoru; Asai, Kiyoshi

1981-08-01

A mathematical model for calculating the optimal replacement interval of computer systems is described. This model is made to estimate the best economical interval of computer replacement when computing demand, cost and performance of computer, etc. are known. The computing demand is assumed to monotonously increase every year. Four kinds of models are described. In the model 1, a computer system is represented by only a central processing unit (CPU) and all the computing demand is to be processed on the present computer until the next replacement. On the other hand in the model 2, the excessive demand is admitted and may be transferred to other computing center and processed costly there. In the model 3, the computer system is represented by a CPU, memories (MEM) and input/output devices (I/O) and it must process all the demand. Model 4 is same as model 3, but the excessive demand is admitted to be processed in other center. (1) Computing demand at the JAERI, (2) conformity of Grosch's law for the recent computers, (3) replacement cost of computer systems, etc. are also described. (author)

On the theories, techniques, and computer codes used in numerical reactor criticality and burnup calculations

International Nuclear Information System (INIS)

El-Osery, I.A.

1981-01-01

The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented

Reactor physics simulations with coupled Monte Carlo calculation and computational fluid dynamics

International Nuclear Information System (INIS)

Seker, V.; Thomas, J. W.; Downar, T. J.

2007-01-01

The interest in high fidelity modeling of nuclear reactor cores has increased over the last few years and has become computationally more feasible because of the dramatic improvements in processor speed and the availability of low cost parallel platforms. In the research here high fidelity, multi-physics analyses was performed by solving the neutron transport equation using Monte Carlo methods and by solving the thermal-hydraulics equations using computational fluid dynamics. A computation tool based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program 'McSTAR' along with the verification and validation efforts. McSTAR is written in PERL programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STAR-CD for every region. Three different methods were investigated and two of them are implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping between MCNP and STAR-CD cells. The necessary input file manipulation, data file generation, normalization and multi-processor calculation settings are all done through the program flow in McSTAR. Initial testing of the code was performed using a single pin cell and a 3X3 PWR pin-cell problem. The preliminary results of the single pin-cell problem are compared with those obtained from a STAR-CD coupled calculation with the deterministic transport code De

A computer program to calculate the committed dose equivalent after the inhalation of radioactivity

International Nuclear Information System (INIS)

Van der Woude, S.

1989-03-01

A growing number of people are, as part of their occupation, at risk of being exposed to radiation originating from sources inside their bodies. The quantification of this exposure is an important part of health physics. The International Commission on Radiological Protection (ICRP) developed a first-order kinetics compartmental model to determine the transport of radioactive material through the human body. The model and the parameters involved in its use, are discussed. A versatile computer program was developed to do the following after the in vivo measurement of either the organ- or whole-body activity: calculate the original amount of radioactive material which was inhaled (intake) by employing the ICRP compartmental model of the human body; compare this intake to calculated reference levels and state any action to be taken for the case under consideration; calculate the committed dose equivalent resulting from this intake. In the execution of the above-mentioned calculations, the computer program makes provision for different aerosol particle sizes and the effect of previous intakes. Model parameters can easily be changed to take the effects of, for instance, medical intervention into account. The computer program and the organization of the data in the input files are such that the computer program can be applied to any first-order kinetics compartmental model. The computer program can also conveniently be used for research on problems related to the application of the ICRP model. 18 refs., 25 figs., 5 tabs

Development of a locally mass flux conservative computer code for calculating 3-D viscous flow in turbomachines

Science.gov (United States)

Walitt, L.

1982-01-01

The VANS successive approximation numerical method was extended to the computation of three dimensional, viscous, transonic flows in turbomachines. A cross-sectional computer code, which conserves mass flux at each point of the cross-sectional surface of computation was developed. In the VANS numerical method, the cross-sectional computation follows a blade-to-blade calculation. Numerical calculations were made for an axial annular turbine cascade and a transonic, centrifugal impeller with splitter vanes. The subsonic turbine cascade computation was generated in blade-to-blade surface to evaluate the accuracy of the blade-to-blade mode of marching. Calculated blade pressures at the hub, mid, and tip radii of the cascade agreed with corresponding measurements. The transonic impeller computation was conducted to test the newly developed locally mass flux conservative cross-sectional computer code. Both blade-to-blade and cross sectional modes of calculation were implemented for this problem. A triplet point shock structure was computed in the inducer region of the impeller. In addition, time-averaged shroud static pressures generally agreed with measured shroud pressures. It is concluded that the blade-to-blade computation produces a useful engineering flow field in regions of subsonic relative flow; and cross-sectional computation, with a locally mass flux conservative continuity equation, is required to compute the shock waves in regions of supersonic relative flow.

Microcomputers, desk calculators and process computers for use in radiation protection

International Nuclear Information System (INIS)

Burgkhardt, B.; Nolte, G.; Schollmeier, W.; Rau, G.

1983-01-01

The goals achievable, or to be pursued, in radiation protection measurement and evaluation by using computers are explained. As there is a large variety of computers available offering a likewise large variety, of performances, use of a computer is justified even for minor measuring and evaluation tasks. The subdivision into: Microcomputers as an installed part of measuring equipment; measuring and evaluation systems with desk calculators; measuring and evaluation systems with process computers is done to explain the importance and extent of the measuring or evaluation tasks and the computing devices suitable for the various purposes. The special requirements to be met in order to fulfill the different tasks are discussed, both in terms of hardware and software and in terms of skill and knowledge of the personnel, and are illustrated by an example showing the usefulness of computers in radiation protection. (orig./HP) [de

Increasing the computational speed of flash calculations with applications for compositional, transient simulations

DEFF Research Database (Denmark)

Rasmussen, Claus P.; Krejbjerg, Kristian; Michelsen, Michael Locht

2006-01-01

Approaches are presented for reducing the computation time spent on flash calculations in compositional, transient simulations. In a conventional flash calculation, the majority of the simulation time is spent on stability analysis, even for systems far into the single-phase region. A criterion has...

An approach to first principles electronic structure calculation by symbolic-numeric computation

Directory of Open Access Journals (Sweden)

Akihito Kikuchi

2013-04-01

Full Text Available There is a wide variety of electronic structure calculation cooperating with symbolic computation. The main purpose of the latter is to play an auxiliary role (but not without importance to the former. In the field of quantum physics [1-9], researchers sometimes have to handle complicated mathematical expressions, whose derivation seems almost beyond human power. Thus one resorts to the intensive use of computers, namely, symbolic computation [10-16]. Examples of this can be seen in various topics: atomic energy levels, molecular dynamics, molecular energy and spectra, collision and scattering, lattice spin models and so on [16]. How to obtain molecular integrals analytically or how to manipulate complex formulas in many body interactions, is one such problem. In the former, when one uses special atomic basis for a specific purpose, to express the integrals by the combination of already known analytic functions, may sometimes be very difficult. In the latter, one must rearrange a number of creation and annihilation operators in a suitable order and calculate the analytical expectation value. It is usual that a quantitative and massive computation follows a symbolic one; for the convenience of the numerical computation, it is necessary to reduce a complicated analytic expression into a tractable and computable form. This is the main motive for the introduction of the symbolic computation as a forerunner of the numerical one and their collaboration has won considerable successes. The present work should be classified as one such trial. Meanwhile, the use of symbolic computation in the present work is not limited to indirect and auxiliary part to the numerical computation. The present work can be applicable to a direct and quantitative estimation of the electronic structure, skipping conventional computational methods.

On application of analytical transformation system using a computer for Feynman intearal calculation

International Nuclear Information System (INIS)

Gerdt, V.P.

1978-01-01

Various systems of analytic transformations for the calculation of Feynman integrals using computers are discussed. The hyperspheric technique Which is used to calculate Feynman integrals enables to perform angular integration for a set of diagrams, thus reducing the multiplicity of integral. All calculations based on this method are made with the ASHMEDAL program. Feynman integrals are calculated in Euclidean space using integration by parts and some differential identities. Analytic calculation of Feynman integral is performed by the MACSYMA system. Dispersion method of integral calculation is implemented in the SCHOONSCHIP system, calculations based on features of Nielsen function are made using efficient SINAC and RSIN programs. A tube of basic Feynman integral parameters calculated using the above techniques is given

OPT13B and OPTIM4 - computer codes for optical model calculations

International Nuclear Information System (INIS)

Pal, S.; Srivastava, D.K.; Mukhopadhyay, S.; Ganguly, N.K.

1975-01-01

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)

Three-phase short circuit calculation method based on pre-computed surface for doubly fed induction generator

Science.gov (United States)

Ma, J.; Liu, Q.

2018-02-01

This paper presents an improved short circuit calculation method, based on pre-computed surface to determine the short circuit current of a distribution system with multiple doubly fed induction generators (DFIGs). The short circuit current, injected into power grid by DFIG, is determined by low voltage ride through (LVRT) control and protection under grid fault. However, the existing methods are difficult to calculate the short circuit current of DFIG in engineering practice due to its complexity. A short circuit calculation method, based on pre-computed surface, was proposed by developing the surface of short circuit current changing with the calculating impedance and the open circuit voltage. And the short circuit currents were derived by taking into account the rotor excitation and crowbar activation time. Finally, the pre-computed surfaces of short circuit current at different time were established, and the procedure of DFIG short circuit calculation considering its LVRT was designed. The correctness of proposed method was verified by simulation.

Shielding calculations using computer techniques; Calculo de blindajes mediante tecnicas de computacion

Energy Technology Data Exchange (ETDEWEB)

Gonzalez Portilla, M. I.; Marquez, J.

2011-07-01

Radiological protection aims to limit the ionizing radiation received by people and equipment, which in numerous occasions requires of protection shields. Although, for certain configurations, there are analytical formulas, to characterize these shields, the design setup may be very intensive in numerical calculations, therefore the most efficient from to design the shields is by means of computer programs to calculate dose and dose rates. In the present article we review the codes most frequently used to perform these calculations, and the techniques used by such codes. (Author) 13 refs.

Installation and testing of the ERANOS computer code for fast reactor calculations

International Nuclear Information System (INIS)

Gren, Milan

2010-12-01

The French ERANOS computer code was acquired and tested by solving benchmark problems. Five problems were calculated: 1D XZ Model, 1D RZ Model, 3D HEX SNR 300 reactor, 2S HEX and 3D HEX VVER 440 reactor. The multi-group diffuse approximation was used. The multiplication coefficients were compared within the first problem, neutron flux density in the calculation points was obtained within the second problem, and powers in the various reactor areas and in the assemblies were calculated within the remaining problems. (P.A.)

Semiempirical Quantum Chemical Calculations Accelerated on a Hybrid Multicore CPU-GPU Computing Platform.

Science.gov (United States)

Wu, Xin; Koslowski, Axel; Thiel, Walter

2012-07-10

In this work, we demonstrate that semiempirical quantum chemical calculations can be accelerated significantly by leveraging the graphics processing unit (GPU) as a coprocessor on a hybrid multicore CPU-GPU computing platform. Semiempirical calculations using the MNDO, AM1, PM3, OM1, OM2, and OM3 model Hamiltonians were systematically profiled for three types of test systems (fullerenes, water clusters, and solvated crambin) to identify the most time-consuming sections of the code. The corresponding routines were ported to the GPU and optimized employing both existing library functions and a GPU kernel that carries out a sequence of noniterative Jacobi transformations during pseudodiagonalization. The overall computation times for single-point energy calculations and geometry optimizations of large molecules were reduced by one order of magnitude for all methods, as compared to runs on a single CPU core.

Computer program for calculation of complex chemical equilibrium compositions and applications. Part 1: Analysis

Science.gov (United States)

Gordon, Sanford; Mcbride, Bonnie J.

1994-01-01

This report presents the latest in a number of versions of chemical equilibrium and applications programs developed at the NASA Lewis Research Center over more than 40 years. These programs have changed over the years to include additional features and improved calculation techniques and to take advantage of constantly improving computer capabilities. The minimization-of-free-energy approach to chemical equilibrium calculations has been used in all versions of the program since 1967. The two principal purposes of this report are presented in two parts. The first purpose, which is accomplished here in part 1, is to present in detail a number of topics of general interest in complex equilibrium calculations. These topics include mathematical analyses and techniques for obtaining chemical equilibrium; formulas for obtaining thermodynamic and transport mixture properties and thermodynamic derivatives; criteria for inclusion of condensed phases; calculations at a triple point; inclusion of ionized species; and various applications, such as constant-pressure or constant-volume combustion, rocket performance based on either a finite- or infinite-chamber-area model, shock wave calculations, and Chapman-Jouguet detonations. The second purpose of this report, to facilitate the use of the computer code, is accomplished in part 2, entitled 'Users Manual and Program Description'. Various aspects of the computer code are discussed, and a number of examples are given to illustrate its versatility.

DOSEFU: Computer application for dose calculation and effluent management in normal operation

International Nuclear Information System (INIS)

Martin Garcia, J. E.; Gonzalvo Manovel, A.; Revuelta Garcia, L.

2002-01-01

DOSEFU is a computer application on Windows that develops the methodology of nuclear power plant Exterior Dose Calculation Manuals (Manuals de Calculo de Dosis al Exterior-MACADE) for calculating doses in normal operation caused by radioactive liquid and gaseous effluents, for the purpose of enforcing the new Spanish Regulation on Health Protection against Ionizing Radiations, Royal Decree 783/2001 resulting from transposition of Directive 96/29/Euratom whereby the basic rules regarding health protection of workers and the population against risks resulting from ionizing radiations are established. In addition to making dose calculations, DOSEFU generates, on a magnetic support, the information regarding radioactive liquid and gaseous effluents that plants must periodically send to the CSN (ELGA format). The computer application has been developed for the specific case of Jose Cabrera NPP, which is called DOEZOR. This application can be easily implemented in any other nuclear or radioactive facility. The application is user-friendly, as the end user inputs data and executes the different modules through keys and dialogue boxes that are enabled by clicking on the mouse (see figures 2, 3, 4 and 5 ), The application runs under Windows 95. Digital Visual Fortran has been used as the development program, as this does not require additional libraries (DLLs), it can be installed in any computer without affecting other programs that are already installed. (Author)

Hamiltonian lattice field theory: Computer calculations using variational methods

International Nuclear Information System (INIS)

Zako, R.L.

1991-01-01

I develop a variational method for systematic numerical computation of physical quantities -- bound state energies and scattering amplitudes -- in quantum field theory. An infinite-volume, continuum theory is approximated by a theory on a finite spatial lattice, which is amenable to numerical computation. I present an algorithm for computing approximate energy eigenvalues and eigenstates in the lattice theory and for bounding the resulting errors. I also show how to select basis states and choose variational parameters in order to minimize errors. The algorithm is based on the Rayleigh-Ritz principle and Kato's generalizations of Temple's formula. The algorithm could be adapted to systems such as atoms and molecules. I show how to compute Green's functions from energy eigenvalues and eigenstates in the lattice theory, and relate these to physical (renormalized) coupling constants, bound state energies and Green's functions. Thus one can compute approximate physical quantities in a lattice theory that approximates a quantum field theory with specified physical coupling constants. I discuss the errors in both approximations. In principle, the errors can be made arbitrarily small by increasing the size of the lattice, decreasing the lattice spacing and computing sufficiently long. Unfortunately, I do not understand the infinite-volume and continuum limits well enough to quantify errors due to the lattice approximation. Thus the method is currently incomplete. I apply the method to real scalar field theories using a Fock basis of free particle states. All needed quantities can be calculated efficiently with this basis. The generalization to more complicated theories is straightforward. I describe a computer implementation of the method and present numerical results for simple quantum mechanical systems

Hamiltonian lattice field theory: Computer calculations using variational methods

International Nuclear Information System (INIS)

Zako, R.L.

1991-01-01

A variational method is developed for systematic numerical computation of physical quantities-bound state energies and scattering amplitudes-in quantum field theory. An infinite-volume, continuum theory is approximated by a theory on a finite spatial lattice, which is amenable to numerical computation. An algorithm is presented for computing approximate energy eigenvalues and eigenstates in the lattice theory and for bounding the resulting errors. It is shown how to select basis states and choose variational parameters in order to minimize errors. The algorithm is based on the Rayleigh-Ritz principle and Kato's generalizations of Temple's formula. The algorithm could be adapted to systems such as atoms and molecules. It is shown how to compute Green's functions from energy eigenvalues and eigenstates in the lattice theory, and relate these to physical (renormalized) coupling constants, bound state energies and Green's functions. Thus one can compute approximate physical quantities in a lattice theory that approximates a quantum field theory with specified physical coupling constants. The author discusses the errors in both approximations. In principle, the errors can be made arbitrarily small by increasing the size of the lattice, decreasing the lattice spacing and computing sufficiently long. Unfortunately, the author does not understand the infinite-volume and continuum limits well enough to quantify errors due to the lattice approximation. Thus the method is currently incomplete. The method is applied to real scalar field theories using a Fock basis of free particle states. All needed quantities can be calculated efficiently with this basis. The generalization to more complicated theories is straightforward. The author describes a computer implementation of the method and present numerical results for simple quantum mechanical systems

A computer calculation of the ternary Mo-Pd-Rh phase diagram

International Nuclear Information System (INIS)

Guerler, R.; Pratt, J.N.

1993-01-01

Thermodynamic coefficients for the phases in the binary Mo-Pd, Pd-Rh and Mo-Rh systems were derived by the assessment of the available experimental data using the binary Lukas optimization program. The resulting coefficients were first successfully utilised in reestablishing the binaries. The coefficients thus obtained in the binary computation were combined with ternary descriptions to compute ternary isothermal sections. Although no ternary interaction term was involved in the construction of the isotherms, the section calculated at 1373 K is found to be consistent with the experimentally established isothermal section at the same temperature. The location of three-phase field (bcc+hcp+fcc) and phase boundaries in both isotherms are matching reasonably well. Combining only binary coefficients of these phases, it is possible to construct reasonable isothermal sections at different temperatures. Following this conclusion, isothermal sections ranging from 1373 to 2673 K of the ternary Mo-Pd-Rh system were calculated. (orig.)

SKYSHIN: A computer code for calculating radiation dose over a barrier

International Nuclear Information System (INIS)

Atwood, C.L.; Boland, J.R.; Dickman, P.T.

1986-11-01

SKYSHIN is a computer code for calculating the radioactive dose (mrem), when there is a barrier between the point source and the receptor. The two geometrical configurations considered are: the source and receptor separated by a rectangular wall, and the source at the bottom of a cylindrical hole in the ground. Each gamma ray traveling over the barrier is assumed to be scattered at a single point. The dose to a receptor from such paths is numerically integrated for the total dose, with symmetry used to reduce the triple integral to a double integral. The buildup factor used along a straight line through air is based on published data, and extrapolated in a stable way to low energy levels. This buildup factor was validated by comparing calculated and experimental line-of-sight doses. The entire code shows good agreement to limited field data. The code runs on a CDC or on a Vax computer, and could be modified easily for others

Direct Monte Carlo dose calculation using polygon-surface computational human model

International Nuclear Information System (INIS)

Jeong, Jong Hwi; Kim, Chan Hyeong; Yeom, Yeon Su; Cho, Sungkoo; Chung, Min Suk; Cho, Kun-Woo

2011-01-01

In the present study, a voxel-type computational human model was converted to a polygon-surface model, after which it was imported directly to the Geant4 code without using a voxelization process, that is, without converting back to a voxel model. The original voxel model was also imported to the Geant4 code, in order to compare the calculated dose values and the computational speed. The average polygon size of the polygon-surface model was ∼0.5 cm 2 , whereas the voxel resolution of the voxel model was 1.981 × 1.981 × 2.0854 mm 3 . The results showed a good agreement between the calculated dose values of the two models. The polygon-surface model was, however, slower than the voxel model by a factor of 6–9 for the photon energies and irradiation geometries considered in the present study, which nonetheless is considered acceptable, considering that direct use of the polygon-surface model does not require a separate voxelization process. (author)

SUBDOSA: a computer program for calculating external doses from accidental atmospheric releases of radionuclides

International Nuclear Information System (INIS)

Strenge, D.L.; Watson, E.C.; Houston, J.R.

1975-06-01

A computer program, SUBDOSA, was developed for calculating external γ and β doses to individuals from the accidental release of radionuclides to the atmosphere. Characteristics of SUBDOSA are: doses from both γ and β radiation are calculated as a function of depth in tissue, summed and reported as skin, eye, gonadal, and total body dose; doses are calculated for releases within each of several release time intervals and nuclide inventories and atmospheric dispersion conditions are considered for each time interval; radioactive decay is considered during the release and/or transit using a chain decay scheme with branching to account for transitions to and from isomeric states; the dose from gamma radiation is calculated using a numerical integration technique to account for the finite size of the plume; and the program computes and lists the normalized air concentrations at ground level as a function of distance from the point of release. (auth)

Computer codes used in the calculation of high-temperature thermodynamic properties of sodium

International Nuclear Information System (INIS)

Fink, J.K.

1979-12-01

Three computer codes - SODIPROP, NAVAPOR, and NASUPER - were written in order to calculate a self-consistent set of thermodynamic properties for saturated, subcooled, and superheated sodium. These calculations incorporate new critical parameters (temperature, pressure, and density) and recently derived single equations for enthalpy and vapor pressure. The following thermodynamic properties have been calculated in these codes: enthalpy, heat capacity, entropy, vapor pressure, heat of vaporization, density, volumetric thermal expansion coefficient, compressibility, and thermal pressure coefficient. In the code SODIPROP, these properties are calculated for saturated and subcooled liquid sodium. Thermodynamic properties of saturated sodium vapor are calculated in the code NAVAPOR. The code NASUPER calculates thermodynamic properties for super-heated sodium vapor only for low (< 1644 K) temperatures. No calculations were made for the supercritical region

Calculation of radiation dose received in computed tomography examinations

International Nuclear Information System (INIS)

Abed Elseed, Eslam Mustafa

2014-07-01

Diagnostic computed tomography (CT) examinations play an important role in the health care of the population. These examination may involve significant irradiation of the patient and probably represent the largest man-made source of radiation exposure for the population. This study was performed to assess the effective dose (ED) received in brain CT examination ( base of skull and cerebrum) and to analyze effective dose distributions among radiological departments under study. The study was performed at Elnileen Medical Center, coverage one CT unit and a sample of 51 patients (25 cerebrum sample and 26 base of skull sample). The following parameters were recorded age, weight, height body mass index (BMI) derived from weight (kg) and height ( m) and exposure factor and CTDI voi , DLP value. The effective dose was measured for brain CT examination. The ED values were calculated from the obtained DLP values using AAPM report No 96 calculation methods. The results of ED values calculated showed that patient exposure were within the normal range of exposure. The mean ED values calculated were 0.35±0.15 for base of skull of brain CT examinations and 0.70±0.32 for cerebrum of brain CT examination, respectively. Further studies are recommended with more number of pa.(Author)

Dynamic stability calculations for power grids employing a parallel computer

Energy Technology Data Exchange (ETDEWEB)

Schmidt, K

1982-06-01

The aim of dynamic contingency calculations in power systems is to estimate the effects of assumed disturbances, such as loss of generation. Due to the large dimensions of the problem these simulations require considerable computing time and costs, to the effect that they are at present only used in a planning state but not for routine checks in power control stations. In view of the homogeneity of the problem, where a multitude of equal generator models, having different parameters, are to be integrated simultaneously, the use of a parallel computer looks very attractive. The results of this study employing a prototype parallel computer (SMS 201) are presented. It consists of up to 128 equal microcomputers bus-connected to a control computer. Each of the modules is programmed to simulate a node of the power grid. Generators with their associated control are represented by models of 13 states each. Passive nodes are complemented by 'phantom'-generators, so that the whole power grid is homogenous, thus removing the need for load-flow-iterations. Programming of microcomputers is essentially performed in FORTRAN.

Large-scale theoretical calculations in molecular science - design of a large computer system for molecular science and necessary conditions for future computers

Energy Technology Data Exchange (ETDEWEB)

Kashiwagi, H [Institute for Molecular Science, Okazaki, Aichi (Japan)

1982-06-01

A large computer system was designed and established for molecular science under the leadership of molecular scientists. Features of the computer system are an automated operation system and an open self-service system. Large-scale theoretical calculations have been performed to solve many problems in molecular science, using the computer system. Necessary conditions for future computers are discussed on the basis of this experience.

Large-scale theoretical calculations in molecular science - design of a large computer system for molecular science and necessary conditions for future computers

International Nuclear Information System (INIS)

Kashiwagi, H.

1982-01-01

A large computer system was designed and established for molecular science under the leadership of molecular scientists. Features of the computer system are an automated operation system and an open self-service system. Large-scale theoretical calculations have been performed to solve many problems in molecular science, using the computer system. Necessary conditions for future computers are discussed on the basis of this experience. (orig.)

Easy calculations of lod scores and genetic risks on small computers.

Science.gov (United States)

Lathrop, G M; Lalouel, J M

1984-01-01

A computer program that calculates lod scores and genetic risks for a wide variety of both qualitative and quantitative genetic traits is discussed. An illustration is given of the joint use of a genetic marker, affection status, and quantitative information in counseling situations regarding Duchenne muscular dystrophy. PMID:6585139

CREST : a computer program for the calculation of composition dependent self-shielded cross-sections

International Nuclear Information System (INIS)

Kapil, S.K.

1977-01-01

A computer program CREST for the calculation of the composition and temperature dependent self-shielded cross-sections using the shielding factor approach has been described. The code includes the editing and formation of the data library, calculation of the effective shielding factors and cross-sections, a fundamental mode calculation to generate the neutron spectrum for the system which is further used to calculate the effective elastic removal cross-sections. Studies to explore the sensitivity of reactor parameters to changes in group cross-sections can also be carried out by using the facility available in the code to temporarily change the desired constants. The final self-shielded and transport corrected group cross-sections can be dumped on cards or magnetic tape in a suitable form for their direct use in a transport or diffusion theory code for detailed reactor calculations. The program is written in FORTRAN and can be accommodated in a computer with 32 K work memory. The input preparation details, sample problem and the listing of the program are given. (author)

Theoretical background and user's manual for the computer code on groundwater flow and radionuclide transport calculation in porous rock

International Nuclear Information System (INIS)

Shirakawa, Toshihiko; Hatanaka, Koichiro

2001-11-01

In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostatistical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report. (author)

Development of a computational methodology for internal dose calculations

International Nuclear Information System (INIS)

Yoriyaz, Helio

2000-01-01

A new approach for calculating internal dose estimates was developed through the use of a more realistic computational model of the human body and a more precise tool for the radiation transport simulation. The present technique shows the capability to build a patient-specific phantom with tomography data (a voxel-based phantom) for the simulation of radiation transport and energy deposition using Monte Carlo methods such as in the MCNP-4B code. In order to utilize the segmented human anatomy as a computational model for the simulation of radiation transport, an interface program, SCMS, was developed to build the geometric configurations for the phantom through the use of tomographic images. This procedure allows to calculate not only average dose values but also spatial distribution of dose in regions of interest. With the present methodology absorbed fractions for photons and electrons in various organs of the Zubal segmented phantom were calculated and compared to those reported for the mathematical phantoms of Snyder and Cristy-Eckerman. Although the differences in the organ's geometry between the phantoms are quite evident, the results demonstrate small discrepancies, however, in some cases, considerable discrepancies were found due to two major causes: differences in the organ masses between the phantoms and the occurrence of organ overlap in the Zubal segmented phantom, which is not considered in the mathematical phantom. This effect was quite evident for organ cross-irradiation from electrons. With the determination of spatial dose distribution it was demonstrated the possibility of evaluation of more detailed doses data than those obtained in conventional methods, which will give important information for the clinical analysis in therapeutic procedures and in radiobiologic studies of the human body. (author)

Calculation and evaluation methodology of the flawed pipe and the compute program development

International Nuclear Information System (INIS)

Liu Chang; Qian Hao; Yao Weida; Liang Xingyun

2013-01-01

Background: The crack will grow gradually under alternating load for a pressurized pipe, whereas the load is less than the fatigue strength limit. Purpose: Both calculation and evaluation methodology for a flawed pipe that have been detected during in-service inspection is elaborated here base on the Elastic Plastic Fracture Mechanics (EPFM) criteria. Methods: In the compute, the depth and length interaction of a flaw has been considered and a compute program is developed per Visual C++. Results: The fluctuating load of the Reactor Coolant System transients, the initial flaw shape, the initial flaw orientation are all accounted here. Conclusions: The calculation and evaluation methodology here is an important basis for continue working or not. (authors)

Depth compensating calculation method of computer-generated holograms using symmetry and similarity of zone plates

Science.gov (United States)

Wei, Hui; Gong, Guanghong; Li, Ni

2017-10-01

Computer-generated hologram (CGH) is a promising 3D display technology while it is challenged by heavy computation load and vast memory requirement. To solve these problems, a depth compensating CGH calculation method based on symmetry and similarity of zone plates is proposed and implemented on graphics processing unit (GPU). An improved LUT method is put forward to compute the distances between object points and hologram pixels in the XY direction. The concept of depth compensating factor is defined and used for calculating the holograms of points with different depth positions instead of layer-based methods. The proposed method is suitable for arbitrary sampling objects with lower memory usage and higher computational efficiency compared to other CGH methods. The effectiveness of the proposed method is validated by numerical and optical experiments.

SIMCRI: a simple computer code for calculating nuclear criticality parameters

International Nuclear Information System (INIS)

Nakamaru, Shou-ichi; Sugawara, Nobuhiko; Naito, Yoshitaka; Katakura, Jun-ichi; Okuno, Hiroshi.

1986-03-01

This is a user's manual for a simple criticality calculation code SIMCRI. The code has been developed to facilitate criticality calculation on a single unit of nuclear fuel. SIMCRI makes an extensive survey with a little computing time. Cross section library MGCL for SIMCRI is the same one for the Monte Carlo criticality code KENOIV; it is, therefore, easy to compare the results of the two codes. SIMCRI solves eigenvalue problems and fixed source problems based on the one space point B 1 equation. The results include infinite and effective multiplication factor, critical buckling, migration area, diffusion coefficient and so on. SIMCRI is comprised in the criticality safety evaluation code system JACS. (author)

Mathematical model of accelerator output characteristics and their calculation on a computer

International Nuclear Information System (INIS)

Mishulina, O.A.; Ul'yanina, M.N.; Kornilova, T.V.

1975-01-01

A mathematical model is described of output characteristics of a linear accelerator. The model is a system of differential equations. Presence of phase limitations is a specific feature of setting the problem which makes it possible to ensure higher simulation accuracy and determine a capture coefficient. An algorithm is elaborated of computing output characteristics based upon the mathematical model suggested. A capture coefficient, coordinate expectation characterizing an average phase value of the beam particles, coordinate expectation characterizing an average value of the reverse relative velocity of the beam particles as well as dispersion of these coordinates are output characteristics of the accelerator. Calculation methods of the accelerator output characteristics are described in detail. The computations have been performed on the BESM-6 computer, the characteristics computing time being 2 min 20 sec. Relative error of parameter computation averages 10 -2

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

International Nuclear Information System (INIS)

Bard, F.E.; Christensen, B.Y.; Gneiting, B.C.

1980-04-01

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

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.

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

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N.

2007-01-01

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

Computer code for calculating personnel doses due to tritium exposures

International Nuclear Information System (INIS)

Graham, C.L.; Parlagreco, J.R.

1977-01-01

This report describes a computer code written in LLL modified Fortran IV that can be used on a CDC 7600 for calculating personnel doses due to internal exposures to tritium. The code is capable of handling various exposure situations and is also capable of detecting a large variety of data input errors that would lead to errors in the dose assessment. The critical organ is the body water

Algorithm of calculation of multicomponent system eutectics using electronic digital computer

International Nuclear Information System (INIS)

Posypajko, V.I.; Stratilatov, B.V.; Pervikova, V.I.; Volkov, V.Ya.

1975-01-01

A computer algorithm is proposed for determining low-temperature equilibrium regions for existing phases. The algorithm has been used in calculating nonvariant parameters (temperatures of melting of eutectics and the concentrations of their components) for a series of trinary systems, among which are Ksub(long)Cl, WO 4 , SO 4 (x 1 =K 2 WO 4 ; x 2 =K 2 SO 4 ), Ag, Cd, Pbsub(long)Cl(x 1 =CdCl 2 , x 2 =PbCl 2 ); Ksub(long)F, Cl, I (x 1 =KF, x 2 =KI). The proposed method of calculating eutectics permits the planning of the subsequent experiment in determining the parameters of the eutectics of multicomponent systems and the forecasting of chemical interaction in such systems. The algorithm can be used in calculating systems containing any number of components

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

International Nuclear Information System (INIS)

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

1983-02-01

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

Calculation of Plutonium content in RSG-GAS spent fuel using IAFUEL computer code

International Nuclear Information System (INIS)

Mochamad-Imron

2003-01-01

It has been calculated the contain of isotopes Pu-239, Pu-240, Pu-241, and isotope Pu-242 in MTR reactor fuel types which have U-235 contain about 250 gram. The calculation was performed in three steps. The first step is to determine the library of calculation output of BOC (Beginning of Cycle). The second step is to determine the core isotope density, the weight of plutonium for one core, and one fuel isotope density. The third step is to calculate weight of plutonium in gram. All calculation is performed by IAFUEL computer code. The calculation was produced content of each Pu isotopes were Pu-239 is 6.7666 gr, Pu-240 is 1.4628 gr, Pu-241 is 0.52951 gr, and Pu-242 is 0.068952 gr

Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

Science.gov (United States)

Barber, Duncan Henry

During some postulated accidents at nuclear power stations, fuel cooling may be impaired. In such cases, the fuel heats up and the subsequent increased fission-gas release from the fuel to the gap may result in fuel sheath failure. After fuel sheath failure, the barrier between the coolant and the fuel pellets is lost or impaired, gases and vapours from the fuel-to-sheath gap and other open voids in the fuel pellets can be vented. Gases and steam from the coolant can enter the broken fuel sheath and interact with the fuel pellet surfaces and the fission-product inclusion on the fuel surface (including material at the surface of the fuel matrix). The chemistry of this interaction is an important mechanism to model in order to assess fission-product releases from fuel. Starting in 1995, the computer program SOURCE 2.0 was developed by the Canadian nuclear industry to model fission-product release from fuel during such accidents. SOURCE 2.0 has employed an early thermochemical model of irradiated uranium dioxide fuel developed at the Royal Military College of Canada. To overcome the limitations of computers of that time, the implementation of the RMC model employed lookup tables to pre-calculated equilibrium conditions. In the intervening years, the RMC model has been improved, the power of computers has increased significantly, and thermodynamic subroutine libraries have become available. This thesis is the result of extensive work based on these three factors. A prototype computer program (referred to as SC11) has been developed that uses a thermodynamic subroutine library to calculate thermodynamic equilibria using Gibbs energy minimization. The Gibbs energy minimization requires the system temperature (T) and pressure (P), and the inventory of chemical elements (n) in the system. In order to calculate the inventory of chemical elements in the fuel, the list of nuclides and nuclear isomers modelled in SC11 had to be expanded from the list used by SOURCE 2.0. A

PABLM: a computer program to calculate accumulated radiation doses from radionuclides in the environment

Energy Technology Data Exchange (ETDEWEB)

Napier, B.A.; Kennedy, W.E. Jr.; Soldat, J.K.

1980-03-01

A computer program, PABLM, was written to facilitate the calculation of internal radiation doses to man from radionuclides in food products and external radiation doses from radionuclides in the environment. This report contains details of mathematical models used and calculational procedures required to run the computer program. Radiation doses from radionuclides in the environment may be calculated from deposition on the soil or plants during an atmospheric or liquid release, or from exposure to residual radionuclides in the environment after the releases have ended. Radioactive decay is considered during the release of radionuclides, after they are deposited on the plants or ground, and during holdup of food after harvest. The radiation dose models consider several exposure pathways. Doses may be calculated for either a maximum-exposed individual or for a population group. The doses calculated are accumulated doses from continuous chronic exposure. A first-year committed dose is calculated as well as an integrated dose for a selected number of years. The equations for calculating internal radiation doses are derived from those given by the International Commission on Radiological Protection (ICRP) for body burdens and MPC's of each radionuclide. The radiation doses from external exposure to contaminated water and soil are calculated using the basic assumption that the contaminated medium is large enough to be considered an infinite volume or plane relative to the range of the emitted radiations. The equations for calculations of the radiation dose from external exposure to shoreline sediments include a correction for the finite width of the contaminated beach.

PABLM: a computer program to calculate accumulated radiation doses from radionuclides in the environment

International Nuclear Information System (INIS)

Napier, B.A.; Kennedy, W.E. Jr.; Soldat, J.K.

1980-03-01

A computer program, PABLM, was written to facilitate the calculation of internal radiation doses to man from radionuclides in food products and external radiation doses from radionuclides in the environment. This report contains details of mathematical models used and calculational procedures required to run the computer program. Radiation doses from radionuclides in the environment may be calculated from deposition on the soil or plants during an atmospheric or liquid release, or from exposure to residual radionuclides in the environment after the releases have ended. Radioactive decay is considered during the release of radionuclides, after they are deposited on the plants or ground, and during holdup of food after harvest. The radiation dose models consider several exposure pathways. Doses may be calculated for either a maximum-exposed individual or for a population group. The doses calculated are accumulated doses from continuous chronic exposure. A first-year committed dose is calculated as well as an integrated dose for a selected number of years. The equations for calculating internal radiation doses are derived from those given by the International Commission on Radiological Protection (ICRP) for body burdens and MPC's of each radionuclide. The radiation doses from external exposure to contaminated water and soil are calculated using the basic assumption that the contaminated medium is large enough to be considered an infinite volume or plane relative to the range of the emitted radiations. The equations for calculations of the radiation dose from external exposure to shoreline sediments include a correction for the finite width of the contaminated beach

A compilation of structural property data for computer impact calculation (1/5)

International Nuclear Information System (INIS)

Ikushima, Takeshi; Nagata, Norio.

1988-10-01

The paper describes structural property data for computer impact calculations of nuclear fuel shipping casks. Four kinds of material data, mild steel, stainless steel, lead and wood are compiled. These materials are main structural elements of shipping casks. Structural data such as, the coefficient of thermal expansion, the modulus of longitudinal elasticity, the modulus of transverse elasticity, the Poisson's ratio and stress and strain relationships, have been tabulated against temperature or strain rate. This volume 1 involve structural property data and data processing computer program. (author)

Easy-to-use application programs for decay heat and delayed neutron calculations on personal computers

Energy Technology Data Exchange (ETDEWEB)

Oyamatsu, Kazuhiro [Nagoya Univ. (Japan)

1998-03-01

Application programs for personal computers are developed to calculate the decay heat power and delayed neutron activity from fission products. The main programs can be used in any computers from personal computers to main frames because their sources are written in Fortran. These programs have user friendly interfaces to be used easily not only for research activities but also for educational purposes. (author)

Calculation of heat-kernel coefficients and usage of computer algebra

International Nuclear Information System (INIS)

Bel'kov, A.A.; Lanev, A.V.; Schaale, A.

1995-01-01

The calculation of heat-kernel coefficients with the classical De Witt algorithm has been discussed. We present the explicit form of the coefficients up to h 5 in the general case and up to h 7 min for the minimal parts. The results are compared with the expressions in other papers. A method to optimize the usage of memory for working with large expressions on universal computer algebra systems has been proposed. 20 refs

Activity computer program for calculating ion irradiation activation

Science.gov (United States)

Palmer, Ben; Connolly, Brian; Read, Mark

2017-07-01

A computer program, Activity, was developed to predict the activity and gamma lines of materials irradiated with an ion beam. It uses the TENDL (Koning and Rochman, 2012) [1] proton reaction cross section database, the Stopping and Range of Ions in Matter (SRIM) (Biersack et al., 2010) code, a Nuclear Data Services (NDS) radioactive decay database (Sonzogni, 2006) [2] and an ENDF gamma decay database (Herman and Chadwick, 2006) [3]. An extended version of Bateman's equation is used to calculate the activity at time t, and this equation is solved analytically, with the option to also solve by numeric inverse Laplace Transform as a failsafe. The program outputs the expected activity and gamma lines of the activated material.

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

International Nuclear Information System (INIS)

Alvarenga, M.A.B.

1984-12-01

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) [pt

High-speed algorithm for calculating the neutron field in a reactor when working in dialog mode with a computer

International Nuclear Information System (INIS)

Afanas'ev, A.M.

1987-01-01

The large-scale construction of atomic power stations results in a need for trainers to instruct power-station personnel. The present work considers one problem of developing training computer software, associated with the development of a high-speed algorithm for calculating the neutron field after control-rod (CR) shift by the operator. The case considered here is that in which training units are developed on the basis of small computers of SM-2 type, which fall significantly short of the BESM-6 and EC-type computers used for the design calculations, in terms of speed and memory capacity. Depending on the apparatus for solving the criticality problem, in a two-dimensional single-group approximation, the physical-calculation programs require ∼ 1 min of machine time on a BESM-6 computer, which translates to ∼ 10 min on an SM-2 machine. In practice, this time is even longer, since ultimately it is necessary to determine not the effective multiplication factor K/sub ef/, but rather the local perturbations of the emergency-control (EC) system (to reach criticality) and change in the neutron field on shifting the CR and the EC rods. This long time means that it is very problematic to use physical-calculation programs to work in dialog mode with a computer. The algorithm presented below allows the neutron field following shift of the CR and EC rods to be calculated in a few seconds on a BESM-6 computer (tens of second on an SM-2 machine. This high speed may be achieved as a result of the preliminary calculation of the influence function (IF) for each CR. The IF may be calculated at high speed on a computer. Then it is stored in the external memory (EM) and, where necessary, used as the initial information

Calculating the nutrient composition of recipes with computers.

Science.gov (United States)

Powers, P M; Hoover, L W

1989-02-01

The objective of this research project was to compare the nutrient values computed by four commonly used computerized recipe calculation methods. The four methods compared were the yield factor, retention factor, summing, and simplified retention factor methods. Two versions of the summing method were modeled. Four pork entrée recipes were selected for analysis: roast pork, pork and noodle casserole, pan-broiled pork chops, and pork chops with vegetables. Assumptions were made about changes expected to occur in the ingredients during preparation and cooking. Models were designed to simulate the algorithms of the calculation methods using a microcomputer spreadsheet software package. Identical results were generated in the yield factor, retention factor, and summing-cooked models for roast pork. The retention factor and summing-cooked models also produced identical results for the recipe for pan-broiled pork chops. The summing-raw model gave the highest value for water in all four recipes and the lowest values for most of the other nutrients. A superior method or methods was not identified. However, on the basis of the capabilities provided with the yield factor and retention factor methods, more serious consideration of these two methods is recommended.

Summaries of recent computer-assisted Feynam diagram calculations

International Nuclear Information System (INIS)

Mark Fischler

2001-01-01

The AIHENP Workshop series has traditionally included cutting edge work on automated computation of Feynman diagrams. The conveners of the Symbolic Problem Solving topic in this ACAT conference felt it would be useful to solicit presentations of brief summaries of the interesting recent calculations. Since this conference was the first in the series to be held in the Western Hemisphere, it was decided that the summaries would be solicited both from attendees and from researchers who could not attend the conference. This would represent a sampling of many of the key calculations being performed. The results were presented at the Poster session; contributions from ten researchers were displayed and posted on the web. Although the poster presentation, which can be viewed at conferences.fnal.gov/acat2000/ placed equal emphasis on results presented at the conference and other contributions, here we primarily discuss the latter, which do not appear in full form in these proceedings. This brief paper can't do full justice to each contribution; interested readers can find details of the work not presented at this conference in references (1), (2), (3), (4), (5), (6), (7)

TRANGE: computer code to calculate the energy beam degradation in target stack

International Nuclear Information System (INIS)

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

FISPRO: a simplified computer program for general fission product formation and decay calculations

International Nuclear Information System (INIS)

Jiacoletti, R.J.; Bailey, P.G.

1979-08-01

This report describes a computer program that solves a general form of the fission product formation and decay equations over given time steps for arbitrary decay chains composed of up to three nuclides. All fission product data and operational history data are input through user-defined input files. The program is very useful in the calculation of fission product activities of specific nuclides for various reactor operational histories and accident consequence calculations

The presence of mathematics and computer anxiety in nursing students and their effects on medication dosage calculations.

Science.gov (United States)

Glaister, Karen

2007-05-01

To determine if the presence of mathematical and computer anxiety in nursing students affects learning of dosage calculations. The quasi-experimental study compared learning outcomes at differing levels of mathematical and computer anxiety when integrative and computer based learning approaches were used. Participants involved a cohort of second year nursing students (n=97). Mathematical anxiety exists in 20% (n=19) of the student nurse population, and 14% (n=13) experienced mathematical testing anxiety. Those students more anxious about mathematics and the testing of mathematics benefited from integrative learning to develop conditional knowledge (F(4,66)=2.52 at pComputer anxiety was present in 12% (n=11) of participants, with those reporting medium and high levels of computer anxiety performing less well than those with low levels (F(1,81)=3.98 at pmathematical and computer anxiety when planning an educational program to develop competency in dosage calculations.

Non-Intrusive Computational Method and Uncertainty Quantification Tool for isolator operability calculations, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Computational fluid dynamics (CFD) simulations are extensively used by NASA for hypersonic aerothermodynamics calculations. The physical models used in CFD codes and...

Meso-microstructural computational simulation of the hydrogen permeation test to calculate intergranular, grain boundary and effective diffusivities

Energy Technology Data Exchange (ETDEWEB)

Jothi, S., E-mail: s.jothi@swansea.ac.uk [College of Engineering, Swansea University, Singleton Park, Swansea SA2 8PP (United Kingdom); Winzer, N. [Fraunhofer Institute for Mechanics of Materials IWM, Wöhlerstraße 11, 79108 Freiburg (Germany); Croft, T.N.; Brown, S.G.R. [College of Engineering, Swansea University, Singleton Park, Swansea SA2 8PP (United Kingdom)

2015-10-05

Highlights: • Characterized polycrystalline nickel microstructure using EBSD analysis. • Development meso-microstructural model based on real microstructure. • Calculated effective diffusivity using experimental electrochemical permeation test. • Calculated intergranular diffusivity of hydrogen using computational FE simulation. • Validated the calculated computation simulation results with experimental results. - Abstract: Hydrogen induced intergranular embrittlement has been identified as a cause of failure of aerospace components such as combustion chambers made from electrodeposited polycrystalline nickel. Accurate computational analysis of this process requires knowledge of the differential in hydrogen transport in the intergranular and intragranular regions. The effective diffusion coefficient of hydrogen may be measured experimentally, though experimental measurement of the intergranular grain boundary diffusion coefficient of hydrogen requires significant effort. Therefore an approach to calculate the intergranular GB hydrogen diffusivity using finite element analysis was developed. The effective diffusivity of hydrogen in polycrystalline nickel was measured using electrochemical permeation tests. Data from electron backscatter diffraction measurements were used to construct microstructural representative volume elements including details of grain size and shape and volume fraction of grains and grain boundaries. A Python optimization code has been developed for the ABAQUS environment to calculate the unknown grain boundary diffusivity.

Hot particle dose calculations using the computer code VARSKIN Mod 2

International Nuclear Information System (INIS)

Durham, J.S.

1991-01-01

The only calculational model recognised by the Nuclear Regulatory Commission (NRC) for hot particle dosimetry is VARSKIN Mod 1. Because the code was designed to calculate skin dose from distributed skin contamination and not hot particles, it is assumed that the particle has no thickness and, therefore, that no self-absorption occurs within the source material. For low energy beta particles such as those emitted from 60 Co, a significant amount of self-shielding occurs in hot particles and VARSKIN Mod 1 overestimates the skin dose. In addition, the presence of protective clothing, which will reduce the calculated skin dose for both high and low energy beta emitters, is not modelled in VARSKIN Mod 1. Finally, there is no provision in VARSKIN Mod 1 to calculate the gamma contribution to skin dose from radionuclides that emit both beta and gamma radiation. The computer code VARSKIN Mod 1 has been modified to model three-dimensional sources, insertion of layers of protective clothing between the source and skin, and gamma dose from appropriate radionuclides. The new code, VARSKIN Mod 2, is described and the sensitivity of the calculated dose to source geometry, diameter, thickness, density, and protective clothing thickness are discussed. Finally, doses calculated using VARSKIN Mod 2 are compared to doses measured from hot particles found in nuclear power plants. (author)

VAMPIR - A two-group two-dimensional diffusion computer code for burnup calculation

International Nuclear Information System (INIS)

Zmijarevic, I.; Petrovic, I.

1985-01-01

VAMPIR is a computer code which simulates the burnup within a reactor coe. It computes the neutron flux, power distribution and burnup taking into account spatial variations of temperature and xenon poisoning. Its overall reactor calculation uses diffusion theory with finite differences approximation in X-Y or R-Z geometry. Two-group macroscopic cross section data are prepared by the lattice cell code WIMS-D4 and stored in the library form of multi entry tabulation against the various parameters that significantly affect the physical conditions in the reactor core. herein, the main features of the program are presented. (author)

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.

Quantum computing applied to calculations of molecular energies: CH2 benchmark.

Science.gov (United States)

Veis, Libor; Pittner, Jiří

2010-11-21

Quantum computers are appealing for their ability to solve some tasks much faster than their classical counterparts. It was shown in [Aspuru-Guzik et al., Science 309, 1704 (2005)] that they, if available, would be able to perform the full configuration interaction (FCI) energy calculations with a polynomial scaling. This is in contrast to conventional computers where FCI scales exponentially. We have developed a code for simulation of quantum computers and implemented our version of the quantum FCI algorithm. We provide a detailed description of this algorithm and the results of the assessment of its performance on the four lowest lying electronic states of CH(2) molecule. This molecule was chosen as a benchmark, since its two lowest lying (1)A(1) states exhibit a multireference character at the equilibrium geometry. It has been shown that with a suitably chosen initial state of the quantum register, one is able to achieve the probability amplification regime of the iterative phase estimation algorithm even in this case.

Improved Patient Size Estimates for Accurate Dose Calculations in Abdomen Computed Tomography

Energy Technology Data Exchange (ETDEWEB)

Lee, Chang-Lae [Yonsei University, Wonju (Korea, Republic of)

2017-07-15

The radiation dose of CT (computed tomography) is generally represented by the CTDI (CT dose index). CTDI, however, does not accurately predict the actual patient doses for different human body sizes because it relies on a cylinder-shaped head (diameter : 16 cm) and body (diameter : 32 cm) phantom. The purpose of this study was to eliminate the drawbacks of the conventional CTDI and to provide more accurate radiation dose information. Projection radiographs were obtained from water cylinder phantoms of various sizes, and the sizes of the water cylinder phantoms were calculated and verified using attenuation profiles. The effective diameter was also calculated using the attenuation of the abdominal projection radiographs of 10 patients. When the results of the attenuation-based method and the geometry-based method shown were compared with the results of the reconstructed-axial-CT-image-based method, the effective diameter of the attenuation-based method was found to be similar to the effective diameter of the reconstructed-axial-CT-image-based method, with a difference of less than 3.8%, but the geometry-based method showed a difference of less than 11.4%. This paper proposes a new method of accurately computing the radiation dose of CT based on the patient sizes. This method computes and provides the exact patient dose before the CT scan, and can therefore be effectively used for imaging and dose control.

Comparison of computer code calculations with FEBA test data

International Nuclear Information System (INIS)

Zhu, Y.M.

1988-06-01

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.) [de

Fatigue analysis - computation of the actual strain range using elastic calculations (factor Ke)

International Nuclear Information System (INIS)

Roche, R.L.

1987-01-01

Pressure vessels are not eternal, their life is not endless, but must be long enough for profitable use. Fatigue is the most important damage limiting life time. It is due to variable loading and especially to deformation-controlled loading like thermal dilatation (thermal stress). Hence, it is of prime importance to perform an fatigue analysis in the design phase in order to be sure the pressure vessel life meet requirement of the design specification. It is also useful to perform such an analysis for assessing the remaining life. To compute the fatigue damage, knowledge of the strain range is needed. As calculation taking into account non linear behavior of the material are very expensive and not always reliable, the current practice is using elastic computation. The aim of this paper is to discuss the methods for correcting the elastically calculated strain range and to propose a sound and practical method

Computer codes for the calculation of vibrations in machines and structures

International Nuclear Information System (INIS)

1989-01-01

After an introductory paper on the typical requirements to be met by vibration calculations, the first two sections of the conference papers present universal as well as specific finite-element codes tailored to solve individual problems. The calculation of dynamic processes increasingly now in addition to the finite elements applies the method of multi-component systems which takes into account rigid bodies or partial structures and linking and joining elements. This method, too, is explained referring to universal computer codes and to special versions. In mechanical engineering, rotary vibrations are a major problem, and under this topic, conference papers exclusively deal with codes that also take into account special effects such as electromechanical coupling, non-linearities in clutches, etc. (orig./HP) [de

Computer code calculations of the TMI-2 accident: initial and boundary conditions

International Nuclear Information System (INIS)

Behling, S.R.

1985-05-01

Initial and boundary conditions during the Three Mile Island Unit 2 (TMI-2) accident are described and detailed. A brief description of the TMI-2 plant configuration is given. Important contributions to the progression of the accident in the reactor coolant system are discussed. Sufficient information is provided to allow calculation of the TMI-2 accident with computer codes

Uncertainties in source term calculations generated by the ORIGEN2 computer code for Hanford Production Reactors

International Nuclear Information System (INIS)

Heeb, C.M.

1991-03-01

The ORIGEN2 computer code is the primary calculational tool for computing isotopic source terms for the Hanford Environmental Dose Reconstruction (HEDR) Project. The ORIGEN2 code computes the amounts of radionuclides that are created or remain in spent nuclear fuel after neutron irradiation and radioactive decay have occurred as a result of nuclear reactor operation. ORIGEN2 was chosen as the primary code for these calculations because it is widely used and accepted by the nuclear industry, both in the United States and the rest of the world. Its comprehensive library of over 1,600 nuclides includes any possible isotope of interest to the HEDR Project. It is important to evaluate the uncertainties expected from use of ORIGEN2 in the HEDR Project because these uncertainties may have a pivotal impact on the final accuracy and credibility of the results of the project. There are three primary sources of uncertainty in an ORIGEN2 calculation: basic nuclear data uncertainty in neutron cross sections, radioactive decay constants, energy per fission, and fission product yields; calculational uncertainty due to input data; and code uncertainties (i.e., numerical approximations, and neutron spectrum-averaged cross-section values from the code library). 15 refs., 5 figs., 5 tabs

TRING: a computer program for calculating radionuclide transport in groundwater

International Nuclear Information System (INIS)

Maul, P.R.

1984-12-01

The computer program TRING is described which enables the transport of radionuclides in groundwater to be calculated for use in long term radiological assessments using methods described previously. Examples of the areas of application of the program are activity transport in groundwater associated with accidental spillage or leakage of activity, the shutdown of reactors subject to delayed decommissioning, shallow land burial of intermediate level waste and geologic disposal of high level waste. Some examples of the use of the program are given, together with full details to enable users to run the program. (author)

On the computation of momentum distributions within wavepacket propagation calculations

International Nuclear Information System (INIS)

Feuerstein, Bernold; Thumm, Uwe

2003-01-01

We present a new method to extract momentum distributions from time-dependent wavepacket calculations. In contrast to established Fourier transformation of the spatial wavepacket at a fixed time, the proposed 'virtual detector' method examines the time dependence of the wavepacket at a fixed position. In first applications to the ionization of model atoms and the dissociation of H 2 + , we find a significant reduction of computing time and are able to extract reliable fragment momentum distributions by using a comparatively small spatial numerical grid for the time-dependent wavefunction

Neutron shielding point kernel integral calculation code for personal computer: PKN-pc

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Sakamoto, Yukio; Nakane, Yoshihiro; Tomita, Ken-ichi; Kurosawa, Naohiro.

1994-07-01

A personal computer version of PKN code, PKN-pc, has been developed to calculate neutron and secondary gamma-ray 1cm depth dose equivalents in water, ordinary concrete and iron for neutron source. Characteristics of PKN code are, to able to calculate dose equivalents in multi-layer three-dimensional system, which are described with two-dimensional surface, for monoenergetic neutron source from 0.01 to 14.9 MeV, 252 Cf fission and 241 Am-Be neutron source quick and easily. In addition to these features, the PKN-pc is possible to process interactive input and to get graphical system configuration and graphical results easily. (author)

Guide for licensing evaluations using CRAC2: A computer program for calculating reactor accident consequences

International Nuclear Information System (INIS)

White, J.E.; Roussin, R.W.; Gilpin, H.

1988-12-01

A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer system. Input preparation is facilitated through the use of an interactive computer program which operates on an IBM personal computer. The resulting CRAC2 input deck is transmitted to the MV/8000 by using an error-free file transfer mechanism. To facilitate the use of CRAC2 at NRC, relevant background material on input requirements and model descriptions has been extracted from four reports - ''Calculations of Reactor Accident Consequences,'' Version 2, NUREG/CR-2326 (SAND81-1994) and ''CRAC2 Model Descriptions,'' NUREG/CR-2552 (SAND82-0342), ''CRAC Calculations for Accident Sections of Environmental Statements, '' NUREG/CR-2901 (SAND82-1693), and ''Sensitivity and Uncertainty Studies of the CRAC2 Computer Code,'' NUREG/CR-4038 (ORNL-6114). When this background information is combined with instructions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific orientation toward applications on the MV/8000. 8 refs., 11 figs., 10 tabs

Impact on Dose Coefficients Calculated with ICRP Adult Mesh-type Reference Computational Phantoms

Energy Technology Data Exchange (ETDEWEB)

Yeom, Yeon Soo; Nguyen, Thang Tat; Choi, Chan Soo; Lee, Han Jin; Han, Hae Gin; Han, Min Cheol; Shin, Bang Ho; Kim, Chan Hyeong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

2017-04-15

In 2016, the International Commission on Radiological Protection (ICRP) formulated a new Task Group (TG) (i.e., TG 103) within Committee 2. The ultimate aim of the TG 103 is to develop the mesh-type reference computational phantoms (MRCPs) that can address dosimetric limitations of the currently used voxel-type reference computational phantoms (VRCPs) due to their limited voxel resolutions. The objective of the present study is to investigate dosimetric impact of the adult MRCPs by comparing dose coefficients (DCs) calculated with the MRCPs for some external and internal exposure cases and the reference DCs in ICRP Publications 116 and 133 that were produced with the adult VRCPs. In the present study, the DCs calculated with the adult MRCPs for some exposure cases were compared with the values in ICRP Publications 116 and 133. This comparison shows that in general the MRCPs provide very similar DCs for uncharged particles, but for charged particles provide significantly different DCs due to the improvement of the MRCPs.

Fast calculation method of computer-generated hologram using a depth camera with point cloud gridding

Science.gov (United States)

Zhao, Yu; Shi, Chen-Xiao; Kwon, Ki-Chul; Piao, Yan-Ling; Piao, Mei-Lan; Kim, Nam

2018-03-01

We propose a fast calculation method for a computer-generated hologram (CGH) of real objects that uses a point cloud gridding method. The depth information of the scene is acquired using a depth camera and the point cloud model is reconstructed virtually. Because each point of the point cloud is distributed precisely to the exact coordinates of each layer, each point of the point cloud can be classified into grids according to its depth. A diffraction calculation is performed on the grids using a fast Fourier transform (FFT) to obtain a CGH. The computational complexity is reduced dramatically in comparison with conventional methods. The feasibility of the proposed method was confirmed by numerical and optical experiments.

A computer program incorporating Pitzer's equations for calculation of geochemical reactions in brines

Science.gov (United States)

Plummer, Niel; Parkhurst, D.L.; Fleming, G.W.; Dunkle, S.A.

1988-01-01

The program named PHRQPITZ is a computer code capable of making geochemical calculations in brines and other electrolyte solutions to high concentrations using the Pitzer virial-coefficient approach for activity-coefficient corrections. Reaction-modeling capabilities include calculation of (1) aqueous speciation and mineral-saturation index, (2) mineral solubility, (3) mixing and titration of aqueous solutions, (4) irreversible reactions and mineral water mass transfer, and (5) reaction path. The computed results for each aqueous solution include the osmotic coefficient, water activity , mineral saturation indices, mean activity coefficients, total activity coefficients, and scale-dependent values of pH, individual-ion activities and individual-ion activity coeffients , and scale-dependent values of pH, individual-ion activities and individual-ion activity coefficients. A data base of Pitzer interaction parameters is provided at 25 C for the system: Na-K-Mg-Ca-H-Cl-SO4-OH-HCO3-CO3-CO2-H2O, and extended to include largely untested literature data for Fe(II), Mn(II), Sr, Ba, Li, and Br with provision for calculations at temperatures other than 25C. An extensive literature review of published Pitzer interaction parameters for many inorganic salts is given. Also described is an interactive input code for PHRQPITZ called PITZINPT. (USGS)

Dose calculation for iridium-192 sources by a personal computer

International Nuclear Information System (INIS)

Takahashi, Kenichi; Ishigaki, Hideyo; Udagawa, Kimio; Saito, Masami; Yamaguchi, Kyoko

1988-01-01

Recently Ir-192 sources have been used for interstitial radiotherapy instead of Ra-226 needles. One end of Ir-192 (single-pin) is formed with circlet and implanted Ir-192 sources are not always straight line. So the authors have developed a new dose calculation system, in which the authers employed conventional method considering oblique filteration for linear source and multi-point source method for curved source. Conventionally the positions of sources in three dimensions are determined from projections of the implanted sources on orthogonal or stereo radiographs. But it is frequentry impossible to define the end of sources on account of overlap. Then the authers have devised a method to determine the positions of sources from two radiographs which were taken with arbitrary directions. For tongue cancer injuries of mandibula so frequently occur after interstitial radiotherapy that the calculation of gingival dose is necessary. The positions of the gingival line are determined from two directional radiographs too. Further the three dimensional dose distributions can be displayed on the cathod ray tube. These calculations are performed by using a personal computer because of its distinctive features such as superiority in cost performance and flexibility for development and modification of programs. (author)

The utilization of Quabox/Cubox computer program for calculating Angra 1 Reactor core

International Nuclear Information System (INIS)

Pina, C.M. de.

1981-01-01

The utilization of Quabox/Cubox computer codes for calculating Angra 1 reactor core is studied. The results shows a dependency between the spent CPU time and the curacy of thermal power distribution in function of the polinomial expansion used. Comparison were mode between Citation code and some results from Westinghouse [pt

First principle calculations of effective exchange integrals: Comparison between SR (BS) and MR computational results

Energy Technology Data Exchange (ETDEWEB)

Yamaguchi, Kizashi [Institute for Nano Science Design Center, Osaka University, 1-3 Machikaneyama, Toyonaka, Osaka 560-8531, Japan and TOYOTA Physical and Chemical Research Institute, Nagakute, Aichi, 480-1192 (Japan); Nishihara, Satomichi; Saito, Toru; Yamanaka, Shusuke; Kitagawa, Yasutaka; Kawakami, Takashi; Yamada, Satoru; Isobe, Hiroshi; Okumura, Mitsutaka [Department of Chemistry, Graduate School of Science, Osaka University, 1-1 Machikaneyama, Toyonaka, Osaka 560-0043 (Japan)

2015-01-22

First principle calculations of effective exchange integrals (J) in the Heisenberg model for diradical species were performed by both symmetry-adapted (SA) multi-reference (MR) and broken-symmetry (BS) single reference (SR) methods. Mukherjee-type (Mk) state specific (SS) MR coupled-cluster (CC) calculations by the use of natural orbital (NO) references of ROHF, UHF, UDFT and CASSCF solutions were carried out to elucidate J values for di- and poly-radical species. Spin-unrestricted Hartree Fock (UHF) based coupled-cluster (CC) computations were also performed to these species. Comparison between UHF-NO(UNO)-MkMRCC and BS UHF-CC computational results indicated that spin-contamination of UHF-CC solutions still remains at the SD level. In order to eliminate the spin contamination, approximate spin-projection (AP) scheme was applied for UCC, and the AP procedure indeed corrected the error to yield good agreement with MkMRCC in energy. The CC double with spin-unrestricted Brueckner's orbital (UBD) was furthermore employed for these species, showing that spin-contamination involved in UHF solutions is largely suppressed, and therefore AP scheme for UBCCD removed easily the rest of spin-contamination. We also performed spin-unrestricted pure- and hybrid-density functional theory (UDFT) calculations of diradical and polyradical species. Three different computational schemes for total spin angular momentums were examined for the AP correction of the hybrid (H) UDFT. HUDFT calculations followed by AP, HUDFT(AP), yielded the S-T gaps that were qualitatively in good agreement with those of MkMRCCSD, UHF-CC(AP) and UB-CC(AP). Thus a systematic comparison among MkMRCCSD, UCC(AP) UBD(AP) and UDFT(AP) was performed concerning with the first principle calculations of J values in di- and poly-radical species. It was found that BS (AP) methods reproduce MkMRCCSD results, indicating their applicability to large exchange coupled systems.

THE COMPARISON BETWEEN COMPUTER SIMULATION AND PHYSICAL MODEL IN CALCULATING ILLUMINANCE LEVEL OF ATRIUM BUILDING

Directory of Open Access Journals (Sweden)

Sushardjanti Felasari

2003-01-01

Full Text Available This research examines the accuracy of computer programmes to simulate the illuminance level in atrium buildings compare to the measurement of those in physical models. The case was taken in atrium building with 4 types of roof i.e. pitched roof, barrel vault roof, monitor pitched roof (both monitor pitched roof and monitor barrel vault roof, and north light roof (both with north orientation and south orientation. The results show that both methods have agreement and disagreement. They show the same pattern of daylight distribution. In the other side, in terms of daylight factors, computer simulation tends to underestimate calculation compared to physical model measurement, while for average and minimum illumination, it tends to overestimate the calculation.

Wakefield calculations on parallel computers

International Nuclear Information System (INIS)

Schoessow, P.

1990-01-01

The use of parallelism in the solution of wakefield problems is illustrated for two different computer architectures (SIMD and MIMD). Results are given for finite difference codes which have been implemented on a Connection Machine and an Alliant FX/8 and which are used to compute wakefields in dielectric loaded structures. Benchmarks on code performance are presented for both cases. 4 refs., 3 figs., 2 tabs

A computer code for calculating neutron cross-sections from resonance parameter data

International Nuclear Information System (INIS)

Mill, A.J.

1979-08-01

A computer code, XSEC, has been written which calculates neutron cross-sections from resonance data. Although the program was originally written in order to identify neutron 'windows' in enriched nuclides, it may be used to evaluate the total neutron cross-section of any medium mass nuclide at intermediate energies. XSEC has proved very useful in identifying suitable nuclides for use as neutron filters at intermediate energies. (author)

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

International Nuclear Information System (INIS)

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

Computational models for probabilistic neutronic calculation in TADSEA

International Nuclear Information System (INIS)

Garcia, Jesus A.R.; Curbelo, Jesus P.; Hernandez, Carlos R.G.; Oliva, Amaury M.; Lira, Carlos A.B.O.

2013-01-01

The Very High Temperature Reactor is one of the main candidates for the next generation of nuclear power plants. In pebble bed reactors, the fuel is contained within graphite pebbles in the form of TRISO particles, which form a randomly packed bed inside a graphite-walled cylindrical cavity. In previous studies, the conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made. The TADSEA is a pebble-bed ADS cooled by helium and moderated by graphite. In order to simulate the TADSEA correctly, the double heterogeneity of the system must be considered. It consists on randomly located pebbles into the core and randomly located TRISO particles into the fuel pebbles. These features are often neglected due to the difficulty to model with MCNP code. The main reason is that there is a limited number of cells and surfaces to be defined. In this paper a computational tool, which allows to get a new geometrical model for fuel pebble to neutronic calculation with MCNPX, was presented. The heterogeneity of system is considered, and also the randomly located TRISO particles inside the pebble. There are also compared several neutronic computational models for TADSEA's fuel pebbles in order to study heterogeneity effects. On the other hand the boundary effect given by the intersection between the pebble surface and the TRISO particles could be significative in the multiplicative properties. A model to study this e ect is also presented. (author)

Mathematical model and computer programme for theoretical calculation of calibration curves of neutron soil moisture probes with highly effective counters

International Nuclear Information System (INIS)

Kolev, N.A.

1981-07-01

A mathematical model based on the three group theory for theoretical calculation by means of computer of the calibration curves of neutron soil moisture probes with highly effective counters, is described. Methods for experimental correction of the mathematical model are discussed and proposed. The computer programme described allows the calibration of neutron probes with high or low effective counters, and central or end geometry, with or without linearizing of the calibration curve. The use of two calculation variants and printing of output data gives the possibility not only for calibration, but also for other researches. The separate data inputs for soil and probe temperature allow the temperature influence analysis. The computer programme and calculation examples are given. (author)

Atmospheric dispersion calculation for posturated accident of nuclear facilities and the computer code: PANDA

International Nuclear Information System (INIS)

Kitahara, Yoshihisa; Kishimoto, Yoichiro; Narita, Osamu; Shinohara, Kunihiko

1979-01-01

Several Calculation methods for relative concentration (X/Q) and relative cloud-gamma dose (D/Q) of the radioactive materials released from nuclear facilities by posturated accident are presented. The procedure has been formulated as a Computer program PANDA and the usage is explained. (author)

ICECON: a computer program used to calculate containment back pressure for LOCA analysis (including ice condenser plants)

International Nuclear Information System (INIS)

1976-07-01

The ICECON computer code provides a method for conservatively calculating the long term back pressure transient in the containment resulting from a hypothetical Loss-of-Coolant Accident (LOCA) for PWR plants including ice condenser containment systems. The ICECON computer code was developed from the CONTEMPT/LT-022 code. A brief discussion of the salient features of a typical ice condenser containment is presented. Details of the ice condenser models are explained. The corrections and improvements made to CONTEMPT/LT-022 are included. The organization of the code, including the calculational procedure, is outlined. The user's manual, to be used in conjunction with the CONTEMPT/LT-022 user's manual, a sample problem, a time-step study (solution convergence) and a comparison of ICECON results with the results of the NSSS vendor are presented. In general, containment pressure calculated with the ICECON code agree with those calculated by the NSSS vendor using the same mass and energy release rates to the containment

A computer program (COSTUM) to calculate confidence intervals for in situ stress measurements. V. 1

International Nuclear Information System (INIS)

Dzik, E.J.; Walker, J.R.; Martin, C.D.

1989-03-01

The state of in situ stress is one of the parameters required both for the design and analysis of underground excavations and for the evaluation of numerical models used to simulate underground conditions. To account for the variability and uncertainty of in situ stress measurements, it is desirable to apply confidence limits to measured stresses. Several measurements of the state of stress along a borehole are often made to estimate the average state of stress at a point. Since stress is a tensor, calculating the mean stress and confidence limits using scalar techniques is inappropriate as well as incorrect. A computer program has been written to calculate and present the mean principle stresses and the confidence limits for the magnitudes and directions of the mean principle stresses. This report describes the computer program, COSTUM

A benchmark test of computer codes for calculating average resonance parameters

International Nuclear Information System (INIS)

Ribon, P.; Thompson, A.

1983-01-01

A set of resonance parameters has been generated from known, but secret, average values; the parameters have then been adjusted to mimic experimental data by including the effects of Doppler broadening, resolution broadening and statistical fluctuations. Average parameters calculated from the dataset by various computer codes are compared with each other, and also with the true values. The benchmark test is fully described in the report NEANDC160-U (NEA Data Bank Newsletter No. 27 July 1982); the present paper is a summary of this document. (Auth.)

Burnup calculations for KIPT accelerator driven subcritical facility using Monte Carlo computer codes-MCB and MCNPX

International Nuclear Information System (INIS)

Gohar, Y.; Zhong, Z.; Talamo, A.

2009-01-01

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is ∼375 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the electrons and the

A finite element computer program for the calculation of the resonant frequencies of anisotropic materials

International Nuclear Information System (INIS)

Fleury, W.H.; Rosinger, H.E.; Ritchie, I.G.

1975-09-01

A set of computer programs for the calculation of the flexural and torsional resonant frequencies of rectangular section bars of materials of orthotropic or higher symmetry are described. The calculations are used in the experimental determination and verification of the elastic constants of anisotropic materials. The simple finite element technique employed separates the inertial and elastic properties of the beam element into station and field transfer matrices respectively. It includes the Timoshenko beam corrections for flexure and Lekhnitskii's theory for torsion-flexure coupling. The programs also calculate the vibration shapes and surface nodal contours or Chladni figures of the vibration modes. (author)

Calculation of radiation losses in cylinder symmetric high pressure discharges by means of a digital computer

International Nuclear Information System (INIS)

Andriessen, F.J.; Boerman, W.; Holtz, I.F.E.M.

1973-08-01

Computer calculations have been made of radiative energy losses in a cylindrically symmetric high pressure discharge. The calculations show that the radiation losses which occur in discharges at pressures of a few atmospheres and central temperatures of about 20000degK when compared with the electrical energy supplied, are only of importance in the neighbourhood of the centre of discharge

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)

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

Computing Moment-Based Probability Tables for Self-Shielding Calculations in Lattice Codes

International Nuclear Information System (INIS)

Hebert, Alain; Coste, Mireille

2002-01-01

As part of the self-shielding model used in the APOLLO2 lattice code, probability tables are required to compute self-shielded cross sections for coarse energy groups (typically with 99 or 172 groups). This paper describes the replacement of the multiband tables (typically with 51 subgroups) with moment-based tables in release 2.5 of APOLLO2. An improved Ribon method is proposed to compute moment-based probability tables, allowing important savings in CPU resources while maintaining the accuracy of the self-shielding algorithm. Finally, a validation is presented where the absorption rates obtained with each of these techniques are compared with exact values obtained using a fine-group elastic slowing-down calculation in the resolved energy domain. Other results, relative to the Rowland's benchmark and to three assembly production cases, are also presented

Calculation of shipboard fire conditions for radioactive materials packages with the methods of computational fluid dynamics

International Nuclear Information System (INIS)

Koski, J.A.; Wix, S.D.; Cole, J.K.

1997-09-01

Shipboard fires both in the same ship hold and in an adjacent hold aboard a break-bulk cargo ship are simulated with a commercial finite-volume computational fluid mechanics code. The fire models and modeling techniques are described and discussed. Temperatures and heat fluxes to a simulated materials package are calculated and compared to experimental values. The overall accuracy of the calculations is assessed

Using the calculational simulating complexes when making the computer process control systems for NPP

International Nuclear Information System (INIS)

Zimakov, V.N.; Chernykh, V.P.

1998-01-01

The problems on creating calculational-simulating (CSC) and their application by developing the program and program-technical means for computer-aided process control systems at NPP are considered. The abo- ve complex is based on the all-mode real time mathematical model, functioning at a special complex of computerized means

Internal radiation dose calculations with the INREM II computer code

International Nuclear Information System (INIS)

Dunning, D.E. Jr.; Killough, G.G.

1978-01-01

A computer code, INREM II, was developed to calculate the internal radiation dose equivalent to organs of man which results from the intake of a radionuclide by inhalation or ingestion. Deposition and removal of radioactivity from the respiratory tract is represented by the Internal Commission on Radiological Protection Task Group Lung Model. A four-segment catenary model of the gastrointestinal tract is used to estimate movement of radioactive material that is ingested, or swallowed after being cleared from the respiratory tract. Retention of radioactivity in other organs is specified by linear combinations of decaying exponential functions. The formation and decay of radioactive daughters is treated explicitly, with each radionuclide in the decay chain having its own uptake and retention parameters, as supplied by the user. The dose equivalent to a target organ is computed as the sum of contributions from each source organ in which radioactivity is assumed to be situated. This calculation utilizes a matrix of dosimetric S-factors (rem/μCi-day) supplied by the user for the particular choice of source and target organs. Output permits the evaluation of components of dose from cross-irradiations when penetrating radiations are present. INREM II has been utilized with current radioactive decay data and metabolic models to produce extensive tabulations of dose conversion factors for a reference adult for approximately 150 radionuclides of interest in environmental assessments of light-water-reactor fuel cycles. These dose conversion factors represent the 50-year dose commitment per microcurie intake of a given radionuclide for 22target organs including contributions from specified source organs and surplus activity in the rest of the body. These tabulations are particularly significant in their consistent use of contemporary models and data and in the detail of documentation

Elasto-plastic benchmark calculations. Step 1: verification of the numerical accuracy of the computer programs

International Nuclear Information System (INIS)

Corsi, F.

1985-01-01

In connection with the design of nuclear reactors components operating at elevated temperature, design criteria need a level of realism in the prediction of inelastic structural behaviour. This concept leads to the necessity of developing non linear computer programmes, and, as a consequence, to the problems of verification and qualification of these tools. Benchmark calculations allow to carry out these two actions, involving at the same time an increased level of confidence in complex phenomena analysis and in inelastic design calculations. With the financial and programmatic support of the Commission of the European Communities (CEE) a programme of elasto-plastic benchmark calculations relevant to the design of structural components for LMFBR has been undertaken by those Member States which are developing a fast reactor project. Four principal progressive aims were initially pointed out that brought to the decision to subdivide the Benchmark effort in a calculations series of four sequential steps: step 1 to 4. The present document tries to summarize Step 1 of the Benchmark exercise, to derive some conclusions on Step 1 by comparison of the results obtained with the various codes and to point out some concluding comments on the first action. It is to point out that even if the work was designed to test the capabilities of the computer codes, another aim was to increase the skill of the users concerned

DCHAIN: A user-friendly computer program for radioactive decay and reaction chain calculations

International Nuclear Information System (INIS)

East, L.V.

1994-05-01

A computer program for calculating the time-dependent daughter populations in radioactive decay and nuclear reaction chains is described. Chain members can have non-zero initial populations and be produced from the preceding chain member as the result of radioactive decay, a nuclear reaction, or both. As presently implemented, chains can contain up to 15 members. Program input can be supplied interactively or read from ASCII data files. Time units for half-lives, etc. can be specified during data entry. Input values are verified and can be modified if necessary, before used in calculations. Output results can be saved in ASCII files in a format suitable for including in reports or other documents. The calculational method, described in some detail, utilizes a generalized form of the Bateman equations. The program is written in the C language in conformance with current ANSI standards and can be used on multiple hardware platforms

Bessel function expansion to reduce the calculation time and memory usage for cylindrical computer-generated holograms.

Science.gov (United States)

Sando, Yusuke; Barada, Daisuke; Jackin, Boaz Jessie; Yatagai, Toyohiko

2017-07-10

This study proposes a method to reduce the calculation time and memory usage required for calculating cylindrical computer-generated holograms. The wavefront on the cylindrical observation surface is represented as a convolution integral in the 3D Fourier domain. The Fourier transformation of the kernel function involving this convolution integral is analytically performed using a Bessel function expansion. The analytical solution can drastically reduce the calculation time and the memory usage without any cost, compared with the numerical method using fast Fourier transform to Fourier transform the kernel function. In this study, we present the analytical derivation, the efficient calculation of Bessel function series, and a numerical simulation. Furthermore, we demonstrate the effectiveness of the analytical solution through comparisons of calculation time and memory usage.

Computer program TMOC for calculating of pressure transients in fluid filled piping networks

International Nuclear Information System (INIS)

Siikonen, T.

1978-01-01

The propagation of a pressure wave in fluid filles tubes is significantly affected by the pipe wall motion and vice versa. A computer code TMOC (Transients by the Method of Characteristics) is being developed for the analysis of the coupled fluid and pipe wall transients. Because of the structural feedback, the pressure can be calculated more accurately than in the programs commonly used. (author)

A compilation of structural property data for computer impact calculation (5/5)

International Nuclear Information System (INIS)

Ikushima, Takeshi

1988-10-01

The paper describes structural property data for computer impact calculations of nuclear fuel shipping casks. Four kinds of material data, mild steel, stainless steel, lead and wood are compiled. These materials are main structural elements of shipping casks. Structural data such as, the coefficient of thermal expansion, the modulus of longitudinal elasticity, the modulus of transverse elasticity, the Poisson's ratio and stress and strain relationships, have been tabulated against temperature or strain rate. This volume 5 involve structural property data of wood. (author)

Calculations detailed progression of fire in NPP ALMARAZ through the code computational fire dynamics SIMULATOR

International Nuclear Information System (INIS)

Villar Sanchez, T.

2012-01-01

(FDS) is an advanced computational model of calculation of simulation of fire that numerically solves the Navier-Stokes equations in each cell of the mesh in each interval of time, having capacity to calculate accurately all those parameters of fire to NUREG-1805 has a limited capacity. The objective of the analysis is to compare the results obtained with the FDS with those obtained from spreadsheets of NUREG-1805 and deal widespread and realistic study of the propagation of a fire in different areas of NPP Almaraz.

POPFOOD - a computer code for calculating ingestion collective doses from continuous atmospheric releases

International Nuclear Information System (INIS)

Hotson, J.; Stacey, A.; Nair, S.

1980-07-01

The basic methodology incorporated in the POPFOOD computer code is described, which may be used to calculate equilibrium collective dose rates associated with continuous atmospheric releases and arising from consumption of a broad range of food products. The standard data libraries associated with the code are also described. These include a data library, based on the 1972 agricultural census, describing the spatial distribution of production, in England, Wales and Scotland, of the following food products: milk; beef and veal; pork bacon and ham; poultrymeat; eggs; mutton and lamb; root vegetables; green vegetables; fruit; cereals. Illustrative collective dose calculations were made for the case of 1 Ci per year emissions of 131 I, tritium and 14 C from a typical rural UK site. The calculations indicate that the ingestion pathway results in a greater collective dose than that via inhalation, with the contributions from consumption of root and green vegetables, and cereals being of comparable significance to that from liquid milk consumption, in all three cases. (author)

Computing tools for accelerator design calculations

International Nuclear Information System (INIS)

Fischler, M.; Nash, T.

1984-01-01

This note is intended as a brief, summary guide for accelerator designers to the new generation of commercial and special processors that allow great increases in computing cost effectiveness. New thinking is required to take best advantage of these computing opportunities, in particular, when moving from analytical approaches to tracking simulations. In this paper, we outline the relevant considerations

A compilation of structural property data for computer impact calculation (3/5)

International Nuclear Information System (INIS)

Ikushima, Takeshi

1988-10-01

The paper describes structural property data for computer impact calculations of nuclear fuel shipping casks. Four kinds of material data, mild steel, stainless steel, lead and wood are compiled. These materials are main structural elements of shipping casks. Structural data such as, the coefficient of thermal expansion, the modulus of longitudinal elasticity, the modulus of transverse elasticity, the Poisson's ratio and stress and strain relationships, have been tabulated against temperature or strain rate. This volume 3 involve structural property data of stainless steel. (author)

A compilation of structural property data for computer impact calculation (2/5)

International Nuclear Information System (INIS)

Ikushima, Takeshi

1988-10-01

The paper describes structural property data for computer impact calculations of nuclear fuel shipping casks. Four kinds of material data, mild steel, stainless steel, lead and wood are compiled. These materials are main structural elements of shipping casks. Structural data such as, the coefficient of thermal expansion, the modulus of longitudinal elasticity, the modulus of transverse elasticity, the Poisson's ratio and stress and strain relationships, have been tabulated against temperature or strain rate. This volume 2 involve structural property data of mild steel. (author)

Modeling of tube current modulation methods in computed tomography dose calculations for adult and pregnant patients

International Nuclear Information System (INIS)

Caracappa, Peter F.; Xu, X. George; Gu, Jianwei

2011-01-01

The comparatively high dose and increasing frequency of computed tomography (CT) examinations have spurred the development of techniques for reducing radiation dose to imaging patients. Among these is the application of tube current modulation (TCM), which can be applied either longitudinally along the body or rotationally along the body, or both. Existing computational models for calculating dose from CT examinations do not include TCM techniques. Dose calculations using Monte Carlo methods have been previously prepared for constant-current rotational exposures at various positions along the body and for the principle exposure projections for several sets of computational phantoms, including adult male and female and pregnant patients. Dose calculations from CT scans with TCM are prepared by appropriately weighting the existing dose data. Longitudinal TCM doses can be obtained by weighting the dose at the z-axis scan position by the relative tube current at that position. Rotational TCM doses are weighted using the relative organ doses from the principle projections as a function of the current at the rotational angle. Significant dose reductions of 15% to 25% to fetal tissues are found from simulations of longitudinal TCM schemes to pregnant patients of different gestational ages. Weighting factors for each organ in rotational TCM schemes applied to adult male and female patients have also been found. As the application of TCM techniques becomes more prevalent, the need for including TCM in CT dose estimates will necessarily increase. (author)

Calculating external doses from contaminated soil with the computer model SOILD

International Nuclear Information System (INIS)

Chen, Y.; LePoire, D.; Yu, C.

1991-01-01

The SOILD computer model was developed for calculating the effective dose equivalent from external exposure to distributed gamma sources in soil. It is designed to assess external doses under various exposure scenarios that may be encountered in environmental restoration programs. The model's four major functional features address (a) dose versus source depth in soil, (b) shielding of clean cover soil, (c) area of contamination, and (d) nonuniform distribution of sources. The model can also adjust doses when there are variations in soil densities for both source and cover soils. It is supported by a data base of ∼500 radionuclides. A sample calculation was performed by SOILD to determine the effective dose equivalent for a uniform source distribution in soil. The soil density was assumed to be 1.6 g/cm 3 , and the source strength was assumed to be 1 pCi/cm 3 . The following radionuclides were studied: 60 C, 131 I, 137+D Cs, 238+D U, and 226+D Ra ('+D' denotes the parent nuclide and daughters)

Convergence acceleration of two-phase flow calculations in FLICA-4. A thermal-hydraulic 3D computer code

International Nuclear Information System (INIS)

Toumi, I.

1995-01-01

Time requirements for 3D two-phase flow steady state calculations are generally long. Usually, numerical methods for steady state problems are iterative methods consisting in time-like methods that are marched to a steady state. Based on the eigenvalue spectrum of the iteration matrix for various flow configuration, two convergence acceleration techniques are discussed; over-relaxation and eigenvalue annihilation. This methods were applied to accelerate the convergence of three dimensional steady state two-phase flow calculations within the FLICA-4 computer code. These acceleration methods are easy to implement and no extra computer memory is required. Successful results are presented for various test problems and a saving of 30 to 50 % in CPU time have been achieved. (author). 10 refs., 4 figs

Computer program for the sensitivity calculation of a CR-39 detector in a diffusion chamber for radon measurements

Energy Technology Data Exchange (ETDEWEB)

Nikezic, D., E-mail: nikezic@kg.ac.rs; Stajic, J. M. [Faculty of Science, University of Kragujevac, R. Domanovica 12, Kragujevac 34000 (Serbia); Yu, K. N. [Department of Physics and Materials Science, City University of Hong Kong, 83 Tat Chee Avenue (Hong Kong)

2014-02-15

Computer software for calculation of the sensitivity of a CR-39 detector closed in a diffusion chamber to radon is described in this work. The software consists of two programs, both written in the standard Fortran 90 programming language. The physical background and a numerical example are given. Presented software is intended for numerous researches in radon measurement community. Previously published computer programs TRACK-TEST.F90 and TRACK-VISION.F90 [D. Nikezic and K. N. Yu, Comput. Phys. Commun. 174, 160 (2006); D. Nikezic and K. N. Yu, Comput. Phys. Commun. 178, 591 (2008)] are used here as subroutines to calculate the track parameters and to determine whether the track is visible or not, based on the incident angle, impact energy, etching conditions, gray level, and visibility criterion. The results obtained by the software, using five different V functions, were compared with the experimental data found in the literature. Application of two functions in this software reproduced experimental data very well, while other three gave lower sensitivity than experiment.

Monteray Mark-I: Computer program (PC-version) for shielding calculation with Monte Carlo method

International Nuclear Information System (INIS)

Pudjijanto, M.S.; Akhmad, Y.R.

1998-01-01

A computer program for gamma ray shielding calculation using Monte Carlo method has been developed. The program is written in WATFOR77 language. The MONTERAY MARH-1 is originally developed by James Wood. The program was modified by the authors that the modified version is easily executed. Applying Monte Carlo method the program observe photon gamma transport in an infinity planar shielding with various thick. A photon gamma is observed till escape from the shielding or when its energy less than the cut off energy. Pair production process is treated as pure absorption process that annihilation photons generated in the process are neglected in the calculation. The out put data calculated by the program are total albedo, build-up factor, and photon spectra. The calculation result for build-up factor of a slab lead and water media with 6 MeV parallel beam gamma source shows that they are in agreement with published data. Hence the program is adequate as a shielding design tool for observing gamma radiation transport in various media

Accuracy of computer-calculated and manual QRS duration assessments: Clinical implications to select candidates for cardiac resynchronization therapy.

Science.gov (United States)

De Pooter, Jan; El Haddad, Milad; Stroobandt, Roland; De Buyzere, Marc; Timmermans, Frank

2017-06-01

QRS duration (QRSD) plays a key role in the field of cardiac resynchronization therapy (CRT). Computer-calculated QRSD assessments are widely used, however inter-manufacturer differences have not been investigated in CRT candidates. QRSD was assessed in 377 digitally stored ECGs: 139 narrow QRS, 140 LBBB and 98 ventricular paced ECGs. Manual QRSD was measured as global QRSD, using digital calipers, by two independent observers. Computer-calculated QRSD was assessed by Marquette 12SL (GE Healthcare, Waukesha, WI, USA) and SEMA3 (Schiller, Baar, Switzerland). Inter-manufacturer differences of computer-calculated QRSD assessments vary among different QRS morphologies: narrow QRSD: 4 [2-9] ms (median [IQR]), p=0.010; LBBB QRSD: 7 [2-10] ms, p=0.003 and paced QRSD: 13 [6-18] ms, p=0.007. Interobserver differences of manual QRSD assessments measured: narrow QRSD: 4 [2-6] ms, p=non-significant; LBBB QRSD: 6 [3-12] ms, p=0.006; paced QRSD: 8 [4-18] ms, p=0.001. In LBBB ECGs, intraclass correlation coefficients (ICCs) were comparable for inter-manufacturer and interobserver agreement (ICC 0.830 versus 0.837). When assessing paced QRSD, manual measurements showed higher ICC compared to inter-manufacturer agreement (ICC 0.902 versus 0.776). Using guideline cutoffs of 130ms, up to 15% of the LBBB ECGs would be misclassified as <130ms or ≥130ms by at least one method. Using a cutoff of 150ms, this number increases to 33% of ECGs being misclassified. However, by combining LBBB-morphology and QRSD, the number of misclassified ECGs can be decreased by half. Inter-manufacturer differences in computer-calculated QRSD assessments are significant and may compromise adequate selection of individual CRT candidates when using QRSD as sole parameter. Paced QRSD should preferentially be assessed by manual QRSD measurements. Copyright © 2017 Elsevier B.V. All rights reserved.

RAP-3A Computer code for thermal and hydraulic calculations in steady state conditions for fuel element clusters

International Nuclear Information System (INIS)

Popescu, C.; Biro, L.; Iftode, I.; Turcu, I.

1975-10-01

The RAP-3A computer code is designed for calculating the main steady state thermo-hydraulic parameters of multirod fuel clusters with liquid metal cooling. The programme provides a double accuracy computation of temperatures and axial enthalpy distributions of pressure losses and axial heat flux distributions in fuel clusters before boiling conditions occur. Physical and mathematical models as well as a sample problem are presented. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer

A computer code 'BEAM' for the ion optics calculation of the JAERI tandem accelerator system

International Nuclear Information System (INIS)

Kikuchi, Shiroh; Takeuchi, Suehiro

1987-11-01

The computer code BEAM is described, together with an outline of the formalism used for the ion optics calculation. The purpose of the code is to obtain the optimum parameters of devices, with which the ion beam is transported through the system without losses. The procedures of the calculation, especially those of searching for the parameters of quadrupole lenses, are discussed in detail. The flow of the code is illustrated as a whole and its constituent subroutines are explained individually. A few resultant beam trajectories and the parameters used to obtain them are shown as examples. (author)

Development of a computational code for calculations of shielding in dental facilities

International Nuclear Information System (INIS)

Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L.

2014-01-01

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

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.

Comparison of image features calculated in different dimensions for computer-aided diagnosis of lung nodules

Science.gov (United States)

Xu, Ye; Lee, Michael C.; Boroczky, Lilla; Cann, Aaron D.; Borczuk, Alain C.; Kawut, Steven M.; Powell, Charles A.

2009-02-01

Features calculated from different dimensions of images capture quantitative information of the lung nodules through one or multiple image slices. Previously published computer-aided diagnosis (CADx) systems have used either twodimensional (2D) or three-dimensional (3D) features, though there has been little systematic analysis of the relevance of the different dimensions and of the impact of combining different dimensions. The aim of this study is to determine the importance of combining features calculated in different dimensions. We have performed CADx experiments on 125 pulmonary nodules imaged using multi-detector row CT (MDCT). The CADx system computed 192 2D, 2.5D, and 3D image features of the lesions. Leave-one-out experiments were performed using five different combinations of features from different dimensions: 2D, 3D, 2.5D, 2D+3D, and 2D+3D+2.5D. The experiments were performed ten times for each group. Accuracy, sensitivity and specificity were used to evaluate the performance. Wilcoxon signed-rank tests were applied to compare the classification results from these five different combinations of features. Our results showed that 3D image features generate the best result compared with other combinations of features. This suggests one approach to potentially reducing the dimensionality of the CADx data space and the computational complexity of the system while maintaining diagnostic accuracy.

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.

A special purpose computer for the calculation of the electric conductivity of a random resistor network

International Nuclear Information System (INIS)

Hajjar, Mansour

1987-01-01

The special purpose computer PERCOLA is designed for long numerical simulations on a percolation problem in Statistical Mechanics of disordered media. Our aim is to improve the actual values of the critical exponents characterizing the behaviour of random resistance networks at percolation threshold. The architecture of PERCOLA is based on an efficient iterative algorithm used to compute the electric conductivity of such networks. The calculator has the characteristics of a general purpose 64 bits floating point micro-programmable computer that can run programs for various types of problems with a peak performance of 25 Mflops. This high computing speed is a result of the pipeline architecture based on internal parallelism and separately micro-code controlled units such as: data memories, a micro-code memory, ALUs and multipliers (both WEITEK components), various data paths, a sequencer (ANALOG DEVICES component), address generators and a random number generator. Thus, the special purpose computer runs percolation problem program 10 percent faster than the supercomputer CRAY XMP. (author) [fr

Computational scheme for pH-dependent binding free energy calculation with explicit solvent.

Science.gov (United States)

Lee, Juyong; Miller, Benjamin T; Brooks, Bernard R

2016-01-01

We present a computational scheme to compute the pH-dependence of binding free energy with explicit solvent. Despite the importance of pH, the effect of pH has been generally neglected in binding free energy calculations because of a lack of accurate methods to model it. To address this limitation, we use a constant-pH methodology to obtain a true ensemble of multiple protonation states of a titratable system at a given pH and analyze the ensemble using the Bennett acceptance ratio (BAR) method. The constant pH method is based on the combination of enveloping distribution sampling (EDS) with the Hamiltonian replica exchange method (HREM), which yields an accurate semi-grand canonical ensemble of a titratable system. By considering the free energy change of constraining multiple protonation states to a single state or releasing a single protonation state to multiple states, the pH dependent binding free energy profile can be obtained. We perform benchmark simulations of a host-guest system: cucurbit[7]uril (CB[7]) and benzimidazole (BZ). BZ experiences a large pKa shift upon complex formation. The pH-dependent binding free energy profiles of the benchmark system are obtained with three different long-range interaction calculation schemes: a cutoff, the particle mesh Ewald (PME), and the isotropic periodic sum (IPS) method. Our scheme captures the pH-dependent behavior of binding free energy successfully. Absolute binding free energy values obtained with the PME and IPS methods are consistent, while cutoff method results are off by 2 kcal mol(-1) . We also discuss the characteristics of three long-range interaction calculation methods for constant-pH simulations. © 2015 The Protein Society.

Radiation damage calculations for compound materials

International Nuclear Information System (INIS)

Greenwood, L.R.

1990-01-01

This paper reports on the SPECOMP computer code, developed to calculate neutron-induced displacement damage cross sections for compound materials such as alloys, insulators, and ceramic tritium breeders for fusion reactors. These new calculations rely on recoil atom energy distributions previously computed with the DISCS computer code, the results of which are stored in SPECTER computer code master libraries. All reaction channels were considered in the DISCS calculations and the neutron cross sections were taken from ENDF/B-V. Compound damage calculations with SPECOMP thus do not need to perform any recoil atom calculations and consequently need no access to ENDF or other neutron data bases. The calculations proceed by determining secondary displacements for each combination of recoil atom and matrix atom using the Lindhard partition of the recoil energy to establish the damage energy

Preparation of functions of computer code GENGTC and improvement for two-dimensional heat transfer calculations for irradiation capsules

International Nuclear Information System (INIS)

Nomura, Yasushi; Someya, Hiroyuki; Ito, Haruhiko.

1992-11-01

Capsules for irradiation tests in the JMTR (Japan Materials Testing Reactor), consist of irradiation specimens surrounded by a cladding tube, holders, an inner tube and a container tube (from 30mm to 65mm in diameter). And the annular gaps between these structural materials in the capsule are filled with liquids or gases. Cooling of the capsule is done by reactor primary coolant flowing down outside the capsule. Most of the heat generated by fission in fuel specimens and gamma absorption in structural materials is directed radially to the capsule container outer surface. In thermal performance calculations for capsule design, an one(r)-dimensional heat transfer computer code entitled (Generalyzed Gap Temperature Calculation), GENGTC, originally developed in Oak Ridge National Laboratory, U.S.A., has been frequently used. In designing a capsule, are needed many cases of parametric calculations with respect to changes materials and gap sizes. And in some cases, two(r,z)-dimensional heat transfer calculations are needed for irradiation test capsules with short length fuel rods. Recently the authors improved the original one-dimensional code GENGTC, (1) to simplify preparation of input data, (2) to perform automatic calculations for parametric survey based on design temperatures, ect. Moreover, the computer code has been improved to perform r-z two-dimensional heat transfer calculation. This report describes contents of the preparation of the one-dimensional code GENGTC and the improvement for the two-dimensional code GENGTC-2, together with their code manuals. (author)

Computer programs for lattice calculations

International Nuclear Information System (INIS)

Keil, E.; Reich, K.H.

1984-01-01

The aim of the workshop was to find out whether some standardisation could be achieved for future work in this field. A certain amount of useful information was unearthed, and desirable features of a ''standard'' program emerged. Progress is not expected to be breathtaking, although participants (practically from all interested US, Canadian and European accelerator laboratories) agreed that the mathematics of the existing programs is more or less the same. Apart from the NIH (not invented here) effect, there is a - to quite some extent understandable - tendency to stay with a program one knows and to add to it if unavoidable rather than to start using a new one. Users of the well supported program TRANSPORT (designed for beam line calculations) would prefer to have it fully extended for lattice calculations (to some extent already possible now), while SYNCH users wish to see that program provided with a user-friendly input, rather than spending time and effort for mastering a new program

A new shielding calculation method for X-ray computed tomography regarding scattered radiation.

Science.gov (United States)

Watanabe, Hiroshi; Noto, Kimiya; Shohji, Tomokazu; Ogawa, Yasuyoshi; Fujibuchi, Toshioh; Yamaguchi, Ichiro; Hiraki, Hitoshi; Kida, Tetsuo; Sasanuma, Kazutoshi; Katsunuma, Yasushi; Nakano, Takurou; Horitsugi, Genki; Hosono, Makoto

2017-06-01

The goal of this study is to develop a more appropriate shielding calculation method for computed tomography (CT) in comparison with the Japanese conventional (JC) method and the National Council on Radiation Protection and Measurements (NCRP)-dose length product (DLP) method. Scattered dose distributions were measured in a CT room with 18 scanners (16 scanners in the case of the JC method) for one week during routine clinical use. The radiation doses were calculated for the same period using the JC and NCRP-DLP methods. The mean (NCRP-DLP-calculated dose)/(measured dose) ratios in each direction ranged from 1.7 ± 0.6 to 55 ± 24 (mean ± standard deviation). The NCRP-DLP method underestimated the dose at 3.4% in fewer shielding directions without the gantry and a subject, and the minimum (NCRP-DLP-calculated dose)/(measured dose) ratio was 0.6. The reduction factors were 0.036 ± 0.014 and 0.24 ± 0.061 for the gantry and couch directions, respectively. The (JC-calculated dose)/(measured dose) ratios ranged from 11 ± 8.7 to 404 ± 340. The air kerma scatter factor κ is expected to be twice as high as that calculated with the NCRP-DLP method and the reduction factors are expected to be 0.1 and 0.4 for the gantry and couch directions, respectively. We, therefore, propose a more appropriate method, the Japanese-DLP method, which resolves the issues of possible underestimation of the scattered radiation and overestimation of the reduction factors in the gantry and couch directions.

Radcalc: A computer program to calculate the radiolytic production of hydrogen gas from radioactive wastes in packages

International Nuclear Information System (INIS)

Green, J.R.; Schwarz, R.A.; Hillesland, K.E.; Roetman, V.E.; Field, J.G.

1995-11-01

Radcalc for Windows' is a menu-driven Microsoft2 Windows-compatible computer code that calculates the radiolytic production of hydrogen gas in high- and low-level radioactive waste. In addition, the code also determines US Department of Transportation (DOT) transportation classifications, calculates the activities of parent and daughter isotopes for a specified period of time, calculates decay heat, and calculates pressure buildup from the production of hydrogen gas in a given package geometry. Radcalc for Windows was developed by Packaging Engineering, Transportation and Packaging, Westinghouse Hanford Company, Richland, Washington, for the US Department of Energy (DOE). It is available from Packaging Engineering and is issued with a user's manual and a technical manual. The code has been verified and validated

LALAGE - a computer program to calculate the TM01 modes of cylindrically symmetrical multicell resonant structures

International Nuclear Information System (INIS)

Fernandes, P.

1982-01-01

An improvement has been made to the LALA program to compute resonant frequencies and fields for all the modes of the lowest TM 01 band-pass of multicell structures. The results are compared with those calculated by another popular rf cavity code and with experimentally measured quantities. (author)

Computational Benchmark Calculations Relevant to the Neutronic Design of the Spallation Neutron Source (SNS)

International Nuclear Information System (INIS)

Gallmeier, F.X.; Glasgow, D.C.; Jerde, E.A.; Johnson, J.O.; Yugo, J.J.

1999-01-01

The Spallation Neutron Source (SNS) will provide an intense source of low-energy neutrons for experimental use. The low-energy neutrons are produced by the interaction of a high-energy (1.0 GeV) proton beam on a mercury (Hg) target and slowed down in liquid hydrogen or light water moderators. Computer codes and computational techniques are being benchmarked against relevant experimental data to validate and verify the tools being used to predict the performance of the SNS. The LAHET Code System (LCS), which includes LAHET, HTAPE ad HMCNP (a modified version of MCNP version 3b), have been applied to the analysis of experiments that were conducted in the Alternating Gradient Synchrotron (AGS) facility at Brookhaven National Laboratory (BNL). In the AGS experiments, foils of various materials were placed around a mercury-filled stainless steel cylinder, which was bombarded with protons at 1.6 GeV. Neutrons created in the mercury target, activated the foils. Activities of the relevant isotopes were accurately measured and compared with calculated predictions. Measurements at BNL were provided in part by collaborating scientists from JAERI as part of the AGS Spallation Target Experiment (ASTE) collaboration. To date, calculations have shown good agreement with measurements

The determination of surface of powders by BET method using nitrogen and krypton with computer calculation of the results

International Nuclear Information System (INIS)

Dembinski, W.; Zlotowski, T.

1973-01-01

A computer program written in FORTRAN language for calculations of final results of specific surface analysis based on BET theory has been described. Two gases - nitrogen and krypton were used. A technical description of measuring apparaturs is presented as well as theoretical basis of the calculations together with statistical analysis of the results for uranium compounds powders. (author)

Computer program for calculation of complex chemical equilibrium compositions and applications. Supplement 1: Transport properties

Science.gov (United States)

Gordon, S.; Mcbride, B.; Zeleznik, F. J.

1984-01-01

An addition to the computer program of NASA SP-273 is given that permits transport property calculations for the gaseous phase. Approximate mixture formulas are used to obtain viscosity and frozen thermal conductivity. Reaction thermal conductivity is obtained by the same method as in NASA TN D-7056. Transport properties for 154 gaseous species were selected for use with the program.

UNIDOSE - a computer program for the calculation of individual and collective doses from airborne radioactive pollutants

International Nuclear Information System (INIS)

Karlberg, O.; Schwartz, H.; Forssen, B.-H.; Marklund, J.-E.

1979-01-01

UNIDOSE is a program system for calculating the consequences of a radioactive release to the atmosphere. The program is applicable for computation of dispersion in a rnage of 0 - 50 km from the release point. The Gaussion plume model is used for calculating the external dose from activity in the atmosphere, on the ground and the internal dose via inhalation. Radioactive decay, as well as growth and decay of daughter products are accounted for. The influence of dry deposition and wash-out are also considered. It is possible to treat time-dependent release-rates of 1 - 24 hours duration and constant release-rates for up to one year. The program system also contains routines for the calculation of collective dose and health effects. The system operates in a statistical manner. Many weather-situations, based on measured data, can be analysed and statistical properties, such as cumulative frequences, can be calculated. (author)

Computer Programs for Calculating and Plotting the Stability Characteristics of a Balloon Tethered in a Wind

Science.gov (United States)

Bennett, R. M.; Bland, S. R.; Redd, L. T.

1973-01-01

Computer programs for calculating the stability characteristics of a balloon tethered in a steady wind are presented. Equilibrium conditions, characteristic roots, and modal ratios are calculated for a range of discrete values of velocity for a fixed tether-line length. Separate programs are used: (1) to calculate longitudinal stability characteristics, (2) to calculate lateral stability characteristics, (3) to plot the characteristic roots versus velocity, (4) to plot the characteristic roots in root-locus form, (5) to plot the longitudinal modes of motion, and (6) to plot the lateral modes for motion. The basic equations, program listings, and the input and output data for sample cases are presented, with a brief discussion of the overall operation and limitations. The programs are based on a linearized, stability-derivative type of analysis, including balloon aerodynamics, apparent mass, buoyancy effects, and static forces which result from the tether line.

MILDOS - a computer program for calculating environmental radiation doses from uranium recovery operations. Research report

International Nuclear Information System (INIS)

Strenge, D.L.; Bander, T.J.

1981-04-01

MILDOS is a Fortran Computer Code which calculates the dose commitments received by individuals and the general population within an 80 kilometer radius of an operating uranium recovery facility. In addition air and ground concentrations are presented for individual locations, as well as for a generalized population grid. Extra-regional population doses resulting from transport of radon and export of agricultural produce are also displayed. The transport of radiological emissions from point and area sources is predicted by using a sector-averaged Gaussian plume dispersion model. Mechanisms such as radioactive decay, plume depletion by deposition, ingrowth of daughter products and resuspension of deposited radionuclides are included in the transport model. Alterations in operation throughout the facility's lifetime can be accounted for in the input stream. The pathways considered are: inhalation; external exposure from ground shine and cloud immersion; and ingestion of vegetables, meat and milk. Dose commitments are calculated primarily on the basis of the recommendations of the International Commission on Radiological Protection (ICRP). Predictive 40 CFR 190 and 10 CFR 20 compliances are also performed. This computer code is designed primarily for uranium milling facilities and should not be used for operations with different radionuclides or processes

A FORTRAN program for an IBM PC compatible computer for calculating kinematical electron diffraction patterns

International Nuclear Information System (INIS)

Skjerpe, P.

1989-01-01

This report describes a computer program which is useful in transmission electron microscopy. The program is written in FORTRAN and calculates kinematical electron diffraction patterns in any zone axis from a given crystal structure. Quite large unit cells, containing up to 2250 atoms, can be handled by the program. The program runs on both the Helcules graphic card and the standard IBM CGA card

POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs

International Nuclear Information System (INIS)

Hardie, R.W.

1982-02-01

POPCYCLE, a computer code designed to calculate levelized life-cycle power costs for nuclear and fossil electrical generating plants is described. Included are (1) derivations of the equations and a discussion of the methodology used by POPCYCLE, (2) a description of the input required by the code, (3) a listing of the input for a sample case, and (4) the output for a sample case

Efficient Probability of Failure Calculations for QMU using Computational Geometry LDRD 13-0144 Final Report

Energy Technology Data Exchange (ETDEWEB)

Mitchell, Scott A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ebeida, Mohamed Salah [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Romero, Vicente J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Swiler, Laura Painton [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rushdi, Ahmad A. [Univ. of Texas, Austin, TX (United States); Abdelkader, Ahmad [Univ. of Maryland, College Park, MD (United States)

2015-09-01

This SAND report summarizes our work on the Sandia National Laboratory LDRD project titled "Efficient Probability of Failure Calculations for QMU using Computational Geometry" which was project #165617 and proposal #13-0144. This report merely summarizes our work. Those interested in the technical details are encouraged to read the full published results, and contact the report authors for the status of the software and follow-on projects.

PopSc: Computing Toolkit for Basic Statistics of Molecular Population Genetics Simultaneously Implemented in Web-Based Calculator, Python and R.

Science.gov (United States)

Chen, Shi-Yi; Deng, Feilong; Huang, Ying; Li, Cao; Liu, Linhai; Jia, Xianbo; Lai, Song-Jia

2016-01-01

Although various computer tools have been elaborately developed to calculate a series of statistics in molecular population genetics for both small- and large-scale DNA data, there is no efficient and easy-to-use toolkit available yet for exclusively focusing on the steps of mathematical calculation. Here, we present PopSc, a bioinformatic toolkit for calculating 45 basic statistics in molecular population genetics, which could be categorized into three classes, including (i) genetic diversity of DNA sequences, (ii) statistical tests for neutral evolution, and (iii) measures of genetic differentiation among populations. In contrast to the existing computer tools, PopSc was designed to directly accept the intermediate metadata, such as allele frequencies, rather than the raw DNA sequences or genotyping results. PopSc is first implemented as the web-based calculator with user-friendly interface, which greatly facilitates the teaching of population genetics in class and also promotes the convenient and straightforward calculation of statistics in research. Additionally, we also provide the Python library and R package of PopSc, which can be flexibly integrated into other advanced bioinformatic packages of population genetics analysis.

PopSc: Computing Toolkit for Basic Statistics of Molecular Population Genetics Simultaneously Implemented in Web-Based Calculator, Python and R.

Directory of Open Access Journals (Sweden)

Shi-Yi Chen

Full Text Available Although various computer tools have been elaborately developed to calculate a series of statistics in molecular population genetics for both small- and large-scale DNA data, there is no efficient and easy-to-use toolkit available yet for exclusively focusing on the steps of mathematical calculation. Here, we present PopSc, a bioinformatic toolkit for calculating 45 basic statistics in molecular population genetics, which could be categorized into three classes, including (i genetic diversity of DNA sequences, (ii statistical tests for neutral evolution, and (iii measures of genetic differentiation among populations. In contrast to the existing computer tools, PopSc was designed to directly accept the intermediate metadata, such as allele frequencies, rather than the raw DNA sequences or genotyping results. PopSc is first implemented as the web-based calculator with user-friendly interface, which greatly facilitates the teaching of population genetics in class and also promotes the convenient and straightforward calculation of statistics in research. Additionally, we also provide the Python library and R package of PopSc, which can be flexibly integrated into other advanced bioinformatic packages of population genetics analysis.

Calculation of normalised organ and effective doses to adult reference computational phantoms from contemporary computed tomography scanners

International Nuclear Information System (INIS)

Jansen, Jan T.M.; Shrimpton, Paul C.

2010-01-01

The general-purpose Monte Carlo radiation transport code MCNPX has been used to simulate photon transport and energy deposition in anthropomorphic phantoms due to the x-ray exposure from the Philips iCT 256 and Siemens Definition CT scanners, together with the previously studied General Electric 9800. The MCNPX code was compiled with the Intel FORTRAN compiler and run on a Linux PC cluster. A patch has been successfully applied to reduce computing times by about 4%. The International Commission on Radiological Protection (ICRP) has recently published the Adult Male (AM) and Adult Female (AF) reference computational voxel phantoms as successors to the Medical Internal Radiation Dose (MIRD) stylised hermaphrodite mathematical phantoms that form the basis for the widely-used ImPACT CT dosimetry tool. Comparisons of normalised organ and effective doses calculated for a range of scanner operating conditions have demonstrated significant differences in results (in excess of 30%) between the voxel and mathematical phantoms as a result of variations in anatomy. These analyses illustrate the significant influence of choice of phantom on normalised organ doses and the need for standardisation to facilitate comparisons of dose. Further such dose simulations are needed in order to update the ImPACT CT Patient Dosimetry spreadsheet for contemporary CT practice. (author)

General-purpose software for science technology calculation

International Nuclear Information System (INIS)

Aikawa, Hiroshi

1999-01-01

We have developed many general-purpose softwares for parallel processing of science technology calculation. This paper reported six softwares such as STA (Seamless Thinking Aid) basic soft, parallel numerical computation library, grid formation software for parallel computer, real-time visualizing system, parallel benchmark test system and object-oriented parallel programing method. STA is a user interface software to perform a total environment for parallel programing, a network computing environment for various parallel computers and a desktop computing environment via Web. Some examples using the above softwares are explained. One of them is a simultaneous parallel calculation of both analysis of flow and structure of supersonic transport to design of them. The other is various kinds of computer parallel calculations for nuclear fusion reaction such as a molecular dynamic calculation and a calculation of reactor structure and fluid. These softs are opened to the public by the home page {http://guide.tokai.jaeri.go.jp/ccse/}. (S.Y.)

Course on hybrid calculation

International Nuclear Information System (INIS)

Weill, J.; Tellier; Bonnemay; Craigne; Chareton; Di Falco

1969-02-01

After a definition of hybrid calculation (combination of analogue and digital calculation) with a distinction between series and parallel hybrid computing, and a description of a hybrid computer structure and of task sharing between computers, this course proposes a description of hybrid hardware used in Saclay and Cadarache computing centres, and of operations performed by these systems. The next part addresses issues related to programming languages and software. The fourth part describes how a problem is organised for its processing on these computers. Methods of hybrid analysis are then addressed: resolution of optimisation problems, of partial differential equations, and of integral equations by means of different methods (gradient, maximum principle, characteristics, functional approximation, time slicing, Monte Carlo, Neumann iteration, Fischer iteration)

Computational tools for the construction of calibration curves for use in dose calculations in radiotherapy treatment planning

International Nuclear Information System (INIS)

Oliveira, Alex C.H.; Vieira, Jose W.; Escola Politecnica de Pernambuco , Recife, PE

2011-01-01

The realization of tissue inhomogeneity corrections in image-based treatment planning improves the accuracy of radiation dose calculations for patients undergoing external-beam radiotherapy. Before the tissue inhomogeneity correction can be applied, the relationship between the computed tomography (CT) numbers and density must be established. This relationship is typically established by a calibration curve empirically obtained from CT images of a phantom that has several inserts of tissue-equivalent materials, covering a wide range of densities. This calibration curve is scanner-dependent and allows the conversion of CT numbers in densities for use in dose calculations. This paper describes the implementation of computational tools necessary to construct calibration curves. These tools are used for reading and displaying of CT images in DICOM format, determination of the mean CT numbers (and their standard deviations) of each tissue-equivalent material and construction of calibration curves by fits with bilinear equations. All these tools have been implemented in the Microsoft Visual Studio 2010 in C≠ programming language. (author)

A comparison of shielding calculation methods for multi-slice computed tomography (CT) systems

International Nuclear Information System (INIS)

Cole, J A; Platten, D J

2008-01-01

Currently in the UK, shielding calculations for computed tomography (CT) systems are based on the BIR-IPEM (British Institute of Radiology and Institute of Physics in Engineering in Medicine) working group publication from 2000. Concerns have been raised internationally regarding the accuracy of the dose plots on which this method depends and the effect that new scanner technologies may have. Additionally, more recent shielding methods have been proposed by the NCRP (National Council on Radiation Protection) from the USA. Thermoluminescent detectors (TLDs) were placed in three CT scanner rooms at different positions for several weeks before being processed. Patient workload and dose data (DLP: the dose length product and mAs: the tube current-time product) were collected for this period. Individual dose data were available for more than 95% of patients scanned and the remainder were estimated. The patient workload data were used to calculate expected scatter radiation for each TLD location by both the NCRP and BIR-IPEM methods. The results were then compared to the measured scattered radiation. Calculated scattered air kerma and the minimum required lead shielding were found to be frequently overestimated compared to the measured air kerma, on average almost five times the measured scattered air kerma.

EXTENDCHAIN: a package of computer programs for calculating the buildup of heavy metals, fission products, and activation products in reactor fuel elements

International Nuclear Information System (INIS)

Robertson, M.W.

1977-01-01

Design of HTGR recycle and refabrication facilities requires a detailed knowledge of the concentrations of around 400 nuclides which are segregated into four different fuel particle types. The EXTENDCHAIN package of computer programs and the supporting input data files were created to provide an efficient method for calculating the 1600 different concentrations required. The EXTENDCHAIN code performs zero-dimensional nuclide burnup, decay, and activation calculations in nine energy groups for up to 108 nuclides per run. Preparation and handling of the input and output for the sixteen EXTENDCHAIN runs required to produce the desired data are the most time consuming tasks in the computation of the spent fuel element composition. The EXTENDCHAIN package of computer programs contains four codes to aid in the preparation and handling of these data. Most of the input data such as cross sections, decay constants, and the nuclide interconnection scheme will not change when calculating new cases. These data were developed for the life cycle of a typical HTGR and stored on archive tapes for future use. The fuel element composition for this typical HTGR life has been calculated and the results for an equilibrium recycle reload are presented. 12 figures, 7 tables

ZOCO V - a computer code for the calculation of time-dependent spatial pressure distribution in reactor containments

International Nuclear Information System (INIS)

Mansfeld, G.; Schally, P.

1978-06-01

ZOCO V is a computer code which can calculate the time- and space- dependent pressure distribution in containments of water-cooled nuclear power reactors (both full pressure containments and pressure suppression systems) following a loss-of-coolant accident, caused by the rupture of a main coolant or steam pipe

Computer code for shielding calculations of x-rays rooms

International Nuclear Information System (INIS)

Affonso, R.R.W.; Borges, D. da S.; Lava, D.D.; Moreira, M. de L.; Guimarães, A.C.F.

2015-01-01

The building an effective barrier against ionizing radiation present in radiographic rooms requires consideration of many variables. The methodology used for thickness specification of primary and secondary, barrier of a traditional radiographic room, considers the following factors: Use Factor, Occupational Factor, distance between the source and the wall, Workload, Kerma in the air and distance between the patient and the source. With these data it was possible to develop a computer code, which aims to identify and use variables in functions obtained through graphics regressions provided by NCRP-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) report, for shielding calculation of room walls, and the walls of the dark room and adjacent areas. With the implemented methodology, it was made a code validation by comparison of results with a study case provided by the report. The obtained values for thickness comprise different materials such as concrete, lead and glass. After validation it was made a case study of an arbitrary radiographic room.The development of the code 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. (authors)

A computer program for accident calculations of a standard pressurized water reactor

International Nuclear Information System (INIS)

Keutner, H.

1979-01-01

In this computer program the dynamic of a standard pressurized water reactor should be realized by both circulation loops with all important components. All important phenomena are taken into consideration, which appear for calculation of disturbances in order to state a realistic process for some minutes after a disturbance or a desired change of condition. In order to optimize the computer time simplifications are introduced in the statement of a differential-algebraic equalization system such that all important effects are taken into consideration. The model analysis starts from the heat production of the fuel rod via cladding material to the cooling medium water and considers the delay time from the core to the steam generator. Alternations of the cooling medium pressure as well as the different temperatures in the primary loop influence the pressuring system - the pressurizer - which is realized by a water and a steam zone with saturated and superheated steam respectively saturated and undercooled water with injection, heating and blow-down devices. The bilance of the steam generator to the secondary loop realizes the process engineering devices. Thereby the control regulation of the steam pressure and the reactor performance is realized. (orig.) [de

Comparison of RESRAD with hand calculations

International Nuclear Information System (INIS)

Rittmann, P.D.

1995-09-01

This report is a continuation of an earlier comparison done with two other computer programs, GENII and PATHRAE. The dose calculations by the two programs were compared with each other and with hand calculations. These band calculations have now been compared with RESRAD Version 5.41 to examine the use of standard models and parameters in this computer program. The hand calculations disclosed a significant computational error in RESRAD. The Pu-241 ingestion doses are five orders of magnitude too small. In addition, the external doses from some nuclides differ greatly from expected values. Both of these deficiencies have been corrected in later versions of RESRAD

Calculator. Owning a Small Business.

Science.gov (United States)

Parma City School District, OH.

Seven activities are presented in this student workbook designed for an exploration of small business ownership and the use of the calculator in this career. Included are simulated situations in which students must use a calculator to compute property taxes; estimate payroll taxes and franchise taxes; compute pricing, approximate salaries,…

Range calculations for spallation recoils in ThF4 by use of the computer code 'Marlowe'

International Nuclear Information System (INIS)

Westmeier, W.; Roessler, K.

1978-12-01

The determination of cross sections of spallation reactions requires a knowledge of the target thickness since only the products recoiling from the target are measured and their yield depends on the range. The effective target thickness is a function of the projectile's Z, A and spallation recoil energy and, thus, varies for the individual products. The computer code MARLOWE was used to evaluate energy vs. range curves in the binary collisions approximation. The program was extended to the high energy regime taking into account the stripping of electrons from the projectile and the concomitant changes in the interaction potentials especially for the inelastic part of the collisions. A complementary computer program LATTIC was developed for the parameterization of the lattice description. This code enables the application of MARLOWE to target materials with complicated crystallographic structure. Test calculations for a series of projectile/target combinations showed a reasonable agreement with experimental recoil ranges of Pd, Ag, Os and Ir isotopes from proton induced spallation in Ag, In and Pb targets, respectively. MARLOWE was then applied to calculate product ranges of the 232 Th(p,spall)X-reaction in the ployatomic system ThF 4 . The calculated energy vs. range curves enabled the evaluation of the mean spallation recoil ranges for all possible products, e.g. 170.8 μg/cm 2 for 192 Tl, 115.2 μg/cm 2 for 208 At and 37.1 μg/cm 2 for 223 Ac. (orig.)

ERATO - a computer program for the calculation of induced eddy-currents in three-dimensional conductive structures

International Nuclear Information System (INIS)

Benner, J.

1985-10-01

The computer code ERATO is used for the calculation of eddy-currents in three-dimensional conductive structures and their secondary magnetic field. ERATO is a revised version of the code FEDIFF, developed at IPP Garching. For the calculation the Finite-Element-Network (FEN) method is used, where the structure is simulated by an equivalent electric network. In the ERATO-code, the calculation of the finite-element discretization, the eddy-current analysis, and the final evaluation of the results are done in separate programs. So the eddy-current analysis as the central step is perfectly independent of a special geometry. For the finite-element discretization there are two so called preprocessors, which treat a torus-segment and a rectangular, flat plate. For the final evaluation postprocessors are used, by which the current-distributions can be printed and plotted. In the report, the theoretical foundation of the FEN-Method is discussed, the structure and the application of the programs (preprocessors, analysis-program, postprocessors, supporting programs) are shown, and two examples for calculations are presented. (orig.) [de

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)

Computing and physical methods to calculate Pu

International Nuclear Information System (INIS)

Mohamed, Ashraf Elsayed Mohamed

2013-01-01

Main limitations due to the enhancement of the plutonium content are related to the coolant void effect as the spectrum becomes faster, the neutron flux in the thermal region tends towards zero and is concentrated in the region from 10 Ke to 1 MeV. Thus, all captures by 240 Pu and 242 Pu in the thermal and epithermal resonance disappear and the 240 Pu and 242 Pu contributions to the void effect became positive. The higher the Pu content and the poorer the Pu quality, the larger the void effect. The core control in nominal or transient conditions Pu enrichment leads to a decrease in (B eff.), the efficiency of soluble boron and control rods. Also, the Doppler effect tends to decrease when Pu replaces U, so, that in case of transients the core could diverge again if the control is not effective enough. As for the voiding effect, the plutonium degradation and the 240 Pu and 242 Pu accumulation after multiple recycling lead to spectrum hardening and to a decrease in control. One solution would be to use enriched boron in soluble boron and shutdown rods. In this paper, I discuss and show the advanced computing and physical methods to calculate Pu inside the nuclear reactors and glovebox and the different solutions to be used to overcome the difficulties that effect, on safety parameters and on reactor performance, and analysis the consequences of plutonium management on the whole fuel cycle like Raw materials savings, fraction of nuclear electric power involved in the Pu management. All through two types of scenario, one involving a low fraction of the nuclear park dedicated to plutonium management, the other involving a dilution of the plutonium in all the nuclear park. (author)

Development of 1-year-old computational phantom and calculation of organ doses during CT scans using Monte Carlo simulation

International Nuclear Information System (INIS)

Pan, Yuxi; Qiu, Rui; Ge, Chaoyong; Xie, Wenzhang; Li, Junli; Gao, Linfeng; Zheng, Junzheng

2014-01-01

With the rapidly growing number of CT examinations, the consequential radiation risk has aroused more and more attention. The average dose in each organ during CT scans can only be obtained by using Monte Carlo simulation with computational phantoms. Since children tend to have higher radiation sensitivity than adults, the radiation dose of pediatric CT examinations requires special attention and needs to be assessed accurately. So far, studies on organ doses from CT exposures for pediatric patients are still limited. In this work, a 1-year-old computational phantom was constructed. The body contour was obtained from the CT images of a 1-year-old physical phantom and the internal organs were deformed from an existing Chinese reference adult phantom. To ensure the organ locations in the 1-year-old computational phantom were consistent with those of the physical phantom, the organ locations in 1-year-old computational phantom were manually adjusted one by one, and the organ masses were adjusted to the corresponding Chinese reference values. Moreover, a CT scanner model was developed using the Monte Carlo technique and the 1-year-old computational phantom was applied to estimate organ doses derived from simulated CT exposures. As a result, a database including doses to 36 organs and tissues from 47 single axial scans was built. It has been verified by calculation that doses of axial scans are close to those of helical scans; therefore, this database could be applied to helical scans as well. Organ doses were calculated using the database and compared with those obtained from the measurements made in the physical phantom for helical scans. The differences between simulation and measurement were less than 25% for all organs. The result shows that the 1-year-old phantom developed in this work can be used to calculate organ doses in CT exposures, and the dose database provides a method for the estimation of 1-year-old patient doses in a variety of CT examinations. (paper)

Three-dimensional coupled Monte Carlo-discrete ordinates computational scheme for shielding calculations of large and complex nuclear facilities

International Nuclear Information System (INIS)

Chen, Y.; Fischer, U.

2005-01-01

Shielding calculations of advanced nuclear facilities such as accelerator based neutron sources or fusion devices of the tokamak type are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. While the complexity of the geometry in the shielding calculation can be hardly handled by the discrete ordinates method, the deep penetration of radiation through bulk shields is a severe challenge for the Monte Carlo particle transport technique. This work proposes a dedicated computational scheme for coupled Monte Carlo-Discrete Ordinates transport calculations to handle this kind of shielding problems. The Monte Carlo technique is used to simulate the particle generation and transport in the target region with both complex geometry and reaction physics, and the discrete ordinates method is used to treat the deep penetration problem in the bulk shield. The coupling scheme has been implemented in a program system by loosely integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a newly developed coupling interface program for mapping process. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities. (authors)

FLUST-2D - A computer code for the calculation of the two-dimensional flow of a compressible medium in coupled retangular areas

International Nuclear Information System (INIS)

Enderle, G.

1979-01-01

The computer-code FLUST-2D is able to calculate the two-dimensional flow of a compressible fluid in arbitrary coupled rectangular areas. In a finite-difference scheme the program computes pressure, density, internal energy and velocity. Starting with a basic set of equations, the difference equations in a rectangular grid are developed. The computational cycle for coupled fluid areas is described. Results of test calculations are compared to analytical solutions and the influence of time step and mesh size are investigated. The program was used to precalculate the blowdown experiments of the HDR experimental program. Downcomer, plena, internal vessel region, blowdown pipe and a containment area have been modelled two-dimensionally. The major results of the precalculations are presented. This report also contains a description of the code structure and user information. (orig.) [de

Computer program MCAP-TOSS calculates steady-state fluid dynamics of coolant in parallel channels and temperature distribution in surrounding heat-generating solid

Science.gov (United States)

Lee, A. Y.

1967-01-01

Computer program calculates the steady state fluid distribution, temperature rise, and pressure drop of a coolant, the material temperature distribution of a heat generating solid, and the heat flux distributions at the fluid-solid interfaces. It performs the necessary iterations automatically within the computer, in one machine run.

Computing NLTE Opacities -- Node Level Parallel Calculation

Energy Technology Data Exchange (ETDEWEB)

Holladay, Daniel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-09-11

Presentation. The goal: to produce a robust library capable of computing reasonably accurate opacities inline with the assumption of LTE relaxed (non-LTE). Near term: demonstrate acceleration of non-LTE opacity computation. Far term (if funded): connect to application codes with in-line capability and compute opacities. Study science problems. Use efficient algorithms that expose many levels of parallelism and utilize good memory access patterns for use on advanced architectures. Portability to multiple types of hardware including multicore processors, manycore processors such as KNL, GPUs, etc. Easily coupled to radiation hydrodynamics and thermal radiative transfer codes.

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.

Ultrafast layer based computer-generated hologram calculation with sparse template holographic fringe pattern for 3-D object.

Science.gov (United States)

Kim, Hak Gu; Man Ro, Yong

2017-11-27

In this paper, we propose a new ultrafast layer based CGH calculation that exploits the sparsity of hologram fringe pattern in 3-D object layer. Specifically, we devise a sparse template holographic fringe pattern. The holographic fringe pattern on a depth layer can be rapidly calculated by adding the sparse template holographic fringe patterns at each object point position. Since the size of sparse template holographic fringe pattern is much smaller than that of the CGH plane, the computational load can be significantly reduced. Experimental results show that the proposed method achieves 10-20 msec for 1024x1024 pixels providing visually plausible results.

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

International Nuclear Information System (INIS)

Ivascu, M.

1983-10-01

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

ParShield: A computer program for calculating attenuation parameters of the gamma rays and the fast neutrons

International Nuclear Information System (INIS)

Elmahroug, Y.; Tellili, B.; Souga, C.; Manai, K.

2015-01-01

Highlights: • Description of the theoretical method used by the ParShield program. • Description of the ParShield program. • Test and validation the ParShield program. - Abstract: This study aims to present a new computer program called ParShield which determines the neutron and gamma-ray shielding parameters. This program can calculate the total mass attenuation coefficients (μ t ), the effective atomic numbers (Z eff ) and the effective electron densities (N eff ) for gamma rays and it can also calculate the effective removal cross-sections (Σ R ) for fast neutrons for mixtures and compounds. The results obtained for the gamma rays by using ParShield were compared with the results calculated by the WinXcom program and the measured results. The obtained values of (Σ R ) were tested by comparing them with the measured results,the manually calculated results and with the results obtained by using MERCSFN program and an excellent agreement was found between them. The ParShield program can be used as a fast and effective tool to choose and compare the shielding materials, especially for the determination of (Z eff ) and (N eff ), there is no other programs in the literature which can calculate

ACRO - a computer program for calculating organ doses from acute or chronic inhalation and ingestion of radionuclides

International Nuclear Information System (INIS)

Hirayama, Akio; Kishimoto, Yoichiro; Shinohara, Kunihiko.

1978-01-01

The computer program ACRO has been developed to calculate organ doses from acute or chronic inhalation and ingestion of radionuclides. The ICRP Task Group Lung Model (TGLM) was used for inhalation model, and a simple one-compartment model for ingestion. This program is written in FORTRAN IV, and can be executed with storage requirements of about 260 K bytes. (auth.)

From parallel to distributed computing for reactive scattering calculations

International Nuclear Information System (INIS)

Lagana, A.; Gervasi, O.; Baraglia, R.

1994-01-01

Some reactive scattering codes have been ported on different innovative computer architectures ranging from massively parallel machines to clustered workstations. The porting has required a drastic restructuring of the codes to single out computationally decoupled cpu intensive subsections. The suitability of different theoretical approaches for parallel and distributed computing restructuring is discussed and the efficiency of related algorithms evaluated

SIMPLE-2: a computer code for calculation of steady-state thermal behavior of rod bundles with flow sweeping

International Nuclear Information System (INIS)

Jones, O.C. Jr.; Yao, S.; Henry, R.E.

1976-01-01

A computer code has been developed for use in making single-phase thermal hydraulic calculations in rod bundle arrays with flow sweeping due to spiral wraps as the predominant crossflow mixing effect. This code, called SIMPLE-2, makes the assumption that the axial pressure gradient is identical for each subchannel over a given axial increment, and is unique in that no empirical coefficients must be specified for its use. Results from this code have been favorably compared with experimental data for both uniform and highly nonuniform power distributions. Typical calculations for various bundle sizes applicable to the LMBR program are included

Ab initio quasi-particle approximation bandgaps of silicon nanowires calculated at density functional theory/local density approximation computational effort

Energy Technology Data Exchange (ETDEWEB)

Ribeiro, M., E-mail: ribeiro.jr@oorbit.com.br [Office of Operational Research for Business Intelligence and Technology, Principal Office, Buffalo, Wyoming 82834 (United States)

2015-06-21

Ab initio calculations of hydrogen-passivated Si nanowires were performed using density functional theory within LDA-1/2, to account for the excited states properties. A range of diameters was calculated to draw conclusions about the ability of the method to correctly describe the main trends of bandgap, quantum confinement, and self-energy corrections versus the diameter of the nanowire. Bandgaps are predicted with excellent accuracy if compared with other theoretical results like GW, and with the experiment as well, but with a low computational cost.

Ab initio quasi-particle approximation bandgaps of silicon nanowires calculated at density functional theory/local density approximation computational effort

International Nuclear Information System (INIS)

Ribeiro, M.

2015-01-01

Ab initio calculations of hydrogen-passivated Si nanowires were performed using density functional theory within LDA-1/2, to account for the excited states properties. A range of diameters was calculated to draw conclusions about the ability of the method to correctly describe the main trends of bandgap, quantum confinement, and self-energy corrections versus the diameter of the nanowire. Bandgaps are predicted with excellent accuracy if compared with other theoretical results like GW, and with the experiment as well, but with a low computational cost

DNAStat, version 2.1--a computer program for processing genetic profile databases and biostatistical calculations.

Science.gov (United States)

Berent, Jarosław

2010-01-01

This paper presents the new DNAStat version 2.1 for processing genetic profile databases and biostatistical calculations. The popularization of DNA studies employed in the judicial system has led to the necessity of developing appropriate computer programs. Such programs must, above all, address two critical problems, i.e. the broadly understood data processing and data storage, and biostatistical calculations. Moreover, in case of terrorist attacks and mass natural disasters, the ability to identify victims by searching related individuals is very important. DNAStat version 2.1 is an adequate program for such purposes. The DNAStat version 1.0 was launched in 2005. In 2006, the program was updated to 1.1 and 1.2 versions. There were, however, slight differences between those versions and the original one. The DNAStat version 2.0 was launched in 2007 and the major program improvement was an introduction of the group calculation options with the potential application to personal identification of mass disasters and terrorism victims. The last 2.1 version has the option of language selection--Polish or English, which will enhance the usage and application of the program also in other countries.

Shielding calculational system for plutonium

International Nuclear Information System (INIS)

Zimmerman, M.G.; Thomsen, D.H.

1975-08-01

A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

TRAFIC, a computer program for calculating the release of metallic fission products from an HTGR core

International Nuclear Information System (INIS)

Smith, P.D.

1978-02-01

A special purpose computer program, TRAFIC, is presented for calculating the release of metallic fission products from an HTGR core. The program is based upon Fick's law of diffusion for radioactive species. One-dimensional transient diffusion calculations are performed for the coated fuel particles and for the structural graphite web. A quasi steady-state calculation is performed for the fuel rod matrix material. The model accounts for nonlinear adsorption behavior in the fuel rod gap and on the coolant hole boundary. The TRAFIC program is designed to operate in a core survey mode; that is, it performs many repetitive calculations for a large number of spatial locations in the core. This is necessary in order to obtain an accurate volume integrated release. For this reason the program has been designed with calculational efficiency as one of its main objectives. A highly efficient numerical method is used in the solution. The method makes use of the Duhamel superposition principle to eliminate interior spatial solutions from consideration. Linear response functions relating the concentrations and mass fluxes on the boundaries of a homogeneous region are derived. Multiple regions are numerically coupled through interface conditions. Algebraic elimination is used to reduce the equations as far as possible. The problem reduces to two nonlinear equations in two unknowns, which are solved using a Newton Raphson technique

Use of Monte Carlo simulation software for calculating effective dose in cone beam computed tomography

Energy Technology Data Exchange (ETDEWEB)

Gomes B, W. O., E-mail: wilsonottobatista@gmail.com [Instituto Federal da Bahia, Rua Emidio dos Santos s/n, Barbalho 40301-015, Salvador de Bahia (Brazil)

2016-10-15

This study aimed to develop a geometry of irradiation applicable to the software PCXMC and the consequent calculation of effective dose in applications of the Computed Tomography Cone Beam (CBCT). We evaluated two different CBCT equipment s for dental applications: Care stream Cs 9000 3-dimensional tomograph; i-CAT and GENDEX GXCB-500. Initially characterize each protocol measuring the surface kerma input and the product kerma air-area, P{sub KA}, with solid state detectors RADCAL and PTW transmission chamber. Then we introduce the technical parameters of each preset protocols and geometric conditions in the PCXMC software to obtain the values of effective dose. The calculated effective dose is within the range of 9.0 to 15.7 μSv for 3-dimensional computer 9000 Cs; within the range 44.5 to 89 μSv for GXCB-500 equipment and in the range of 62-111 μSv for equipment Classical i-CAT. These values were compared with results obtained dosimetry using TLD implanted in anthropomorphic phantom and are considered consistent. Os effective dose results are very sensitive to the geometry of radiation (beam position in mathematical phantom). This factor translates to a factor of fragility software usage. But it is very useful to get quick answers to regarding process optimization tool conclusions protocols. We conclude that use software PCXMC Monte Carlo simulation is useful assessment protocols for CBCT tests in dental applications. (Author)

Heterogeneous Calculation of {epsilon}

Energy Technology Data Exchange (ETDEWEB)

Jonsson, Alf

1961-02-15

A heterogeneous method of calculating the fast fission factor given by Naudet has been applied to the Carlvik - Pershagen definition of {epsilon}. An exact calculation of the collision probabilities is included in the programme developed for the Ferranti - Mercury computer.

Heterogeneous Calculation of ε

International Nuclear Information System (INIS)

Jonsson, Alf

1961-02-01

A heterogeneous method of calculating the fast fission factor given by Naudet has been applied to the Carlvik - Pershagen definition of ε. An exact calculation of the collision probabilities is included in the programme developed for the Ferranti - Mercury computer

BUSH: A computer code for calculating steady state heat transfer in LWR rod bundles under accident conditions

International Nuclear Information System (INIS)

Shepherd, I.M.

1982-01-01

The computer code BUSH has been developed for the calculation of steady state heat transfer in a rod bundle. For a given power, flow and geometry it can calculate the temperatures in the rods, coolant and shroud assuming that at any axial level each rod can be described by one temperature and the coolant fluid is also radially uniform at this level. Heat transfer by convection and radiation are handled and the geometry is flexible enough to model nearly all types of envisaged shroud design for the SUPERSARA test series. The modular way in which BUSH has been written makes it suitable for future development, either within the present BUSH framework or as part of a more advanced code

Calculation of the number of branches of multi-valued decision trees in computer aided importance rank of parameters

Directory of Open Access Journals (Sweden)

Tiszbierek Agnieszka

2017-01-01

Full Text Available An elaborated digital computer programme supporting the time-consuming process of selecting the importance rank of construction and operation parameters by means of stating optimum sets is based on the Quine – McCluskey algorithm of minimizing individual partial multi-valued logic functions. The example with real time data, calculated by means of the programme, showed that among the obtained optimum sets there were such which had a different number of real branches after being presented on the multi-valued logic decision tree. That is why an idea of elaborating another functionality of the programme – a module calculating the number of branches of real, multi-valued logic decision trees presenting optimum sets chosen by the programme was pursued. This paper presents the idea and the method for developing a module calculating the number of branches, real for each of optimum sets indicated by the programme, as well as to the calculation process.

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

International Nuclear Information System (INIS)

Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F.

2014-01-01

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

Computational methods in calculating superconducting current problems

Science.gov (United States)

Brown, David John, II

Various computational problems in treating superconducting currents are examined. First, field inversion in spatial Fourier transform space is reviewed to obtain both one-dimensional transport currents flowing down a long thin tape, and a localized two-dimensional current. The problems associated with spatial high-frequency noise, created by finite resolution and experimental equipment, are presented, and resolved with a smooth Gaussian cutoff in spatial frequency space. Convergence of the Green's functions for the one-dimensional transport current densities is discussed, and particular attention is devoted to the negative effects of performing discrete Fourier transforms alone on fields asymptotically dropping like 1/r. Results of imaging simulated current densities are favorably compared to the original distributions after the resulting magnetic fields undergo the imaging procedure. The behavior of high-frequency spatial noise, and the behavior of the fields with a 1/r asymptote in the imaging procedure in our simulations is analyzed, and compared to the treatment of these phenomena in the published literature. Next, we examine calculation of Mathieu and spheroidal wave functions, solutions to the wave equation in elliptical cylindrical and oblate and prolate spheroidal coordinates, respectively. These functions are also solutions to Schrodinger's equations with certain potential wells, and are useful in solving time-varying superconducting problems. The Mathieu functions are Fourier expanded, and the spheroidal functions expanded in associated Legendre polynomials to convert the defining differential equations to recursion relations. The infinite number of linear recursion equations is converted to an infinite matrix, multiplied by a vector of expansion coefficients, thus becoming an eigenvalue problem. The eigenvalue problem is solved with root solvers, and the eigenvector problem is solved using a Jacobi-type iteration method, after preconditioning the

GAPCON-THERMAL-2: a computer program for calculating the thermal behavior of an oxide fuel rod

International Nuclear Information System (INIS)

Beyer, C.E.; Hann, C.R.; Lanning, D.D.; Panisko, F.E.; Parchen, L.J.

1975-11-01

A description is presented of the computer code GAPCON THERMAL-2, a light water reactor (LWR) fuel thermal performance prediction code. GAPCON-THERMAL-2, is intended to be used as a calculational tool for reactor fuel steady-state thermal performance and to provide input for accident analyses. Some models used in the code provide best estimate as well as conservative predictions. Each of the individual models in the code is based on the best available data

Development of selective photoionization spectroscopy technology - Development of a computer program to calculate selective ionization of atoms with multistep processes

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Soon; Nam, Baek Il [Myongji University, Seoul (Korea, Republic of)

1995-08-01

We have developed computer programs to calculate 2-and 3-step selective resonant multiphoton ionization of atoms. Autoionization resonances in the final continuum can be put into account via B-Spline basis set method. 8 refs., 5 figs. (author)

VVER 1000 SBO calculations with pressuriser relief valve stuck open with ASTEC computer code

International Nuclear Information System (INIS)

Atanasova, B.P.; Stefanova, A.E.; Groudev, P.P.

2012-01-01

Highlights: ► We modelled the ASTEC input file for accident scenario (SBO) and focused analyses on the behaviour of core degradation. ► We assumed opening and stuck-open of pressurizer relief valve during performance of SBO scenario. ► ASTEC v1.3.2 has been used as a reference code for the comparison study with the new version of ASTEC code. - Abstract: The objective of this paper is to present the results obtained from performing the calculations with ASTEC computer code for the Source Term evaluation for specific severe accident transient. The calculations have been performed with the new version of ASTEC. The ASTEC V2 code version is released by the French IRSN (Institut de Radioprotection at de surete nucleaire) and Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), Germany. This investigation has been performed in the framework of the SARNET2 project (under the Euratom 7th framework program) by Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Science (INRNE-BAS).

GAMMA-CLOUD: a computer code for calculating gamma-exposure due to a radioactive cloud released from a point source

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, O [Chugoku Electric Power Co. Inc., Hiroshima (Japan); Sawaguchi, Y; Kaneko, M

1979-03-01

A computer code, designated GAMMA-CLOUD, has been developed by specialists of electric power companies to meet requests from the companies to have a unified means of calculating annual external doses from routine releases of radioactive gaseous effluents from nuclear power plants, based on the Japan Atomic Energy Commission's guides for environmental dose evaluation. GAMMA-CLOUD is written in FORTRAN language and its required capacity is less than 100 kilobytes. The average ..gamma..-exposure at an observation point can be calculated within a few minutes with comparable precision to other existing codes.

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

International Nuclear Information System (INIS)

Vieira, Jose Wilson

2004-07-01

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)

Variations of dose distribution in high energy electron beams as a function of geometrical parameters of irradiation. Application to computer calculation

International Nuclear Information System (INIS)

Villeret, O.

1985-04-01

An algorithm is developed for the purpose of compter treatment planning of electron therapy. The method uses experimental absorbed dose distribution data in the irradiated medium for electron beams in the 8-20 MeV range delivered by the Sagittaire linear accelerator (study of central axis depth dose, beam profiles) in various geometrical conditions. Experimental verification of the computer program showed agreement with 2% between dose measurement and computer calculation [fr

ROBOT3: a computer program to calculate the in-pile three-dimensional bowing of cylindrical fuel rods (AWBA Development Program)

International Nuclear Information System (INIS)

Kovscek, S.E.; Martin, S.E.

1982-10-01

ROBOT3 is a FORTRAN computer program which is used in conjunction with the CYGRO5 computer program to calculate the time-dependent inelastic bowing of a fuel rod using an incremental finite element method. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the CYGRO5 axisymmetric model. Fuel rod supports are modeled as displacement, force, or spring-type nodal boundary conditions. The program input is described and a sample problem is given

STOSS - A computer module which can be used in Monte-Carlo-calculation for determining the path of a particle in a heterogeneous medium in three dimensions

International Nuclear Information System (INIS)

Sdouz, G.

1980-09-01

The computer program STOSS determines the path of a particle in a heterogenous medium in three dimensions. The program can be used as a module in Monte-Carlo-calculations. The collision can be transferred from the centre-of-mass system into a fixed cartesian coordinate-system by means of appropriate transformations. Then the path length is determined and the location of the next collision is calculated. The computational details are discussed at some length. (auth.)

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.

Detecting number processing and mental calculation in patients with disorders of consciousness using a hybrid brain-computer interface system.

Science.gov (United States)

Li, Yuanqing; Pan, Jiahui; He, Yanbin; Wang, Fei; Laureys, Steven; Xie, Qiuyou; Yu, Ronghao

2015-12-15

For patients with disorders of consciousness such as coma, a vegetative state or a minimally conscious state, one challenge is to detect and assess the residual cognitive functions in their brains. Number processing and mental calculation are important brain functions but are difficult to detect in patients with disorders of consciousness using motor response-based clinical assessment scales such as the Coma Recovery Scale-Revised due to the patients' motor impairments and inability to provide sufficient motor responses for number- and calculation-based communication. In this study, we presented a hybrid brain-computer interface that combines P300 and steady state visual evoked potentials to detect number processing and mental calculation in Han Chinese patients with disorders of consciousness. Eleven patients with disorders of consciousness who were in a vegetative state (n = 6) or in a minimally conscious state (n = 3) or who emerged from a minimally conscious state (n = 2) participated in the brain-computer interface-based experiment. During the experiment, the patients with disorders of consciousness were instructed to perform three tasks, i.e., number recognition, number comparison, and mental calculation, including addition and subtraction. In each experimental trial, an arithmetic problem was first presented. Next, two number buttons, only one of which was the correct answer to the problem, flickered at different frequencies to evoke steady state visual evoked potentials, while the frames of the two buttons flashed in a random order to evoke P300 potentials. The patients needed to focus on the target number button (the correct answer). Finally, the brain-computer interface system detected P300 and steady state visual evoked potentials to determine the button to which the patients attended, further presenting the results as feedback. Two of the six patients who were in a vegetative state, one of the three patients who were in a minimally conscious state, and

Efficient computer program EPAS-J1 for calculating stress intensity factors of three-dimensional surface cracks

International Nuclear Information System (INIS)

Miyazaki, Noriyuki; Watanabe, Takayuki; Yagawa, Genki.

1982-03-01

A finite element computer program EPAS-J1 was developed to calculate the stress intensity factors of three-dimensional cracks. In the program, the stress intensity factor is determined by the virtual crack extension method together with the distorted elements allocated along the crack front. This program also includes the connection elements based on the Lagrange multiplier concept to connect such different kinds of elements as the solid and shell elements, or the shell and beam elements. For the structure including three-dimensional surface cracks, the solid elements are employed only at the neighborhood of a surface crack, while the remainder of the structure is modeled by the shell or beam elements due to the reason that the crack singularity is very local. Computer storage and computational time can be highly reduced with the application of the above modeling technique for the calculation of the stress intensity factors of the three-dimensional surface cracks, because the three-dimensional solid elements are required only around the crack front. Several numerical analyses were performed by the EPAS-J1 program. At first, the accuracies of the connection element and the virtual crack extension method were confirmed using the simple structures. Compared with other techniques of connecting different kinds of elements such as the tying method or the method using anisotropic plate element, the present connection element is found to provide better results than the others. It is also found that the virtual crack extension method provides the accurate stress intensity factor. Furthermore, the results are also presented for the stress intensity factor analyses of cylinders with longitudinal or circumferential surface cracks using the combination of the various kinds of elements together with the connection elements. (author)

MRI-Based Computed Tomography Metal Artifact Correction Method for Improving Proton Range Calculation Accuracy

International Nuclear Information System (INIS)

Park, Peter C.; Schreibmann, Eduard; Roper, Justin; Elder, Eric; Crocker, Ian; Fox, Tim; Zhu, X. Ronald; Dong, Lei; Dhabaan, Anees

2015-01-01

Purpose: Computed tomography (CT) artifacts can severely degrade dose calculation accuracy in proton therapy. Prompted by the recently increased popularity of magnetic resonance imaging (MRI) in the radiation therapy clinic, we developed an MRI-based CT artifact correction method for improving the accuracy of proton range calculations. Methods and Materials: The proposed method replaces corrupted CT data by mapping CT Hounsfield units (HU number) from a nearby artifact-free slice, using a coregistered MRI. MRI and CT volumetric images were registered with use of 3-dimensional (3D) deformable image registration (DIR). The registration was fine-tuned on a slice-by-slice basis by using 2D DIR. Based on the intensity of paired MRI pixel values and HU from an artifact-free slice, we performed a comprehensive analysis to predict the correct HU for the corrupted region. For a proof-of-concept validation, metal artifacts were simulated on a reference data set. Proton range was calculated using reference, artifactual, and corrected images to quantify the reduction in proton range error. The correction method was applied to 4 unique clinical cases. Results: The correction method resulted in substantial artifact reduction, both quantitatively and qualitatively. On respective simulated brain and head and neck CT images, the mean error was reduced from 495 and 370 HU to 108 and 92 HU after correction. Correspondingly, the absolute mean proton range errors of 2.4 cm and 1.7 cm were reduced to less than 2 mm in both cases. Conclusions: Our MRI-based CT artifact correction method can improve CT image quality and proton range calculation accuracy for patients with severe CT artifacts

Dose calculation for electrons

International Nuclear Information System (INIS)

Hirayama, Hideo

1995-01-01

The joint working group of ICRP/ICRU is advancing the works of reviewing the ICRP publication 51 by investigating the data related to radiation protection. In order to introduce the 1990 recommendation, it has been demanded to carry out calculation for neutrons, photons and electrons. As for electrons, EURADOS WG4 (Numerical Dosimetry) rearranged the data to be calculated at the meeting held in PTB Braunschweig in June, 1992, and the question and request were presented by Dr. J.L. Chartier, the responsible person, to the researchers who are likely to undertake electron transport Monte Carlo calculation. The author also has carried out the requested calculation as it was the good chance to do the mutual comparison among various computation codes regarding electron transport calculation. The content that the WG requested to calculate was the absorbed dose at depth d mm when parallel electron beam enters at angle α into flat plate phantoms of PMMA, water and ICRU4-element tissue, which were placed in vacuum. The calculation was carried out by the versatile electron-photon shower computation Monte Carlo code, EGS4. As the results, depth dose curves and the dependence of absorbed dose on electron energy, incident angle and material are reported. The subjects to be investigated are pointed out. (K.I.)

Monte Carlo Calculated Effective Dose to Teenage Girls from Computed Tomography Examinations

International Nuclear Information System (INIS)

Caon, M.; Bibbo, G.; Pattison, J.

2000-01-01

Effective doses from CT to paediatric patients are not common in the literature. This article reports some effective doses to teenage girls from CT examinations. The voxel computational model ADELAIDE, representative of a 14-year-old girl, was scaled in size by ±5% to represent also 11-12-year-old and 16-year-old girls. The EGS4 Monte Carlo code was used to calculate the effective dose from chest, abdomen and whole torso CT examinations to the three version of ADELAIDE using a 120 kV spectrum. For the whole torso CT examination, in order of increasing model size, the effective doses were 9.0, 8.2 and 7.8 mSv per 100 mA.s. Data are presented that allow the estimation of effective dose from CT examinations of the torso for girls between the ages of 11 and 16. (author)

Efficient Finite Element Calculation of Nγ

DEFF Research Database (Denmark)

Clausen, Johan; Damkilde, Lars; Krabbenhøft, K.

2007-01-01

This paper deals with the computational aspects of the Mohr-Coulomb material model, in particular the calculation of the bearing capacity factor Nγfor a strip and a circular footing.......This paper deals with the computational aspects of the Mohr-Coulomb material model, in particular the calculation of the bearing capacity factor Nγfor a strip and a circular footing....

Probability calculations for three-part mineral resource assessments

Science.gov (United States)

Ellefsen, Karl J.

2017-06-27

Three-part mineral resource assessment is a methodology for predicting, in a specified geographic region, both the number of undiscovered mineral deposits and the amount of mineral resources in those deposits. These predictions are based on probability calculations that are performed with computer software that is newly implemented. Compared to the previous implementation, the new implementation includes new features for the probability calculations themselves and for checks of those calculations. The development of the new implementation lead to a new understanding of the probability calculations, namely the assumptions inherent in the probability calculations. Several assumptions strongly affect the mineral resource predictions, so it is crucial that they are checked during an assessment. The evaluation of the new implementation leads to new findings about the probability calculations,namely findings regarding the precision of the computations,the computation time, and the sensitivity of the calculation results to the input.

Calculating three loop ladder and V-topologies for massive operator matrix elements by computer algebra

International Nuclear Information System (INIS)

Ablinger, J.; Schneider, C.; Manteuffel, A. von

2015-09-01

Three loop ladder and V-topology diagrams contributing to the massive operator matrix element A Qg are calculated. The corresponding objects can all be expressed in terms of nested sums and recurrences depending on the Mellin variable N and the dimensional parameter ε. Given these representations, the desired Laurent series expansions in ε can be obtained with the help of our computer algebra toolbox. Here we rely on generalized hypergeometric functions and Mellin-Barnes representations, on difference ring algorithms for symbolic summation, on an optimized version of the multivariate Almkvist-Zeilberger algorithm for symbolic integration, and on new methods to calculate Laurent series solutions of coupled systems of differential equations. The solutions can be computed for general coefficient matrices directly for any basis also performing the expansion in the dimensional parameter in case it is expressible in terms of indefinite nested product-sum expressions. This structural result is based on new results of our difference ring theory. In the cases discussed we deal with iterative sum- and integral-solutions over general alphabets. The final results are expressed in terms of special sums, forming quasi-shuffle algebras, such as nested harmonic sums, generalized harmonic sums, and nested binomially weighted (cyclotomic) sums. Analytic continuations to complex values of N are possible through the recursion relations obeyed by these quantities and their analytic asymptotic expansions. The latter lead to a host of new constants beyond the multiple zeta values, the infinite generalized harmonic and cyclotomic sums in the case of V-topologies.

Calculating three loop ladder and V-topologies for massive operator matrix elements by computer algebra

Science.gov (United States)

Ablinger, J.; Behring, A.; Blümlein, J.; De Freitas, A.; von Manteuffel, A.; Schneider, C.

2016-05-01

Three loop ladder and V-topology diagrams contributing to the massive operator matrix element AQg are calculated. The corresponding objects can all be expressed in terms of nested sums and recurrences depending on the Mellin variable N and the dimensional parameter ε. Given these representations, the desired Laurent series expansions in ε can be obtained with the help of our computer algebra toolbox. Here we rely on generalized hypergeometric functions and Mellin-Barnes representations, on difference ring algorithms for symbolic summation, on an optimized version of the multivariate Almkvist-Zeilberger algorithm for symbolic integration, and on new methods to calculate Laurent series solutions of coupled systems of differential equations. The solutions can be computed for general coefficient matrices directly for any basis also performing the expansion in the dimensional parameter in case it is expressible in terms of indefinite nested product-sum expressions. This structural result is based on new results of our difference ring theory. In the cases discussed we deal with iterative sum- and integral-solutions over general alphabets. The final results are expressed in terms of special sums, forming quasi-shuffle algebras, such as nested harmonic sums, generalized harmonic sums, and nested binomially weighted (cyclotomic) sums. Analytic continuations to complex values of N are possible through the recursion relations obeyed by these quantities and their analytic asymptotic expansions. The latter lead to a host of new constants beyond the multiple zeta values, the infinite generalized harmonic and cyclotomic sums in the case of V-topologies.

The digital computer

CERN Document Server

Parton, K C

2014-01-01

The Digital Computer focuses on the principles, methodologies, and applications of the digital computer. The publication takes a look at the basic concepts involved in using a digital computer, simple autocode examples, and examples of working advanced design programs. Discussions focus on transformer design synthesis program, machine design analysis program, solution of standard quadratic equations, harmonic analysis, elementary wage calculation, and scientific calculations. The manuscript then examines commercial and automatic programming, how computers work, and the components of a computer

Calculation of local skin doses with ICRP adult mesh-type reference computational phantoms

Science.gov (United States)

Yeom, Yeon Soo; Han, Haegin; Choi, Chansoo; Nguyen, Thang Tat; Lee, Hanjin; Shin, Bangho; Kim, Chan Hyeong; Han, Min Cheol

2018-01-01

Recently, Task Group 103 of the International Commission on Radiological Protection (ICRP) developed new mesh-type reference computational phantoms (MRCPs) for adult males and females in order to address the limitations of the current voxel-type reference phantoms described in ICRP Publication 110 due to their limited voxel resolutions and the nature of the voxel geometry. One of the substantial advantages of the MRCPs over the ICRP-110 reference phantoms is the inclusion of a 50-μm-thick radiosensitive skin basal-cell layer; however, a methodology for calculating the local skin dose (LSD), i.e., the maximum dose to the basal layer averaged over a 1-cm2 area, has yet to be developed. In the present study, a dedicated program for the LSD calculation with the MRCPs was developed based on the mean shift algorithm and the Geant4 Monte Carlo code. The developed program was used to calculate local skin dose coefficients (LSDCs) for electrons and alpha particles, which were then compared with the values given in ICRP Publication 116 that were produced with a simple tissue-equivalent cube model. The results of the present study show that the LSDCs of the MRCPs are generally in good agreement with the ICRP-116 values for alpha particles, but for electrons, significant differences are found at energies higher than 0.15 MeV. The LSDCs of the MRCPs are greater than the ICRP-116 values by as much as 2.7 times at 10 MeV, which is due mainly to the different curvature between realistic MRCPs ( i.e., curved) and the simple cube model ( i.e., flat).

Development of a computer model using the EGS4 simulation code to calculate scattered X-rays through some materials

International Nuclear Information System (INIS)

Al-Ghorabie, F.H.H.

2003-01-01

In this paper a computer model based on the use of the well-known Monte Carlo simulation code EGS4 was developed to simulate the scattering of polyenergetic X-ray beams through some materials. These materials are: lucite, polyethylene, polypropylene and aluminium. In particular, the ratio of the scattered to total X-ray fluence (scatter fraction) has been calculated for X-ray beams in the energy region 30-120 keV. In addition scatter fractions have been determined experimentally using a polyenergetic superficial X-ray unit. Comparison of the measured and the calculated results has been performed. The Monte Carlo calculations have also been carried out for water, bakelite and bone to examine the dependence of scatter fraction on the density of the scatterer. Good agreement (estimated statistical error < 5%) was obtained between the measured and the calculated values of the scatter fractions for materials with Z < 20 that were studied in this paper. Copyright (2003) Australasian College of Physical Scientists and Engineers in Medicine

GRUCAL, a computer program for calculating macroscopic group constants

International Nuclear Information System (INIS)

Woll, D.

1975-06-01

Nuclear reactor calculations require material- and composition-dependent, energy averaged nuclear data to describe the interaction of neutrons with individual isotopes in material compositions of reactor zones. The code GRUCAL calculates these macroscopic group constants for given compositions from the material-dependent data of the group constant library GRUBA. The instructions for calculating group constants are not fixed in the program, but will be read at the actual execution time from a separate instruction file. This allows to accomodate GRUCAL to various problems or different group constant concepts. (orig.) [de

ACDOS1: a computer code to calculate dose rates from neutron activation of neutral beamlines and other fusion-reactor components

International Nuclear Information System (INIS)

Keney, G.S.

1981-08-01

A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV

FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup

International Nuclear Information System (INIS)

Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

1997-12-01

FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes' integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6)

Parallel computer calculation of quantum spin lattices

International Nuclear Information System (INIS)

Lamarcq, J.

1998-01-01

Numerical simulation allows the theorists to convince themselves about the validity of the models they use. Particularly by simulating the spin lattices one can judge about the validity of a conjecture. Simulating a system defined by a large number of degrees of freedom requires highly sophisticated machines. This study deals with modelling the magnetic interactions between the ions of a crystal. Many exact results have been found for spin 1/2 systems but not for systems of other spins for which many simulation have been carried out. The interest for simulations has been renewed by the Haldane's conjecture stipulating the existence of a energy gap between the ground state and the first excited states of a spin 1 lattice. The existence of this gap has been experimentally demonstrated. This report contains the following four chapters: 1. Spin systems; 2. Calculation of eigenvalues; 3. Programming; 4. Parallel calculation

TISCON, a BASIC computer program for the calculation of the biodistribution of radionuclide-labelled drugs in rats and mice

International Nuclear Information System (INIS)

Maddalena, D.J.

1983-09-01

Animal biodistribution studies on radionuclide-labelled drugs are labour-intensive and time-consuming. A method for rapidly carrying out these studies on rats and mice is presented. An interactive computer program, written in BASIC, is used to calculate parameters of interest, such as per cent injected dose (%ID),%ID per gram and target to non-target ratios

Numerical Nuclear Second Derivatives on a Computing Grid: Enabling and Accelerating Frequency Calculations on Complex Molecular Systems.

Science.gov (United States)

Yang, Tzuhsiung; Berry, John F

2018-06-04

The computation of nuclear second derivatives of energy, or the nuclear Hessian, is an essential routine in quantum chemical investigations of ground and transition states, thermodynamic calculations, and molecular vibrations. Analytic nuclear Hessian computations require the resolution of costly coupled-perturbed self-consistent field (CP-SCF) equations, while numerical differentiation of analytic first derivatives has an unfavorable 6 N ( N = number of atoms) prefactor. Herein, we present a new method in which grid computing is used to accelerate and/or enable the evaluation of the nuclear Hessian via numerical differentiation: NUMFREQ@Grid. Nuclear Hessians were successfully evaluated by NUMFREQ@Grid at the DFT level as well as using RIJCOSX-ZORA-MP2 or RIJCOSX-ZORA-B2PLYP for a set of linear polyacenes with systematically increasing size. For the larger members of this group, NUMFREQ@Grid was found to outperform the wall clock time of analytic Hessian evaluation; at the MP2 or B2LYP levels, these Hessians cannot even be evaluated analytically. We also evaluated a 156-atom catalytically relevant open-shell transition metal complex and found that NUMFREQ@Grid is faster (7.7 times shorter wall clock time) and less demanding (4.4 times less memory requirement) than an analytic Hessian. Capitalizing on the capabilities of parallel grid computing, NUMFREQ@Grid can outperform analytic methods in terms of wall time, memory requirements, and treatable system size. The NUMFREQ@Grid method presented herein demonstrates how grid computing can be used to facilitate embarrassingly parallel computational procedures and is a pioneer for future implementations.

Methods and experimental coefficients used in the computation of reactor shielding; Methodes et coefficients experimentaux pour le calcul des protections de reacteurs

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J; Lafore, P; Millot, J P; Rastoin, J; Vathaire, F de [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

1) The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2) Starting from this concept, we endeavoured to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3) Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author) [French] 1) La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2) A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3) Enfin une methode de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

Methods and experimental coefficients used in the computation of reactor shielding; Methodes et coefficients experimentaux pour le calcul des protections de reacteurs

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J; Lafore, P; Millot, J P; Rastoin, J; Vathaire, F de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

1. The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2. Starting from this concept, we endeavored to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3. Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author)Fren. [French] 1. La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2. A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3. Enfin une mde de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

Development of hybrid core calculation system using 2-D full-core heterogeneous transport calculation and 3-D advanced nodal calculation

International Nuclear Information System (INIS)

Sugimura, Naoki; Mori, Masaaki; Hijiya, Masayuki; Ushio, Tadashi; Arakawa, Yasushi

2004-01-01

This paper presents the Hybrid Core Calculation System which is a very rigorous but a practical calculation system applicable to best estimate core design calculations taking advantage of the recent remarkable progress of computers. The basic idea of this system is to generate the correction factors for assembly homogenized cross sections, discontinuity factors, etc. by comparing the CASMO-4 and SIMULATE-3 2-D core calculation results under the consistent calculation condition and then apply them for SIMULATE-3 3-D calculation. The CASMO-4 2-D heterogeneous core calculation is performed for each depletion step with the core conditions previously determined by ordinary SIMULATE-3 core calculation to avoid time consuming iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. The final SIMULATE-3 3-D calculation using the correction factors is performed with iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. (author)

Hybrid reduced order modeling for assembly calculations

Energy Technology Data Exchange (ETDEWEB)

Bang, Youngsuk, E-mail: ysbang00@fnctech.com [FNC Technology, Co. Ltd., Yongin-si (Korea, Republic of); Abdel-Khalik, Hany S., E-mail: abdelkhalik@purdue.edu [Purdue University, West Lafayette, IN (United States); Jessee, Matthew A., E-mail: jesseema@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mertyurek, Ugur, E-mail: mertyurek@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

2015-12-15

Highlights: • Reducing computational cost in engineering calculations. • Reduced order modeling algorithm for multi-physics problem like assembly calculation. • Non-intrusive algorithm with random sampling. • Pattern recognition in the components with high sensitive and large variation. - Abstract: While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system.

Hybrid reduced order modeling for assembly calculations

International Nuclear Information System (INIS)

Bang, Youngsuk; Abdel-Khalik, Hany S.; Jessee, Matthew A.; Mertyurek, Ugur

2015-01-01

Highlights: • Reducing computational cost in engineering calculations. • Reduced order modeling algorithm for multi-physics problem like assembly calculation. • Non-intrusive algorithm with random sampling. • Pattern recognition in the components with high sensitive and large variation. - Abstract: While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system.

Learning to Calculate and Learning Mathematics.

Science.gov (United States)

Fearnley-Sander, Desmond

1980-01-01

A calculator solution of a simple computational problem is discussed with emphasis on its ramifications for the understanding of some fundamental theorems of pure mathematics and techniques of computing. (Author/MK)

Methodology of shielding calculation for nuclear reactors

International Nuclear Information System (INIS)

Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo

1982-01-01

A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt

User's guide to SERICPAC: A computer program for calculating electric-utility avoided costs rates

Energy Technology Data Exchange (ETDEWEB)

Wirtshafter, R.; Abrash, M.; Koved, M.; Feldman, S.

1982-05-01

SERICPAC is a computer program developed to calculate average avoided cost rates for decentralized power producers and cogenerators that sell electricity to electric utilities. SERICPAC works in tandem with SERICOST, a program to calculate avoided costs, and determines the appropriate rates for buying and selling of electricity from electric utilities to qualifying facilities (QF) as stipulated under Section 210 of PURA. SERICPAC contains simulation models for eight technologies including wind, hydro, biogas, and cogeneration. The simulations are converted in a diversified utility production which can be either gross production or net production, which accounts for an internal electricity usage by the QF. The program allows for adjustments to the production to be made for scheduled and forced outages. The final output of the model is a technology-specific average annual rate. The report contains a description of the technologies and the simulations as well as complete user's guide to SERICPAC.

Second reference calculation for the WIPP

International Nuclear Information System (INIS)

Branstetter, L.J.

1985-03-01

Results of the second reference calculation for the Waste Isolation Pilot Plant (WIPP) project using the dynamic relaxation finite element code SANCHO are presented. This reference calculation is intended to predict the response of a typical panel of excavated rooms designed for storage of nonheat-producing nuclear waste. Results are presented that include relevant deformations, relative clay seam displacements, and stress and strain profiles. This calculation is a particular solution obtained by a computer code, which has proven analytic capabilities when compared with other structural finite element codes. It is hoped that the results presented here will be useful in providing scoping values for defining experiments and for developing instrumentation. It is also hoped that the calculation will be useful as part of an exercise in developing a methodology for performing important design calculations by more than one analyst using more than one computer code, and for defining internal Quality Assurance (QA) procedures for such calculations. 27 refs., 15 figs

Program POD; A computer code to calculate nuclear elastic scattering cross sections with the optical model and neutron inelastic scattering cross sections by the distorted-wave born approximation

International Nuclear Information System (INIS)

Ichihara, Akira; Kunieda, Satoshi; Chiba, Satoshi; Iwamoto, Osamu; Shibata, Keiichi; Nakagawa, Tsuneo; Fukahori, Tokio; Katakura, Jun-ichi

2005-07-01

The computer code, POD, was developed to calculate angle-differential cross sections and analyzing powers for shape-elastic scattering for collisions of neutron or light ions with target nucleus. The cross sections are computed with the optical model. Angle-differential cross sections for neutron inelastic scattering can also be calculated with the distorted-wave Born approximation. The optical model potential parameters are the most essential inputs for those model computations. In this program, the cross sections and analyzing powers are obtained by using the existing local or global parameters. The parameters can also be inputted by users. In this report, the theoretical formulas, the computational methods, and the input parameters are explained. The sample inputs and outputs are also presented. (author)

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.

Reactor perturbation calculations by Monte Carlo methods

International Nuclear Information System (INIS)

Gubbins, M.E.

1965-09-01

Whilst Monte Carlo methods are useful for reactor calculations involving complicated geometry, it is difficult to apply them to the calculation of perturbation worths because of the large amount of computing time needed to obtain good accuracy. Various ways of overcoming these difficulties are investigated in this report, with the problem of estimating absorbing control rod worths particularly in mind. As a basis for discussion a method of carrying out multigroup reactor calculations by Monte Carlo methods is described. Two methods of estimating a perturbation worth directly, without differencing two quantities of like magnitude, are examined closely but are passed over in favour of a third method based on a correlation technique. This correlation method is described, and demonstrated by a limited range of calculations for absorbing control rods in a fast reactor. In these calculations control rod worths of between 1% and 7% in reactivity are estimated to an accuracy better than 10% (3 standard errors) in about one hour's computing time on the English Electric KDF.9 digital computer. (author)

3-D calculations for comparison with the experiments

Energy Technology Data Exchange (ETDEWEB)

Alrsen, A M; Bosser, R

1973-09-27

In order to analyse the axial power profile measurements an attempt has been made to do full 3-D calculations for the Dragon reactor. The calculations are still at a very early stage, but the methods used will be outlined here together with the plans for investigations to be carried out in the near future. Some preliminary-results are reported as no final results have yet been obtained. 3-D calculations are rather expensive because of the computer time consumption. It is therefore essential, before too many big computer jobs are spent, to find approximations which can save calculation time. On the other hand some savings, for instance in the number of mesh points, may cause totally wrong results. The ''proper'' calculations have therefore to be proceeded by a number of preliminary investigations, to ensure optimum accuracy and computer expenses. This report contains some of these preliminary studies.

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

International Nuclear Information System (INIS)

Carassiti, F.; Liuzzo, G.; Morelli, A.

1982-01-01

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)

Unsteady State Two Phase Flow Pressure Drop Calculations

OpenAIRE

Ayatollahi, Shahaboddin

1992-01-01

A method is presented to calculate unsteady state two phase flow in a gas-liquid line based on a quasi-steady state approach. A computer program for numerical solution of this method was prepared. Results of calculations using the computer program are presented for several unsteady state two phase flow systems

PHYSICOCHEMICAL PROPERTY CALCULATIONS

Science.gov (United States)

Computer models have been developed to estimate a wide range of physical-chemical properties from molecular structure. The SPARC modeling system approaches calculations as site specific reactions (pKa, hydrolysis, hydration) and `whole molecule' properties (vapor pressure, boilin...

Radioimmunoassay evaluation and quality control by use of a simple computer program for a low cost desk top calculator

International Nuclear Information System (INIS)

Schwarz, S.

1980-01-01

A simple computer program for the data processing and quality control of radioimmunoassays is presented. It is written for low cost programmable desk top calculator (Hewlett Packard 97), which can be afforded by smaller laboratories. The untreated counts from the scintillation spectrometer are entered manually; the printout gives the following results: initial data, logit-log transformed calibration points, parameters of goodness of fit and of the position of the standard curve, control and unknown samples dose estimates (mean value from single dose interpolations and scatter of replicates) together with the automatic calculation of within assay variance and, by use of magnetic cards holding the control parameters of all previous assays, between assay variance. (orig.) [de

Magnetic Field Grid Calculator

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The Magnetic Field Properties Calculator will computes the estimated values of Earth's magnetic field(declination, inclination, vertical component, northerly...

Molecular calculations with B functions

International Nuclear Information System (INIS)

Steinborn, E.O.; Homeier, H.H.H.; Ema, I.; Lopez, R.; Ramirez, G.

2000-01-01

A program for molecular calculations with B functions is reported and its performance is analyzed. All the one- and two-center integrals and the three-center nuclear attraction integrals are computed by direct procedures, using previously developed algorithms. The three- and four-center electron repulsion integrals are computed by means of Gaussian expansions of the B functions. A new procedure for obtaining these expansions is also reported. Some results on full molecular calculations are included to show the capabilities of the program and the quality of the B functions to represent the electronic functions in molecules

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)

Leal, L.C.; de Saussure, G.; Perez, R.B.

1990-01-01

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

Calculation of Monte Carlo importance functions for use in nuclear-well logging calculations

International Nuclear Information System (INIS)

Soran, P.D.; McKeon, D.C.; Booth, T.E.

1989-07-01

Importance sampling is essential to the timely solution of Monte Carlo nuclear-logging computer simulations. Achieving minimum variance (maximum precision) of a response in minimum computation time is one criteria for the choice of an importance function. Various methods for calculating importance functions will be presented, new methods investigated, and comparisons with porosity and density tools will be shown. 5 refs., 1 tab

Radioactive cloud dose calculations

International Nuclear Information System (INIS)

Healy, J.W.

1984-01-01

Radiological dosage principles, as well as methods for calculating external and internal dose rates, following dispersion and deposition of radioactive materials in the atmosphere are described. Emphasis has been placed on analytical solutions that are appropriate for hand calculations. In addition, the methods for calculating dose rates from ingestion are discussed. A brief description of several computer programs are included for information on radionuclides. There has been no attempt to be comprehensive, and only a sampling of programs has been selected to illustrate the variety available

Computational Calorimetry: High-Precision Calculation of Host–Guest Binding Thermodynamics

Science.gov (United States)

2015-01-01

We present a strategy for carrying out high-precision calculations of binding free energy and binding enthalpy values from molecular dynamics simulations with explicit solvent. The approach is used to calculate the thermodynamic profiles for binding of nine small molecule guests to either the cucurbit[7]uril (CB7) or β-cyclodextrin (βCD) host. For these systems, calculations using commodity hardware can yield binding free energy and binding enthalpy values with a precision of ∼0.5 kcal/mol (95% CI) in a matter of days. Crucially, the self-consistency of the approach is established by calculating the binding enthalpy directly, via end point potential energy calculations, and indirectly, via the temperature dependence of the binding free energy, i.e., by the van’t Hoff equation. Excellent agreement between the direct and van’t Hoff methods is demonstrated for both host–guest systems and an ion-pair model system for which particularly well-converged results are attainable. Additionally, we find that hydrogen mass repartitioning allows marked acceleration of the calculations with no discernible cost in precision or accuracy. Finally, we provide guidance for accurately assessing numerical uncertainty of the results in settings where complex correlations in the time series can pose challenges to statistical analysis. The routine nature and high precision of these binding calculations opens the possibility of including measured binding thermodynamics as target data in force field optimization so that simulations may be used to reliably interpret experimental data and guide molecular design. PMID:26523125

Subjective Organization Calculator for Free Recall

Directory of Open Access Journals (Sweden)

Olesya Senkova

2015-11-01

Full Text Available The free recall measure has an advantage over other memory measures because the free recall measure can provide organization measures, which can reveal the strategies participants used to maximize recall. For instance, even when a study list does not show a clear organizational scheme, recall outputs are often far from random, evidenced by participants recalling the same two or more items together repeatedly across multiple test trials. Unfortunately, computing organizational measures is laborious. The present article introduces a calculator to compute subjective organization (SO measures. The calculator is based on a popular platform accessible to most researchers and is designed to compute commonly used SO measures for each participant.

Recommendations for computer code selection of a flow and transport code to be used in undisturbed vadose zone calculations for TWRS immobilized wastes environmental analyses

International Nuclear Information System (INIS)

VOOGD, J.A.

1999-01-01

An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis

User's Guide to Handlens - A Computer Program that Calculates the Chemistry of Minerals in Mixtures

Science.gov (United States)

Eberl, D.D.

2008-01-01

HandLens is a computer program, written in Excel macro language, that calculates the chemistry of minerals in mineral mixtures (for example, in rocks, soils and sediments) for related samples from inputs of quantitative mineralogy and chemistry. For best results, the related samples should contain minerals having the same chemical compositions; that is, the samples should differ only in the proportions of minerals present. This manual describes how to use the program, discusses the theory behind its operation, and presents test results of the program's accuracy. Required input for HandLens includes quantitative mineralogical data, obtained, for example, by RockJock analysis of X-ray diffraction (XRD) patterns, and quantitative chemical data, obtained, for example, by X-ray florescence (XRF) analysis of the same samples. Other quantitative data, such as sample depth, temperature, surface area, also can be entered. The minerals present in the samples are selected from a list, and the program is started. The results of the calculation include: (1) a table of linear coefficients of determination (r2's) which relate pairs of input data (for example, Si versus quartz weight percents); (2) a utility for plotting all input data, either as pairs of variables, or as sums of up to eight variables; (3) a table that presents the calculated chemical formulae for minerals in the samples; (4) a table that lists the calculated concentrations of major, minor, and trace elements in the various minerals; and (5) a table that presents chemical formulae for the minerals that have been corrected for possible systematic errors in the mineralogical and/or chemical analyses. In addition, the program contains a method for testing the assumption of constant chemistry of the minerals within a sample set.

Calculator calculus

CERN Document Server

McCarty, George

1982-01-01

How THIS BOOK DIFFERS This book is about the calculus. What distinguishes it, however, from other books is that it uses the pocket calculator to illustrate the theory. A computation that requires hours of labor when done by hand with tables is quite inappropriate as an example or exercise in a beginning calculus course. But that same computation can become a delicate illustration of the theory when the student does it in seconds on his calculator. t Furthermore, the student's own personal involvement and easy accomplishment give hi~ reassurance and en­ couragement. The machine is like a microscope, and its magnification is a hundred millionfold. We shall be interested in limits, and no stage of numerical approximation proves anything about the limit. However, the derivative of fex) = 67.SgX, for instance, acquires real meaning when a student first appreciates its values as numbers, as limits of 10 100 1000 t A quick example is 1.1 , 1.01 , 1.001 , •••• Another example is t = 0.1, 0.01, in the functio...

Computer calculation of Witten's 3-manifold invariant

International Nuclear Information System (INIS)

Freed, D.S.; Gompf, R.E.

1991-01-01

Witten's 2+1 dimensional Chern-Simons theory is exactly solvable. We compute the partition function, a topological invariant of 3-manifolds, on generalized Seifert spaces. Thus we test the path integral using the theory of 3-manifolds. In particular, we compare the exact solution with the asymptotic formula predicted by perturbation theory. We conclude that this path integral works as advertised and gives an effective topological invariant. (orig.)

Ensuring the validity of calculated subcritical limits

International Nuclear Information System (INIS)

Clark, H.K.

1977-01-01

The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionally subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin

Calculation of Rydberg interaction potentials

DEFF Research Database (Denmark)

Weber, Sebastian; Tresp, Christoph; Menke, Henri

2017-01-01

for calculating the required electric multipole moments and the inclusion of electromagnetic fields with arbitrary direction. We focus specifically on symmetry arguments and selection rules, which greatly reduce the size of the Hamiltonian matrix, enabling the direct diagonalization of the Hamiltonian up...... to higher multipole orders on a desktop computer. Finally, we present example calculations showing the relevance of the full interaction calculation to current experiments. Our software for calculating Rydberg potentials including all features discussed in this tutorial is available as open source....

Description and application of the EAP computer program for calculating life-cycle energy use and greenhouse gas emissions of household consumption items

NARCIS (Netherlands)

Benders, R.M.J.; Wilting, H.C.; Kramer, K.J.; Moll, H.C.

2001-01-01

Focusing on reduction in energy use and greenhouse gas emissions, a life-cycle-based analysis tool has been developed. The energy analysis program (EAP) is a computer program for determining energy use and greenhouse gas emissions related to household consumption items, using a hybrid calculation

Good Practices in Free-energy Calculations

Science.gov (United States)

Pohorille, Andrew; Jarzynski, Christopher; Chipot, Christopher

2013-01-01

As access to computational resources continues to increase, free-energy calculations have emerged as a powerful tool that can play a predictive role in drug design. Yet, in a number of instances, the reliability of these calculations can be improved significantly if a number of precepts, or good practices are followed. For the most part, the theory upon which these good practices rely has been known for many years, but often overlooked, or simply ignored. In other cases, the theoretical developments are too recent for their potential to be fully grasped and merged into popular platforms for the computation of free-energy differences. The current best practices for carrying out free-energy calculations will be reviewed demonstrating that, at little to no additional cost, free-energy estimates could be markedly improved and bounded by meaningful error estimates. In energy perturbation and nonequilibrium work methods, monitoring the probability distributions that underlie the transformation between the states of interest, performing the calculation bidirectionally, stratifying the reaction pathway and choosing the most appropriate paradigms and algorithms for transforming between states offer significant gains in both accuracy and precision. In thermodynamic integration and probability distribution (histogramming) methods, properly designed adaptive techniques yield nearly uniform sampling of the relevant degrees of freedom and, by doing so, could markedly improve efficiency and accuracy of free energy calculations without incurring any additional computational expense.

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)

Leal, L.C.; de Saussure, G.; Perez, R.B.

1989-01-01

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

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

1993-02-01

This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

International Nuclear Information System (INIS)

Yuan, Y.C.; Chen, S.Y.; LePoire, D.J.

1993-02-01

This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors

Benchmark neutron porosity log calculations

International Nuclear Information System (INIS)

Little, R.C.; Michael, M.; Verghese, K.; Gardner, R.P.

1989-01-01

Calculations have been made for a benchmark neutron porosity log problem with the general purpose Monte Carlo code MCNP and the specific purpose Monte Carlo code McDNL. For accuracy and timing comparison purposes the CRAY XMP and MicroVax II computers have been used with these codes. The CRAY has been used for an analog version of the MCNP code while the MicroVax II has been used for the optimized variance reduction versions of both codes. Results indicate that the two codes give the same results within calculated standard deviations. Comparisons are given and discussed for accuracy (precision) and computation times for the two codes

Quantum Computing's Classical Problem, Classical Computing's Quantum Problem

OpenAIRE

Van Meter, Rodney

2013-01-01

Tasked with the challenge to build better and better computers, quantum computing and classical computing face the same conundrum: the success of classical computing systems. Small quantum computing systems have been demonstrated, and intermediate-scale systems are on the horizon, capable of calculating numeric results or simulating physical systems far beyond what humans can do by hand. However, to be commercially viable, they must surpass what our wildly successful, highly advanced classica...

SU-E-T-531: Performance Evaluation of Multithreaded Geant4 for Proton Therapy Dose Calculations in a High Performance Computing Facility

International Nuclear Information System (INIS)

Shin, J; Coss, D; McMurry, J; Farr, J; Faddegon, B

2014-01-01

Purpose: To evaluate the efficiency of multithreaded Geant4 (Geant4-MT, version 10.0) for proton Monte Carlo dose calculations using a high performance computing facility. Methods: Geant4-MT was used to calculate 3D dose distributions in 1×1×1 mm3 voxels in a water phantom and patient's head with a 150 MeV proton beam covering approximately 5×5 cm2 in the water phantom. Three timestamps were measured on the fly to separately analyze the required time for initialization (which cannot be parallelized), processing time of individual threads, and completion time. Scalability of averaged processing time per thread was calculated as a function of thread number (1, 100, 150, and 200) for both 1M and 50 M histories. The total memory usage was recorded. Results: Simulations with 50 M histories were fastest with 100 threads, taking approximately 1.3 hours and 6 hours for the water phantom and the CT data, respectively with better than 1.0 % statistical uncertainty. The calculations show 1/N scalability in the event loops for both cases. The gains from parallel calculations started to decrease with 150 threads. The memory usage increases linearly with number of threads. No critical failures were observed during the simulations. Conclusion: Multithreading in Geant4-MT decreased simulation time in proton dose distribution calculations by a factor of 64 and 54 at a near optimal 100 threads for water phantom and patient's data respectively. Further simulations will be done to determine the efficiency at the optimal thread number. Considering the trend of computer architecture development, utilizing Geant4-MT for radiotherapy simulations is an excellent cost-effective alternative for a distributed batch queuing system. However, because the scalability depends highly on simulation details, i.e., the ratio of the processing time of one event versus waiting time to access for the shared event queue, a performance evaluation as described is recommended

GASCON and MHDGAS: FORTRAN IV computer codes for calculating gas and condensed-phase compositions in the coal-fired open-cycle MHD system

Energy Technology Data Exchange (ETDEWEB)

Blackburn, P E

1977-12-01

Fortran IV computer codes have been written to calculate the equilibrium partial pressures of the gaseous phase and the quantity and composition of the condensed phases in the open-cycle MHD system. The codes are based on temperature-dependent equilibrium constants, mass conservation, the mass action law, and assumed ideal solution of compounds in each of two condensed phases. It is assumed that the phases are an oxide-silicate phase and a sulfate-carbonate-hydroxide phase. Calculations are iterated for gas and condensate concentrations while increasing or decreasing the total moles of elements, but keeping mole ratios constant, to achieve the desired total pressure. During iteration the oxygen partial pressure is incrementally changed. The decision to increase or decrease the oxygen pressure in this process depends on comparison of the oxygen content calculated in the gas and condensate phases with the initial amount of oxygen in the ash, coal, seed, and air. This process, together with a normalization step, allows the elements to converge to their initial quantities. Two versions of the computer code have been written. GASCON calculates the equilibrium gas partial pressures and the quantity and composition of the condensed phases in steps of thirteen temperature and pressure combinations in which the condensate is removed after each step, simulating continuous slag removal from the MHD system. MHDGAS retains the condensate for each step, simulating flow of condensate (and gas) through the MHD system.

Acquisition of data from on-line laser turbidimeter and calculation of some kinetic variables in computer-coupled automated fed-batch culture

International Nuclear Information System (INIS)

Kadotani, Y.; Miyamoto, K.; Mishima, N.; Kominami, M.; Yamane, T.

1995-01-01

Output signals of a commercially available on-line laser turbidimeter exhibit fluctuations due to air and/or CO 2 bubbles. A simple data processing algorithm and a personal computer software have been developed to smooth the noisy turbidity data acquired, and to utilize them for the on-line calculations of some kinetic variables involved in batch and fed-batch cultures of uniformly dispersed microorganisms. With this software, about 10 3 instantaneous turbidity data acquired over 55 s are averaged and convert it to dry cell concentration, X, every minute. Also, volume of the culture broth, V, is estimated from the averaged output data of weight loss of feed solution reservoir, W, using an electronic balance on which the reservoir is placed. Then, the computer software is used to perform linear regression analyses over the past 30 min of the total biomass, VX, the natural logarithm of the total biomass, ln(VX), and the weight loss, W, in order to calculate volumetric growth rate, d(VX)/dt, specific growth rate, μ [ = dln(VX)/dt] and the rate of W, dW/dt, every minute in a fed-batch culture. The software used to perform the first-order regression analyses of VX, ln(VX) and W was applied to batch or fed-batch cultures of Escherichia coli on minimum synthetic or natural complex media. Sample determination coefficients of the three different variables (VX, ln(VX) and W) were close to unity, indicating that the calculations are accurate. Furthermore, growth yield, Y x/s , and specific substrate consumption rate, q sc , were approximately estimated from the data, dW/dt and in a ‘balanced’ fed-batch culture of E. coli on the minimum synthetic medium where the computer-aided substrate-feeding system automatically matches well with the cell growth. (author)

Two computational approaches for Monte Carlo based shutdown dose rate calculation with applications to the JET fusion machine

Energy Technology Data Exchange (ETDEWEB)

Petrizzi, L.; Batistoni, P.; Migliori, S. [Associazione EURATOM ENEA sulla Fusione, Frascati (Roma) (Italy); Chen, Y.; Fischer, U.; Pereslavtsev, P. [Association FZK-EURATOM Forschungszentrum Karlsruhe (Germany); Loughlin, M. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, OX (United Kingdom); Secco, A. [Nice Srl Via Serra 33 Camerano Casasco AT (Italy)

2003-07-01

shortly after the deuterium-tritium experiment (DTE1) in 1997. Large computing power, both in terms of amount of data handling and storage and the CPU computing time is needed by the two methods, partly due to the complexity of the problem. With parallel versions of the MCNP code, running on two different platforms, a satisfying accuracy of the calculation has been reached in reasonable times. (authors)

Two computational approaches for Monte Carlo based shutdown dose rate calculation with applications to the JET fusion machine

International Nuclear Information System (INIS)

Petrizzi, L.; Batistoni, P.; Migliori, S.; Chen, Y.; Fischer, U.; Pereslavtsev, P.; Loughlin, M.; Secco, A.

2003-01-01

shortly after the deuterium-tritium experiment (DTE1) in 1997. Large computing power, both in terms of amount of data handling and storage and the CPU computing time is needed by the two methods, partly due to the complexity of the problem. With parallel versions of the MCNP code, running on two different platforms, a satisfying accuracy of the calculation has been reached in reasonable times. (authors)

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.

Radiation damage calculations for compound materials

International Nuclear Information System (INIS)

Greenwood, L.R.

1989-01-01

Displacement damage calculations can be performed for 40 elements in the energy range up to 20 MeV with the SPECTER computer code. A recent addition to the code, called SPECOMP, can intermix atomic recoil energy distributions for any four elements to calculate the proper displacement damage for compound materials. The calculations take advantage of the atomic recoil data in the SPECTER libraries, which were determined by the DISCS computer code, using evaluated neutron cross section and angular distribution data in ENDF/B-V. Resultant damage cross sections for any compound can be added to the SPECTER libraries for the routine calculation of displacements in any given neutron field. Users do not require access to neutron cross section files. Results are presented for a variety of fusion materials and a new ceramic superconductor material. Future plans and nuclear data needs are discussed. 11 refs., 6 figs., 1 tab

Systems for neutronic, thermohydraulic and shielding calculation in personal computers

International Nuclear Information System (INIS)

Villarino, E.A.; Abbate, P.; Lovotti, O.; Santini, M.

1990-01-01

The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author) [es

Continuum variational and diffusion quantum Monte Carlo calculations

International Nuclear Information System (INIS)

Needs, R J; Towler, M D; Drummond, N D; Lopez RIos, P

2010-01-01

This topical review describes the methodology of continuum variational and diffusion quantum Monte Carlo calculations. These stochastic methods are based on many-body wavefunctions and are capable of achieving very high accuracy. The algorithms are intrinsically parallel and well suited to implementation on petascale computers, and the computational cost scales as a polynomial in the number of particles. A guide to the systems and topics which have been investigated using these methods is given. The bulk of the article is devoted to an overview of the basic quantum Monte Carlo methods, the forms and optimization of wavefunctions, performing calculations under periodic boundary conditions, using pseudopotentials, excited-state calculations, sources of calculational inaccuracy, and calculating energy differences and forces. (topical review)

IRIS core criticality calculations

International Nuclear Information System (INIS)

Jecmenica, R.; Trontl, K.; Pevec, D.; Grgic, D.

2003-01-01

Three-dimensional Monte Carlo computer code KENO-VI of CSAS26 sequence of SCALE-4.4 code system was applied for pin-by-pin calculations of the effective multiplication factor for the first cycle IRIS reactor core. The effective multiplication factors obtained by the above mentioned Monte Carlo calculations using 27-group ENDF/B-IV library and 238-group ENDF/B-V library have been compared with the effective multiplication factors achieved by HELIOS/NESTLE, CASMO/SIMULATE, and modified CORD-2 nodal calculations. The results of Monte Carlo calculations are found to be in good agreement with the results obtained by the nodal codes. The discrepancies in effective multiplication factor are typically within 1%. (author)

Parallel processing of dose calculation for external photon beam therapy

International Nuclear Information System (INIS)

Kunieda, Etsuo; Ando, Yutaka; Tsukamoto, Nobuhiro; Ito, Hisao; Kubo, Atsushi

1994-01-01

We implemented external photon beam dose calculation programs into a parallel processor system consisting of Transputers, 32-bit processors especially suitable for multi-processor configuration. Two network conformations, binary-tree and pipeline, were evaluated for rectangular and irregular field dose calculation algorithms. Although computation speed increased in proportion to the number of CPU, substantial overhead caused by inter-processor communication occurred when a smaller computation load was delivered to each processor. On the other hand, for irregular field calculation, which requires more computation capability for each calculation point, the communication overhead was still less even when more than 50 processors were involved. Real-time responses could be expected for more complex algorithms by increasing the number of processors. (author)

Dose calculation system for remotely supporting radiotherapy

International Nuclear Information System (INIS)

Saito, K.; Kunieda, E.; Narita, Y.; Kimura, H.; Hirai, M.; Deloar, H. M.; Kaneko, K.; Ozaki, M.; Fujisaki, T.; Myojoyama, A.; Saitoh, H.

2005-01-01

The dose calculation system IMAGINE is being developed keeping in mind remotely supporting external radiation therapy using photon beams. The system is expected to provide an accurate picture of the dose distribution in a patient body, using a Monte Carlo calculation that employs precise models of the patient body and irradiation head. The dose calculation will be performed utilising super-parallel computing at the dose calculation centre, which is equipped with the ITBL computer, and the calculated results will be transferred through a network. The system is intended to support the quality assurance of current, widely carried out radiotherapy and, further, to promote the prevalence of advanced radiotherapy. Prototypes of the modules constituting the system have already been constructed and used to obtain basic data that are necessary in order to decide on the concrete design of the system. The final system will be completed in 2007. (authors)

Laser Beam and Resonator Calculations on Desktop Computers.

Science.gov (United States)

Doumont, Jean-Luc

There is a continuing interest in the design and calculation of laser resonators and optical beam propagation. In particular, recently, interest has increased in developing concepts such as one-sided unstable resonators, supergaussian reflectivity profiles, diode laser modes, beam quality concepts, mode competition, excess noise factors, and nonlinear Kerr lenses. To meet these calculation needs, I developed a general-purpose software package named PARAXIA ^{rm TM}, aimed at providing optical scientists and engineers with a set of powerful design and analysis tools that provide rapid and accurate results and are extremely easy to use. PARAXIA can handle separable paraxial optical systems in cartesian or cylindrical coordinates, including complex-valued and misaligned ray matrices, with full diffraction effects between apertures. It includes the following programs:. ABCD provides complex-valued ray-matrix and gaussian -mode analyses for arbitrary paraxial resonators and optical systems, including astigmatism and misalignment in each element. This program required that I generalize the theory of gaussian beam propagation to the case of an off-axis gaussian beam propagating through a misaligned, complex -valued ray matrix. FRESNEL uses FFT and FHT methods to propagate an arbitrary wavefront through an arbitrary paraxial optical system using Huygens' integral in rectangular or radial coordinates. The wavefront can be multiplied by an arbitrary mirror profile and/or saturable gain sheet on each successive propagation through the system. I used FRESNEL to design a one-sided negative-branch unstable resonator for a free -electron laser, and to show how a variable internal aperture influences the mode competition and beam quality in a stable cavity. VSOURCE implements the virtual source analysis to calculate eigenvalues and eigenmodes for unstable resonators with both circular and rectangular hard-edged mirrors (including misaligned rectangular systems). I used VSOURCE to

Effective action calculation in lattice QCD

International Nuclear Information System (INIS)

Hoek, J.

1983-01-01

A method (called the effective action method) devised to make analytic calculations in Quantum Chromodynamics in the region of strong coupling is presented. First, the author deals with developing the calculation of a strong coupling expansion of the generating functional for gauge systems on a lattice with arbitrary sources. An accompanying manual describes the implementation of this calculation on a computer. The next step consists of substituting the expressions for the one-link free energies for a specific gauge group in the result of the previous calculation. This process of substitution, together with the replacement of the sources by a bilinear combination of fermion fields, is described for the group SU(3). More details on the implementation of the substitution scheme on a computer can be found in the accompanying manual. From the effective action thus obtained in terms of meson fields and baryon fields the Green functions of the theory can be derived. As an illustrative application the effective potential determining the vacuum expectation value of the meson field is calculated. (Auth.)

A users manual for a computer program which calculates time optical geocentric transfers using solar or nuclear electric and high thrust propulsion

Science.gov (United States)

Sackett, L. L.; Edelbaum, T. N.; Malchow, H. L.

1974-01-01

This manual is a guide for using a computer program which calculates time optimal trajectories for high-and low-thrust geocentric transfers. Either SEP or NEP may be assumed and a one or two impulse, fixed total delta V, initial high thrust phase may be included. Also a single impulse of specified delta V may be included after the low thrust state. The low thrust phase utilizes equinoctial orbital elements to avoid the classical singularities and Kryloff-Boguliuboff averaging to help insure more rapid computation time. The program is written in FORTRAN 4 in double precision for use on an IBM 360 computer. The manual includes a description of the problem treated, input/output information, examples of runs, and source code listings.

Quantum computing applied to calculations of molecular energies

Czech Academy of Sciences Publication Activity Database

Pittner, Jiří; Veis, L.

2011-01-01

Roč. 241, - (2011), 151-phys ISSN 0065-7727. [National Meeting and Exposition of the American-Chemical-Society (ACS) /241./. 27.03.2011-31.03.2011, Anaheim] Institutional research plan: CEZ:AV0Z40400503 Keywords : molecular energie * quantum computers Subject RIV: CF - Physical ; Theoretical Chemistry

First-principles calculations of mobility

Science.gov (United States)

Krishnaswamy, Karthik

First-principles calculations can be a powerful predictive tool for studying, modeling and understanding the fundamental scattering mechanisms impacting carrier transport in materials. In the past, calculations have provided important qualitative insights, but numerical accuracy has been limited due to computational challenges. In this talk, we will discuss some of the challenges involved in calculating electron-phonon scattering and carrier mobility, and outline approaches to overcome them. Topics will include the limitations of models for electron-phonon interaction, the importance of grid sampling, and the use of Gaussian smearing to replace energy-conserving delta functions. Using prototypical examples of oxides that are of technological importance-SrTiO3, BaSnO3, Ga2O3, and WO3-we will demonstrate computational approaches to overcome these challenges and improve the accuracy. One approach that leads to a distinct improvement in the accuracy is the use of analytic functions for the band dispersion, which allows for an exact solution of the energy-conserving delta function. For select cases, we also discuss direct quantitative comparisons with experimental results. The computational approaches and methodologies discussed in the talk are general and applicable to other materials, and greatly improve the numerical accuracy of the calculated transport properties, such as carrier mobility, conductivity and Seebeck coefficient. This work was performed in collaboration with B. Himmetoglu, Y. Kang, W. Wang, A. Janotti and C. G. Van de Walle, and supported by the LEAST Center, the ONR EXEDE MURI, and NSF.

FCJ-135 Feral Computing: From Ubiquitous Calculation to Wild Interactions

Directory of Open Access Journals (Sweden)

Matthew Fuller

2011-12-01

Full Text Available In ‘The Coming Age of Calm Technology’, Mark Weiser and John Seely Brown are clear in their assertions, what really ‘matters’ about technology is not technology in itself, rather, its capacity to continuously recreate our relationship with the world at large (Brown and Weiser 1996. Even though they promote such an idea under the banner of ‘calm technology’, what is central to their thesis is the mutational capacities brought into the world by the spillage of computation out from its customary boxes. What their work tends to occlude is that in setting the sinking of technology almost imperceptibly, but deeply into the ‘everyday’ as a target for ubiquitous computing, other possibilities are masked, for instance, those of greater hackability or interrogability of such technologies. Our contention is that making ubicomp seamless (MacColl et al, 2002 tends to obfuscate the potential of computation in reworking computational subjects, including societies, modes of life, and inter-relations with the dynamics of thought and the composition of experience and understanding.

Study of Variation in Dose Calculation Accuracy Between kV Cone-Beam Computed Tomography and kV fan-Beam Computed Tomography

Science.gov (United States)

Kaliyaperumal, Venkatesan; Raphael, C. Jomon; Varghese, K. Mathew; Gopu, Paul; Sivakumar, S.; Boban, Minu; Raj, N. Arunai Nambi; Senthilnathan, K.; Babu, P. Ramesh

2017-01-01

Cone-beam computed tomography (CBCT) images are presently used for geometric verification for daily patient positioning. In this work, we have compared the images of CBCT with the images of conventional fan beam CT (FBCT) in terms of image quality and Hounsfield units (HUs). We also compared the dose calculated using CBCT with that of FBCT. Homogenous RW3 plates and Catphan phantom were scanned by FBCT and CBCT. In RW3 and Catphan phantom, percentage depth dose (PDD), profiles, isodose distributions (for intensity modulated radiotherapy plans), and calculated dose volume histograms were compared. The HU difference was within ± 20 HU (central region) and ± 30 HU (peripheral region) for homogeneous RW3 plates. In the Catphan phantom, the difference in HU was ± 20 HU in the central area and peripheral areas. The HU differences were within ± 30 HU for all HU ranges starting from −1000 to 990 in phantom and patient images. In treatment plans done with simple symmetric and asymmetric fields, dose difference (DD) between CBCT plan and FBCT plan was within 1.2% for both phantoms. In intensity modulated radiotherapy (IMRT) treatment plans, for different target volumes, the difference was <2%. This feasibility study investigated HU variation and dose calculation accuracy between FBCT and CBCT based planning and has validated inverse planning algorithms with CBCT. In our study, we observed a larger deviation of HU values in the peripheral region compared to the central region. This is due to the ring artifact and scatter contribution which may prevent the use of CBCT as the primary imaging modality for radiotherapy treatment planning. The reconstruction algorithm needs to be modified further for improving the image quality and accuracy in HU values. However, our study with TG-119 and intensity modulated radiotherapy test targets shows that CBCT can be used for adaptive replanning as the recalculation of dose with the anisotropic analytical algorithm is in full accord

Calculation of molecular free energies in classical potentials

International Nuclear Information System (INIS)

Farhi, Asaf; Singh, Bipin

2016-01-01

Free energies of molecules can be calculated by quantum chemistry computations or by normal mode classical calculations. However, the first can be computationally impractical for large molecules and the second is based on the assumption of harmonic dynamics. We present a novel, accurate and complete calculation of molecular free energies in standard classical potentials. In this method we transform the molecule by relaxing potential terms which depend on the coordinates of a group of atoms in that molecule and calculate the free energy difference associated with the transformation. Then, since the transformed molecule can be treated as non-interacting systems, the free energy associated with these atoms is analytically or numerically calculated. This two-step calculation can be applied to calculate free energies of molecules or free energy difference between (possibly large) molecules in a general environment. We demonstrate the method in free energy calculations for methanethiol and butane molecules in vacuum and solvent. We suggest the potential application of free energy calculation of chemical reactions in classical molecular simulations. (paper)

Parallel scalability of Hartree-Fock calculations

Science.gov (United States)

Chow, Edmond; Liu, Xing; Smelyanskiy, Mikhail; Hammond, Jeff R.

2015-03-01

Quantum chemistry is increasingly performed using large cluster computers consisting of multiple interconnected nodes. For a fixed molecular problem, the efficiency of a calculation usually decreases as more nodes are used, due to the cost of communication between the nodes. This paper empirically investigates the parallel scalability of Hartree-Fock calculations. The construction of the Fock matrix and the density matrix calculation are analyzed separately. For the former, we use a parallelization of Fock matrix construction based on a static partitioning of work followed by a work stealing phase. For the latter, we use density matrix purification from the linear scaling methods literature, but without using sparsity. When using large numbers of nodes for moderately sized problems, density matrix computations are network-bandwidth bound, making purification methods potentially faster than eigendecomposition methods.

Impact of neutron resonance treatments on reactor calculation

International Nuclear Information System (INIS)

Leszczynski, F.

1988-01-01

The neutron resonance treatment on reactor calculation is one of the not completely resolved problems of reactor theory. The calculation required on design, fuel management and accident analysis of nuclear reactors contains adjust coefficients and semi-empirical values introduced on the computer codes; these values are obtained comparing calculation results with experimental values and more exact calculation results. This is made when the characteristics of the analyzed system are such that this type of comparisons are possible. The impact that one fixed resonance treatment method have on the final evaluation of physics reactor parameters, reactivity, power distribution, etc., is useful to know. In this work, the differences between calculated parameters with two different methods of resonance treatment in cell calculations are shown. It is concluded that improvements on resonance treatment are necessary for growing the reliability on core calculations results. Finally, possible improvements, easy to implement in current computer codes, are presented. (Author) [es

Computer Program for Calculation of Complex Chemical Equilibrium Compositions, Rocket Performance, Incident and Reflected Shocks, and Chapman-Jouguet Detonations. Interim Revision, March 1976

Science.gov (United States)

Gordon, S.; Mcbride, B. J.

1976-01-01

A detailed description of the equations and computer program for computations involving chemical equilibria in complex systems is given. A free-energy minimization technique is used. The program permits calculations such as (1) chemical equilibrium for assigned thermodynamic states (T,P), (H,P), (S,P), (T,V), (U,V), or (S,V), (2) theoretical rocket performance for both equilibrium and frozen compositions during expansion, (3) incident and reflected shock properties, and (4) Chapman-Jouguet detonation properties. The program considers condensed species as well as gaseous species.

Prospects in deterministic three dimensional whole-core transport calculations

International Nuclear Information System (INIS)

Sanchez, Richard

2012-01-01

The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

Calculation and construction of electron-diffraction photographs using computer

International Nuclear Information System (INIS)

Khayurov, S.S.; Notkin, A.B.

1981-01-01

A method of computer construction and indexing of theoretical electronograms for monophase structures with arbitrary type of crystal lattice and for polyphase ones with known orientational coorrelations between phases is presented. Electron-diffraction photograph is presented, obtained from the foil area of two-phase VT22 alloy at β phase orientation in comparison with theoretical electron-diffraction photographs, built ap by computer, with the [100] β phase zone axis and with three variants of α phase orientation relatively to β phase. It is shown that on the experimental electron-diffraction photograph simultaneously presents α-phase three orientations, which reflexes can be indexing correctly [ru

Hybrid reduced order modeling for assembly calculations

Energy Technology Data Exchange (ETDEWEB)

Bang, Y.; Abdel-Khalik, H. S. [North Carolina State University, Raleigh, NC (United States); Jessee, M. A.; Mertyurek, U. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

2013-07-01

While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system. (authors)

FISPIN - a computer code for nuclide inventory calculations

International Nuclear Information System (INIS)

Burstall, R.F.

1979-10-01

The code is used for assessment of three groups of nuclides, the actinides, the fission products, and structural materials. The methods of calculation are described, together with the input and output of the code and examples of both. Recommendations are given for the best use of the code. (author)

Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

International Nuclear Information System (INIS)

Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae

2016-01-01

The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air

Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

Energy Technology Data Exchange (ETDEWEB)

Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air.

Bending Moment Calculations for Piles Based on the Finite Element Method

Directory of Open Access Journals (Sweden)

Yu-xin Jie

2013-01-01

Full Text Available Using the finite element analysis program ABAQUS, a series of calculations on a cantilever beam, pile, and sheet pile wall were made to investigate the bending moment computational methods. The analyses demonstrated that the shear locking is not significant for the passive pile embedded in soil. Therefore, higher-order elements are not always necessary in the computation. The number of grids across the pile section is important for bending moment calculated with stress and less significant for that calculated with displacement. Although computing bending moment with displacement requires fewer grid numbers across the pile section, it sometimes results in variation of the results. For displacement calculation, a pile row can be suitably represented by an equivalent sheet pile wall, whereas the resulting bending moments may be different. Calculated results of bending moment may differ greatly with different grid partitions and computational methods. Therefore, a comparison of results is necessary when performing the analysis.

Structural system reliability calculation using a probabilistic fault tree analysis method

Science.gov (United States)

Torng, T. Y.; Wu, Y.-T.; Millwater, H. R.

1992-01-01

The development of a new probabilistic fault tree analysis (PFTA) method for calculating structural system reliability is summarized. The proposed PFTA procedure includes: developing a fault tree to represent the complex structural system, constructing an approximation function for each bottom event, determining a dominant sampling sequence for all bottom events, and calculating the system reliability using an adaptive importance sampling method. PFTA is suitable for complicated structural problems that require computer-intensive computer calculations. A computer program has been developed to implement the PFTA.

Fast electrostatic force calculation on parallel computer clusters

International Nuclear Information System (INIS)

Kia, Amirali; Kim, Daejoong; Darve, Eric

2008-01-01

The fast multipole method (FMM) and smooth particle mesh Ewald (SPME) are well known fast algorithms to evaluate long range electrostatic interactions in molecular dynamics and other fields. FMM is a multi-scale method which reduces the computation cost by approximating the potential due to a group of particles at a large distance using few multipole functions. This algorithm scales like O(N) for N particles. SPME algorithm is an O(NlnN) method which is based on an interpolation of the Fourier space part of the Ewald sum and evaluating the resulting convolutions using fast Fourier transform (FFT). Those algorithms suffer from relatively poor efficiency on large parallel machines especially for mid-size problems around hundreds of thousands of atoms. A variation of the FMM, called PWA, based on plane wave expansions is presented in this paper. A new parallelization strategy for PWA, which takes advantage of the specific form of this expansion, is described. Its parallel efficiency is compared with SPME through detail time measurements on two different computer clusters

Analytical predictions of SGEMP response and comparisons with computer calculations

International Nuclear Information System (INIS)

de Plomb, E.P.

1976-01-01

An analytical formulation for the prediction of SGEMP surface current response is presented. Only two independent dimensionless parameters are required to predict the peak magnitude and rise time of SGEMP induced surface currents. The analysis applies to limited (high fluence) emission as well as unlimited (low fluence) emission. Cause-effect relationships for SGEMP response are treated quantitatively, and yield simple power law dependencies between several physical variables. Analytical predictions for a large matrix of SGEMP cases are compared with an array of about thirty-five computer solutions of similar SGEMP problems, which were collected from three independent research groups. The theoretical solutions generally agree with the computer solutions as well as the computer solutions agree with one another. Such comparisons typically show variations less than a ''factor of two.''

Investigation of an inventory calculation model for a solvent extraction system and the development of its computer programme - SEPHIS-J

International Nuclear Information System (INIS)

Ihara, Hitoshi; Nishimura, Hideo; Ikawa, Koji; Ido, Masaru.

1986-11-01

In order to improve the applicability of near-real-time materials accountancy (N.R.T.MA) to a reprocessing plant, it is necessary to develop an estimation method for the nuclear material inventory at a solvent extraction system under operation. For designing the solvent extraction system, such computer codes as SEPHIS, SOLVEX and TRANSIENTS had been used. Accuracy of these codes in tracing operations and predicting inventories in the extraction system had been discussed. Then, much better codes, e.g., SEPHIS Mod4 and PUBG, were developed. Unfortunately, SEPHIS Mod4 was not available in countries other than the USA and PUBG was not suitable for use with a mini-computer which would be practical as a field computer because of quite a lot of computing time needed. The authors investigated an inventory estimation model compatible with PUBG in functions and developed the corresponding computer programme, SEPHIS-J, based on the SEPHIS Mod3 code, resulting in a third of computing time compared with PUBG. They also validated the programme by calculating a static state as well as a dynamic one of the solvent extraction process and by comparing them among the programme, SEPHIS Mod3 and PUBG. Using the programme, it was shown that the inventory changes due to changes of feed flow and concentration were not so small that they might be neglected although the changes of feed flow and concentration were within measurement errors. (author)

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Yuan, Y.C. [Square Y Consultants, Orchard Park, NY (US); Chen, S.Y.; Biwer, B.M.; LePoire, D.J. [Argonne National Lab., IL (US)

1995-11-01

This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows{trademark} environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident.

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

International Nuclear Information System (INIS)

Yuan, Y.C.; Chen, S.Y.; Biwer, B.M.; LePoire, D.J.

1995-11-01

This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows trademark environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident

Preliminary isodose calculation for gynecological curietherapy

International Nuclear Information System (INIS)

Bridier, A.; Dutreix, A.; Gerbaulet, A.; Chassagne, D.

1981-01-01

We present a preliminary method of calculating the dimensions of the reference isodose, based upon the geometrical distribution and length of the sources used, their linear activity and the length of treatment, that does not require use of a computer. Inversely, this method can be used to determine the factors necessary to produce a given shape of isodose, and also to predict the change in shape of the isodose that will be produced by altering the various factors. This method was derived from a systematic computer study of dose distribution in which each factor was varied independently of all others. The dimensions of the isodoses, calculated by this method, were found to be in agreement with those derived from computer calculation to within an error of about 2 mm. The method is only applicable for a limited range of positions of the vaginal sources. The influence of the positions of these sources along the line of the axis of uterine catheter and of their inclination to this line, are currently being studied. The results are presented as mathematical formulae relating each dimension of the isodose curves to the features of the application, but could equally well be expressed in tabular form that would be more convenient for everyday use. An example of the calculation used is given to facilitate understanding of the method [fr

A computer code for the prediction of mill gases and hot air distribution between burners sections as input parameters for 3D CFD furnace calculation

International Nuclear Information System (INIS)

Tucakovic, Dragan; Zivanovic, Titoslav; Beloshevic, Srdjan

2006-01-01

Current computer technology development enables application of powerful software packages that can provide a reliable insight into real operating conditions of a steam boiler in the Thermal Power Plant. Namely, an application of CFD code to the 3D analysis of combustion and heat transfer in a furnace provides temperature, velocity and concentration fields in both cross sectional and longitudinal planes of the observed furnace. In order to obtain reliable analytical results, which corresponds to real furnace conditions, it is necessary to accurately predict a distribution of mill gases and hot air between burners' sections, because these parameters are input values for the furnace 3D calculation. Regarding these tasks, the computer code for the prediction of mill gases and hot air distribution has been developed at the Department for steam boilers of the Faculty of Mechanical Engineering in Belgrade. The code is based on simultaneous calculations of material and heat balances for fan mill and air tracts. The aim of this paper is to present a methodology of performed calculations and results obtained for the steam boiler furnace of 350 MWe Thermal Power Plant equipped with eight fan mills. Key words: mill gases, hot air, aerodynamic calculation, air tract, mill tract.

Biases in Monte Carlo eigenvalue calculations

Energy Technology Data Exchange (ETDEWEB)

Gelbard, E.M.

1992-12-01

The Monte Carlo method has been used for many years to analyze the neutronics of nuclear reactors. In fact, as the power of computers has increased the importance of Monte Carlo in neutronics has also increased, until today this method plays a central role in reactor analysis and design. Monte Carlo is used in neutronics for two somewhat different purposes, i.e., (a) to compute the distribution of neutrons in a given medium when the neutron source-density is specified, and (b) to compute the neutron distribution in a self-sustaining chain reaction, in which case the source is determined as the eigenvector of a certain linear operator. In (b), then, the source is not given, but must be computed. In the first case (the ``fixed-source`` case) the Monte Carlo calculation is unbiased. That is to say that, if the calculation is repeated (``replicated``) over and over, with independent random number sequences for each replica, then averages over all replicas will approach the correct neutron distribution as the number of replicas goes to infinity. Unfortunately, the computation is not unbiased in the second case, which we discuss here.

Biases in Monte Carlo eigenvalue calculations

Energy Technology Data Exchange (ETDEWEB)

Gelbard, E.M.

1992-01-01

The Monte Carlo method has been used for many years to analyze the neutronics of nuclear reactors. In fact, as the power of computers has increased the importance of Monte Carlo in neutronics has also increased, until today this method plays a central role in reactor analysis and design. Monte Carlo is used in neutronics for two somewhat different purposes, i.e., (a) to compute the distribution of neutrons in a given medium when the neutron source-density is specified, and (b) to compute the neutron distribution in a self-sustaining chain reaction, in which case the source is determined as the eigenvector of a certain linear operator. In (b), then, the source is not given, but must be computed. In the first case (the fixed-source'' case) the Monte Carlo calculation is unbiased. That is to say that, if the calculation is repeated ( replicated'') over and over, with independent random number sequences for each replica, then averages over all replicas will approach the correct neutron distribution as the number of replicas goes to infinity. Unfortunately, the computation is not unbiased in the second case, which we discuss here.

Biases in Monte Carlo eigenvalue calculations

International Nuclear Information System (INIS)

Gelbard, E.M.

1992-01-01

The Monte Carlo method has been used for many years to analyze the neutronics of nuclear reactors. In fact, as the power of computers has increased the importance of Monte Carlo in neutronics has also increased, until today this method plays a central role in reactor analysis and design. Monte Carlo is used in neutronics for two somewhat different purposes, i.e., (a) to compute the distribution of neutrons in a given medium when the neutron source-density is specified, and (b) to compute the neutron distribution in a self-sustaining chain reaction, in which case the source is determined as the eigenvector of a certain linear operator. In (b), then, the source is not given, but must be computed. In the first case (the ''fixed-source'' case) the Monte Carlo calculation is unbiased. That is to say that, if the calculation is repeated (''replicated'') over and over, with independent random number sequences for each replica, then averages over all replicas will approach the correct neutron distribution as the number of replicas goes to infinity. Unfortunately, the computation is not unbiased in the second case, which we discuss here

Dual-energy imaging method to improve the image quality and the accuracy of dose calculation for cone-beam computed tomography.

Science.gov (United States)

Men, Kuo; Dai, Jianrong; Chen, Xinyuan; Li, Minghui; Zhang, Ke; Huang, Peng

2017-04-01

To improve the image quality and accuracy of dose calculation for cone-beam computed tomography (CT) images through implementation of a dual-energy cone-beam computed tomography method (DE-CBCT), and evaluate the improvement quantitatively. Two sets of CBCT projections were acquired using the X-ray volumetric imaging (XVI) system on a Synergy (Elekta, Stockholm, Sweden) system with 120kV (high) and 70kV (low) X-rays, respectively. Then, the electron density relative to water (relative electron density (RED)) of each voxel was calculated using a projection-based dual-energy decomposition method. As a comparison, single-energy cone-beam computed tomography (SE-CBCT) was used to calculate RED with the Hounsfield unit-RED calibration curve generated by a CIRS phantom scan with identical imaging parameters. The imaging dose was measured with a dosimetry phantom. The image quality was evaluated quantitatively using a Catphan 503 phantom with the evaluation indices of the reproducibility of the RED values, high-contrast resolution (MTF 50% ), uniformity, and signal-to-noise ratio (SNR). Dose calculation of two simulated volumetric-modulated arc therapy plans using an Eclipse treatment-planning system (Varian Medical Systems, Palo Alto, CA, USA) was performed on an Alderson Rando Head and Neck (H&N) phantom and a Pelvis phantom. Fan-beam planning CT images for the H&N and Pelvis phantom were set as the reference. A global three-dimensional gamma analysis was used to compare dose distributions with the reference. The average gamma values for targets and OAR were analyzed with paired t-tests between DE-CBCT and SE-CBCT. In two scans (H&N scan and body scan), the imaging dose of DE-CBCT increased by 1.0% and decreased by 1.3%. It had a better reproducibility of the RED values (mean bias: 0.03 and 0.07) compared with SE-CBCT (mean bias: 0.13 and 0.16). It also improved the image uniformity (57.5% and 30.1%) and SNR (9.7% and 2.3%), but did not affect the MTF 50% . Gamma

Calculating Cumulative Binomial-Distribution Probabilities

Science.gov (United States)

Scheuer, Ernest M.; Bowerman, Paul N.

1989-01-01

Cumulative-binomial computer program, CUMBIN, one of set of three programs, calculates cumulative binomial probability distributions for arbitrary inputs. CUMBIN, NEWTONP (NPO-17556), and CROSSER (NPO-17557), used independently of one another. Reliabilities and availabilities of k-out-of-n systems analyzed. Used by statisticians and users of statistical procedures, test planners, designers, and numerical analysts. Used for calculations of reliability and availability. Program written in C.

Online plasma calculator

Science.gov (United States)

Wisniewski, H.; Gourdain, P.-A.

2017-10-01

APOLLO is an online, Linux based plasma calculator. Users can input variables that correspond to their specific plasma, such as ion and electron densities, temperatures, and external magnetic fields. The system is based on a webserver where a FastCGI protocol computes key plasma parameters including frequencies, lengths, velocities, and dimensionless numbers. FastCGI was chosen to overcome security problems caused by JAVA-based plugins. The FastCGI also speeds up calculations over PHP based systems. APOLLO is built upon the WT library, which turns any web browser into a versatile, fast graphic user interface. All values with units are expressed in SI units except temperature, which is in electron-volts. SI units were chosen over cgs units because of the gradual shift to using SI units within the plasma community. APOLLO is intended to be a fast calculator that also provides the user with the proper equations used to calculate the plasma parameters. This system is intended to be used by undergraduates taking plasma courses as well as graduate students and researchers who need a quick reference calculation.

Improved scFv Anti-HIV-1 p17 Binding Affinity Guided from the Theoretical Calculation of Pairwise Decomposition Energies and Computational Alanine Scanning

Directory of Open Access Journals (Sweden)

Panthip Tue-ngeun

2013-01-01

Full Text Available Computational approaches have been used to evaluate and define important residues for protein-protein interactions, especially antigen-antibody complexes. In our previous study, pairwise decomposition of residue interaction energies of single chain Fv with HIV-1 p17 epitope variants has indicated the key specific residues in the complementary determining regions (CDRs of scFv anti-p17. In this present investigation in order to determine whether a specific side chain group of residue in CDRs plays an important role in bioactivity, computational alanine scanning has been applied. Molecular dynamics simulations were done with several complexes of original scFv anti-p17 and scFv anti-p17mutants with HIV-1 p17 epitope variants with a production run up to 10 ns. With the combination of pairwise decomposition residue interaction and alanine scanning calculations, the point mutation has been initially selected at the position MET100 to improve the residue binding affinity. The calculated docking interaction energy between a single mutation from methionine to either arginine or glycine has shown the improved binding affinity, contributed from the electrostatic interaction with the negative favorably interaction energy, compared to the wild type. Theoretical calculations agreed well with the results from the peptide ELISA results.

Computer Programs for Uncertainty Analysis of Solubility Calculations: Windows Version and Other Updates of the SENVAR and UNCCON. Program Description and Handling Instructions

International Nuclear Information System (INIS)

Ekberg, Christian; Oedegaard Jensen, Arvid

2004-04-01

Uncertainty and sensitivity analysis is becoming more and more important for testing the reliability of computer predictions. Solubility estimations play important roles for, e.g. underground repositories for nuclear waste, other hazardous materials as well as simple dissolution problems in general or industrial chemistry applications. The calculated solubility of a solid phase is dependent on several input data, e.g. the stability constants for the complexes formed in the solution, the enthalpies of reaction for the formation of these complexes and also the content of other elements in the water used for the dissolution. These input data are determined with more or less accuracy and thus the results of the calculations are uncertain. For the purpose of investigating the effects of these uncertainties several computer programs were developed in the 1990s, e.g. SENVAR, MINVAR and UNCCON. Of these SENVAR and UNCCON now exist as windows programs based on a newer speciation code. In this report we have given an explanation of how the codes work and also given some test cases as handling instructions. The results are naturally similar to the previous ones but the advantages are easier handling and more stable solubility calculations. With these improvements the programs presented here will be more publically accessible

Numericware i: Identical by State Matrix Calculator

Directory of Open Access Journals (Sweden)

Bongsong Kim

2017-02-01

Full Text Available We introduce software, Numericware i, to compute identical by state (IBS matrix based on genotypic data. Calculating an IBS matrix with a large dataset requires large computer memory and takes lengthy processing time. Numericware i addresses these challenges with 2 algorithmic methods: multithreading and forward chopping. The multithreading allows computational routines to concurrently run on multiple central processing unit (CPU processors. The forward chopping addresses memory limitation by dividing a dataset into appropriately sized subsets. Numericware i allows calculation of the IBS matrix for a large genotypic dataset using a laptop or a desktop computer. For comparison with different software, we calculated genetic relationship matrices using Numericware i, SPAGeDi, and TASSEL with the same genotypic dataset. Numericware i calculates IBS coefficients between 0 and 2, whereas SPAGeDi and TASSEL produce different ranges of values including negative values. The Pearson correlation coefficient between the matrices from Numericware i and TASSEL was high at .9972, whereas SPAGeDi showed low correlation with Numericware i (.0505 and TASSEL (.0587. With a high-dimensional dataset of 500 entities by 10 000 000 SNPs, Numericware i spent 382 minutes using 19 CPU threads and 64 GB memory by dividing the dataset into 3 pieces, whereas SPAGeDi and TASSEL failed with the same dataset. Numericware i is freely available for Windows and Linux under CC-BY 4.0 license at https://figshare.com/s/f100f33a8857131eb2db .

Languages for structural calculations

International Nuclear Information System (INIS)

Thomas, J.B.; Chambon, M.R.

1988-01-01

The differences between human and computing languages are recalled. It is argued that they are to some extent structured in antagonistic ways. Languages in structural calculation, in the past, present, and future, are considered. The contribution of artificial intelligence is stressed [fr

Benchmark calculation programme concerning typical LMFBR structures

International Nuclear Information System (INIS)

Donea, J.; Ferrari, G.; Grossetie, J.C.; Terzaghi, A.

1982-01-01

This programme, which is part of a comprehensive activity aimed at resolving difficulties encountered in using design procedures based on ASME Code Case N-47, should allow to get confidence in computer codes which are supposed to provide a realistic prediction of the LMFBR component behaviour. The calculations started on static analysis of typical structures made of non linear materials stressed by cyclic loads. The fluid structure interaction analysis is also being considered. Reasons and details of the different benchmark calculations are described, results obtained are commented and future computational exercise indicated

Temperature Calculations in the Coastal Modeling System

Science.gov (United States)

2017-04-01

ERDC/CHL CHETN-IV-110 April 2017 Approved for public release; distribution is unlimited . Temperature Calculations in the Coastal Modeling...tide) and river discharge at model boundaries, wave radiation stress, and wind forcing over a model computational domain. Physical processes calculated...calculated in the CMS using the following meteorological parameters: solar radiation, cloud cover, air temperature, wind speed, and surface water temperature

Calculation results and experimental testing of doppler feedback coefficients

International Nuclear Information System (INIS)

Yang Shunhai

1989-01-01

The Doppler feedback coefficients are calculated by the interpolation and group collapsing method from multigroup self-shielding factors and infinite dilution cross sections rather than effective resonance integrals by using resonance data base. Since many updated sets of multigroup data are in existence to be selected, the calculation process can be simplified. The heterogeneous effects are taken into account by equivalence relation. The computer code of Doppler feedback coefficients is created on computer CYBER-825 and PDP-11. The results calculated are in good agreement with the experiments

Thermohydraulic calculations of PWR primary circuits

International Nuclear Information System (INIS)

Botelho, D.A.

1984-01-01

Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

Study on high density multi-scale calculation technique

International Nuclear Information System (INIS)

Sekiguchi, S.; Tanaka, Y.; Nakada, H.; Nishikawa, T.; Yamamoto, N.; Yokokawa, M.

2004-01-01

To understand degradation of nuclear materials under irradiation, it is essential to know as much about each phenomenon observed from multi-scale points of view; they are micro-scale in atomic-level, macro-level in structural scale and intermediate level. In this study for application to meso-scale materials (100A ∼ 2μm), computer technology approaching from micro- and macro-scales was developed including modeling and computer application using computational science and technology method. And environmental condition of grid technology for multi-scale calculation was prepared. The software and MD (molecular dynamics) stencil for verifying the multi-scale calculation were improved and their movement was confirmed. (A. Hishinuma)

Molecular dynamics calculation of shear viscosity for molten salt

International Nuclear Information System (INIS)

Okamoto, Yoshihiro; Yokokawa, Mitsuo; Ogawa, Toru

1993-12-01

A computer program of molecular dynamics simulation has been made to calculate shear viscosity of molten salt. Correlation function for an off-diagonal component of stress tensor can be obtained as the results of calculation. Shear viscosity is calculated by integration of the correlation function based on the Kubo-type formula. Shear viscosities for a molten KCl ranging in temperature from 1047K to 1273K were calculated using the program. Calculation of 10 5 steps (1 step corresponds to 5 x 10 -15 s) was performed for each temperature in the 216 ions system. The obtained results were in good agreement with the reported experimental values. The program has been vectorized to achieve a faster computation in supercomputer. It makes possible to calculate the viscosity using a large number of statistics amounting to several million MD steps. (author)

Parallel computations

CERN Document Server

1982-01-01

Parallel Computations focuses on parallel computation, with emphasis on algorithms used in a variety of numerical and physical applications and for many different types of parallel computers. Topics covered range from vectorization of fast Fourier transforms (FFTs) and of the incomplete Cholesky conjugate gradient (ICCG) algorithm on the Cray-1 to calculation of table lookups and piecewise functions. Single tridiagonal linear systems and vectorized computation of reactive flow are also discussed.Comprised of 13 chapters, this volume begins by classifying parallel computers and describing techn

Computer applications in thermochemistry

International Nuclear Information System (INIS)

Vana Varamban, S.

1996-01-01

Knowledge of equilibrium is needed under many practical situations. Simple stoichiometric calculations can be performed by the use of hand calculators. Multi-component, multi-phase gas - solid chemical equilibrium calculations are far beyond the conventional devices and methods. Iterative techniques have to be resorted. Such problems are most elegantly handled by the use of modern computers. This report demonstrates the possible use of computers for chemical equilibrium calculations in the field of thermochemistry and chemical metallurgy. Four modules are explained. To fit the experimental C p data and to generate the thermal functions, to perform equilibrium calculations to the defined conditions, to prepare the elaborate input to the equilibrium and to analyse the calculated results graphically. The principles of thermochemical calculations are briefly described. An extensive input guide is given. Several illustrations are included to help the understanding and usage. (author)

Some computer realizations of the REDUCE-3 calculations for exclusive processes

International Nuclear Information System (INIS)

Darbaidze, Ya.Z.; Merebashvili, Z.V.; Rostovtsev, V.A.

1990-01-01

The REDUCE-3 algorithm for the calculation of the squared gauge invariant set of tree diagrams is given in the α 3 order of the perturbation theory. The necessity of using such program packages as factorizator, 'COLOR'-factor and so on is shown. The correctness of calculation for the infrared radiation corrections as compared with manual calculations is discussed. An example of applying the programs is given for the matrix and noncommutative algebras when the well-known supersymmetric commutative relation is proved. (author)

An approximate but efficient method to calculate free energy trends by computer simulation: Application to dihydrofolate reductase-inhibitor complexes

Science.gov (United States)

Gerber, Paul R.; Mark, Alan E.; van Gunsteren, Wilfred F.

1993-06-01

Derivatives of free energy differences have been calculated by molecular dynamics techniques. The systems under study were ternary complexes of Trimethoprim (TMP) with dihydrofolate reductases of E. coli and chicken liver, containing the cofactor NADPH. Derivatives are taken with respect to modification of TMP, with emphasis on altering the 3-, 4- and 5-substituents of the phenyl ring. A linear approximation allows the encompassing of a whole set of modifications in a single simulation, as opposed to a full perturbation calculation, which requires a separate simulation for each modification. In the case considered here, the proposed technique requires a factor of 1000 less computing effort than a full free energy perturbation calculation. For the linear approximation to yield a significant result, one has to find ways of choosing the perturbation evolution, such that the initial trend mirrors the full calculation. The generation of new atoms requires a careful treatment of the singular terms in the non-bonded interaction. The result can be represented by maps of the changed molecule, which indicate whether complex formation is favoured under movement of partial charges and change in atom polarizabilities. Comparison with experimental measurements of inhibition constants reveals fair agreement in the range of values covered. However, detailed comparison fails to show a significant correlation. Possible reasons for the most pronounced deviations are given.

Large scale calculations for hadron spectroscopy

International Nuclear Information System (INIS)

Rebbi, C.

1985-01-01

The talk reviews some recent Monte Carlo calculations for Quantum Chromodynamics, performed on Euclidean lattices of rather large extent. Purpose of the calculations is to provide accurate determinations of quantities, such as interquark potentials or mass eigenvalues, which are relevant for hadronic spectroscopy. Results obtained in quenched QCD on 16 3 x 32 lattices are illustrated, and a discussion of computational resources and techniques required for the calculations is presented. 18 refs.,3 figs., 2 tabs

Examples of 3-D field calculations using GFUN

International Nuclear Information System (INIS)

Lari, R.J.

1981-01-01

Magnets are described that were calculated using GFUN. A four-step procedure is used to calculate magnets using GFUN. First, the interactive TSO system is used to draw the geometry of the magnet on a Tektronix 4012 graphic display unit. When the geometry is correct, it is stored on a disc file that is shared by the batch computers 3033 and 195. A file JCL and data can be created on TSO and submitted to the batch computers to calculate the magnetization of the steel tetrahedron elements. The results of this approx. 1 hour batch job are stored on disc. In the same job, or in a separate one, the fields can be calculated at desired points and stored on a shared disc. The fourth step is to plot these fields interactively in TSO

An ultrafast line-by-line algorithm for calculating spectral transmittance and radiance

International Nuclear Information System (INIS)

Tan, X.

2013-01-01

An ultrafast line-by-line algorithm for calculating spectral transmittance and radiance of gases is presented. The algorithm is based on fast convolution of the Voigt line profile using Fourier transform and a binning technique. The algorithm breaks a radiative transfer calculation into two steps: a one-time pre-computation step in which a set of pressure independent coefficients are computed using the spectral line information; a normal calculation step in which the Fourier transform coefficients of the optical depth are calculated using the line of sight information and the coefficients pre-computed in the first step, the optical depth is then calculated using an inverse Fourier transform and the spectral transmittance and radiance are calculated. The algorithm is significantly faster than line-by-line algorithms that do not employ special speedup techniques by a factor of 10 3 –10 6 . A case study of the 2.7 μm band of H 2 O vapor is presented. -- Highlights: •An ultrafast line-by-line model based on FFT and a binning technique is presented. •Computationally expensive calculations are factored out into a pre-computation step. •It is 10 3 –10 8 times faster than LBL algorithms that do not employ speedup techniques. •Good agreement with experimental data for the 2.7 μm band of H 2 O

Study of the acceleration of nuclide burnup calculation using GPU with CUDA

International Nuclear Information System (INIS)

Okui, S.; Ohoka, Y.; Tatsumi, M.

2009-01-01

The computation costs of neutronics calculation code become higher as physics models and methods are complicated. The degree of them in neutronics calculation tends to be limited due to available computing power. In order to open a door to the new world, use of GPU for general purpose computing, called GPGPU, has been studied [1]. GPU has multi-threads computing mechanism enabled with multi-processors which realize mush higher performance than CPUs. NVIDIA recently released the CUDA language for general purpose computation which is a C-like programming language. It is relatively easy to learn compared to the conventional ones used for GPGPU, such as OpenGL or CG. Therefore application of GPU to the numerical calculation became much easier. In this paper, we tried to accelerate nuclide burnup calculation, which is important to predict nuclides time dependence in the core, using GPU with CUDA. We chose the 4.-order Runge-Kutta method to solve the nuclide burnup equation. The nuclide burnup calculation and the 4.-order Runge-Kutta method were suitable to the first step of introduction CUDA into numerical calculation because these consist of simple operations of matrices and vectors of single precision where actual codes were written in the C++ language. Our experimental results showed that nuclide burnup calculations with GPU have possibility of speedup by factor of 100 compared to that with CPU. (authors)

FPDCYS and FPSPEC: computer programs for calculating fission-product beta and gamma multigroup spectra from ENDF/B-IV data

International Nuclear Information System (INIS)

Stamatelatos, M.G.; England, T.R.

1977-05-01

FPDCYS and FPSPEC are two FORTRAN computer programs used at the Los Alamos Scientific Laboratory (LASL), in conjunction with the CINDER-10 program, for calculating cumulative fission-product beta and/or gamma multigroup spectra in arbitrary energy structures, and for arbitrary neutron irradiation periods and cooling times. FPDCYS processes ENDF/B-IV fission-product decay energy data to generate multigroup beta and gamma spectra from individual ENDF/B-IV fission-product nuclides. FPSPEC further uses these spectra and the corresponding nuclide activities calculated by the CINDER-10 code to produce cumulative beta and gamma spectra in the same energy grids in which FPDCYS generates individual isotope decay spectra. The code system consisting of CINDER-10, FPDCYS, and FPSPEC has been used for comparisons with experimental spectra and continues to be used at LASL for generating spectra in special user-oriented group structures. 3 figures

Microcomputer generated pipe support calculations

International Nuclear Information System (INIS)

Hankinson, R.F.; Czarnowski, P.; Roemer, R.E.

1991-01-01

The cost and complexity of pipe support design has been a continuing challenge to the construction and modification of commercial nuclear facilities. Typically, pipe support design or qualification projects have required large numbers of engineers centrally located with access to mainframe computer facilities. Much engineering time has been spent repetitively performing a sequence of tasks to address complex design criteria and consolidating the results of calculations into documentation packages in accordance with strict quality requirements. The continuing challenges of cost and quality, the need for support engineering services at operating plant sites, and the substantial recent advances in microcomputer systems suggested that a stand-alone microcomputer pipe support calculation generator was feasible and had become a necessity for providing cost-effective and high quality pipe support engineering services to the industry. This paper outlines the preparation for, and the development of, an integrated pipe support design/evaluation software system which maintains all computer programs in the same environment, minimizes manual performance of standard or repetitive tasks, and generates a high quality calculation which is consistent and easily followed

Accelerating three-dimensional FDTD calculations on GPU clusters for electromagnetic field simulation.

Science.gov (United States)

Nagaoka, Tomoaki; Watanabe, Soichi

2012-01-01

Electromagnetic simulation with anatomically realistic computational human model using the finite-difference time domain (FDTD) method has recently been performed in a number of fields in biomedical engineering. To improve the method's calculation speed and realize large-scale computing with the computational human model, we adapt three-dimensional FDTD code to a multi-GPU cluster environment with Compute Unified Device Architecture and Message Passing Interface. Our multi-GPU cluster system consists of three nodes. The seven GPU boards (NVIDIA Tesla C2070) are mounted on each node. We examined the performance of the FDTD calculation on multi-GPU cluster environment. We confirmed that the FDTD calculation on the multi-GPU clusters is faster than that on a multi-GPU (a single workstation), and we also found that the GPU cluster system calculate faster than a vector supercomputer. In addition, our GPU cluster system allowed us to perform the large-scale FDTD calculation because were able to use GPU memory of over 100 GB.

Neutronic calculation of the next fuel elements for the Argonaut reactor

International Nuclear Information System (INIS)

Oliveira, C.R.E.; Brito Aghina, L.O. de

1981-01-01

The best parameters of the next fuel elements of the Argonaut reactor, at IEN (Instituto de Engenharia Nuclear - Brazil), were determined. The next fuel elements will be rods of an uranium mixture (19.98% enriched), graphite and bakelite. The parameters to be determined are: mixture density, percentage of uranium in the mixture, pellet radius, rod material and elements arrangement (step). The calculations routines consisted in the analysis of several steps, using the LEOPARD computer code for cell calculations and RMAT1D for one dimensional spatial calculations (criticality) with four energy groups. Finally a neutronic study of the Argounat reactors present configuration was done, using the HAMMER computer code (cell), the EXTERMINATOR computer code (two-dimensional calculations) and RAMAT1D. (Author) [pt

Applications of computer algebra

CERN Document Server

1985-01-01

Today, certain computer software systems exist which surpass the computational ability of researchers when their mathematical techniques are applied to many areas of science and engineering. These computer systems can perform a large portion of the calculations seen in mathematical analysis. Despite this massive power, thousands of people use these systems as a routine resource for everyday calculations. These software programs are commonly called "Computer Algebra" systems. They have names such as MACSYMA, MAPLE, muMATH, REDUCE and SMP. They are receiving credit as a computational aid with in­ creasing regularity in articles in the scientific and engineering literature. When most people think about computers and scientific research these days, they imagine a machine grinding away, processing numbers arithmetically. It is not generally realized that, for a number of years, computers have been performing non-numeric computations. This means, for example, that one inputs an equa­ tion and obtains a closed for...

Infinite slab-shield dose calculations

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

I calculated neutron and gamma-ray equivalent doses leaking through a variety of infinite (laminate) slab-shields. In the shield computations, I used, as the incident neutron spectrum, the leakage spectrum (<20 MeV) calculated for the LANSCE tungsten production target at 90 degree to the target axis. The shield thickness was fixed at 60 cm. The results of the shield calculations show a minimum in the total leakage equivalent dose if the shield is 40-45 cm of iron followed by 20-15 cm of borated (5% B) polyethylene. High-performance shields can be attained by using multiple laminations. The calculated dose at the shield surface is very dependent on shield material. 4 refs., 4 figs., 1 tab

Computational study of the hydrolysis of carbonyl sulphide: Thermodynamics and kinetic constants estimation using ab initio calculations

International Nuclear Information System (INIS)

Vidal-Vidal, Á.; Pérez-Rodríguez, M.; Piñeiro, M.M.

2017-01-01

Highlights: • OCS hydrolysis equilibrium constants were calculated using QM composite methods. • CBS-QB3 was found to be the most adequate method for OCS thermodynamic calculations. • Calculated hydrolysis yields decrease when temperature increases. • The isotopic effect is less significant than temperature or initial concentration dependences. - Abstract: Carbonyl sulphide is the predominant sulphur compound in the atmosphere, contributing to the formation of aerosol particles affecting global climate. Human activity has significantly increased its total amount since the beginning of the Industrial Revolution due to its presence in petroleum and coal, reason why it is necessary to understand and control its emissions. On the other hand, carbonyl sulphide is an undesired substance for catalysis in important industrial processes. Hydrolysis is the most promising among the different strategies to reduce its presence, giving as products carbon dioxide and hydrogen sulphide. In the present work, the mechanism of reaction of carbonyl sulphide hydrolysis process in gas phase was studied from 400 K to 1500 K, equilibrium constants were obtained and reaction yields were estimated, by means of composite quantum-computational methods. Good agreement with literature experimental results confirms the suitability of the chosen methods, specially CBS-QB3, in supporting the reaction mechanism, giving accurate equilibrium constant values, and obtaining realistic yields. The effect of isotopic substitution in OCS was also studied, from 300 K to 1500 K, being much less significant than temperature dependence.

Calculation of brain atrophy using computed tomography and a new atrophy measurement tool

Science.gov (United States)

Bin Zahid, Abdullah; Mikheev, Artem; Yang, Andrew Il; Samadani, Uzma; Rusinek, Henry

2015-03-01

Purpose: To determine if brain atrophy can be calculated by performing volumetric analysis on conventional computed tomography (CT) scans in spite of relatively low contrast for this modality. Materials & Method: CTs for 73 patients from the local Veteran Affairs database were selected. Exclusion criteria: AD, NPH, tumor, and alcohol abuse. Protocol: conventional clinical acquisition (Toshiba; helical, 120 kVp, X-ray tube current 300mA, slice thickness 3-5mm). Locally developed, automatic algorithm was used to segment intracranial cavity (ICC) using (a) white matter seed (b) constrained growth, limited by inner skull layer and (c) topological connectivity. ICC was further segmented into CSF and brain parenchyma using a threshold of 16 Hu. Results: Age distribution: 25-95yrs; (Mean 67+/-17.5yrs.). Significant correlation was found between age and CSF/ICC(r=0.695, pautomated software and conventional CT. Compared to MRI, CT is more widely available, cheaper, and less affected by head motion due to ~100 times shorter scan time. Work is in progress to improve the precision of the measurements, possibly leading to assessment of longitudinal changes within the patient.

Adaptation of HAMMER computer code to CYBER 170/750 computer

International Nuclear Information System (INIS)

Pinheiro, A.M.B.S.; Nair, R.P.K.

1982-01-01

The adaptation of HAMMER computer code to CYBER 170/750 computer is presented. The HAMMER code calculates cell parameters by multigroup transport theory and reactor parameters by few group diffusion theory. The auxiliary programs, the carried out modifications and the use of HAMMER system adapted to CYBER 170/750 computer are described. (M.C.K.) [pt

Calculating radiation exposure and dose

International Nuclear Information System (INIS)

Hondros, J.

1987-01-01

This paper discusses the methods and procedures used to calculate the radiation exposures and radiation doses to designated employees of the Olympic Dam Project. Each of the three major exposure pathways are examined. These are: gamma irradiation, radon daughter inhalation and radioactive dust inhalation. A further section presents ICRP methodology for combining individual pathway exposures to give a total dose figure. Computer programs used for calculations and data storage are also presented briefly

Monte Carlo calculations of electron transport on microcomputers

International Nuclear Information System (INIS)

Chung, Manho; Jester, W.A.; Levine, S.H.; Foderaro, A.H.

1990-01-01

In the work described in this paper, the Monte Carlo program ZEBRA, developed by Berber and Buxton, was converted to run on the Macintosh computer using Microsoft BASIC to reduce the cost of Monte Carlo calculations using microcomputers. Then the Eltran2 program was transferred to an IBM-compatible computer. Turbo BASIC and Microsoft Quick BASIC have been used on the IBM-compatible Tandy 4000SX computer. The paper shows the running speed of the Monte Carlo programs on the different computers, normalized to one for Eltran2 on the Macintosh-SE or Macintosh-Plus computer. Higher values refer to faster running times proportionally. Since Eltran2 is a one-dimensional program, it calculates energy deposited in a semi-infinite multilayer slab. Eltran2 has been modified to a two-dimensional program called Eltran3 to computer more accurately the case with a point source, a small detector, and a short source-to-detector distance. The running time of Eltran3 is about twice as long as that of Eltran2 for a similar case

Range calculations using multigroup transport methods

International Nuclear Information System (INIS)

Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

1979-01-01

Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems

Method and computer program product for maintenance and modernization backlogging

Science.gov (United States)

Mattimore, Bernard G; Reynolds, Paul E; Farrell, Jill M

2013-02-19

According to one embodiment, a computer program product for determining future facility conditions includes a computer readable medium having computer readable program code stored therein. The computer readable program code includes computer readable program code for calculating a time period specific maintenance cost, for calculating a time period specific modernization factor, and for calculating a time period specific backlog factor. Future facility conditions equal the time period specific maintenance cost plus the time period specific modernization factor plus the time period specific backlog factor. In another embodiment, a computer-implemented method for calculating future facility conditions includes calculating a time period specific maintenance cost, calculating a time period specific modernization factor, and calculating a time period specific backlog factor. Future facility conditions equal the time period specific maintenance cost plus the time period specific modernization factor plus the time period specific backlog factor. Other embodiments are also presented.

Core status computing system

International Nuclear Information System (INIS)

Yoshida, Hiroyuki.

1982-01-01

Purpose: To calculate power distribution, flow rate and the like in the reactor core with high accuracy in a BWR type reactor. Constitution: Total flow rate signals, traverse incore probe (TIP) signals as the neutron detector signals, thermal power signals and pressure signals are inputted into a process computer, where the power distribution and the flow rate distribution in the reactor core are calculated. A function generator connected to the process computer calculates the absolute flow rate passing through optional fuel assemblies using, as variables, flow rate signals from the introduction part for fuel assembly flow rate signals, data signals from the introduction part for the geometrical configuration data at the flow rate measuring site of fuel assemblies, total flow rate signals for the reactor core and the signals from the process computer. Numerical values thus obtained are given to the process computer as correction signals to perform correction for the experimental data. (Moriyama, K.)

Heavy nucleus resonant absorption calculation benchmarks

International Nuclear Information System (INIS)

Tellier, H.; Coste, H.; Raepsaet, C.; Van der Gucht, C.

1993-01-01

The calculation of the space and energy dependence of the heavy nucleus resonant absorption in a heterogeneous lattice is one of the hardest tasks in reactor physics. Because of the computer time and memory needed, it is impossible to represent finely the cross-section behavior in the resonance energy range for everyday computations. Consequently, reactor physicists use a simplified formalism, the self-shielding formalism. As no clean and detailed experimental results are available to validate the self-shielding calculations, Monte Carlo computations are used as a reference. These results, which were obtained with the TRIPOLI continuous-energy Monte Carlo code, constitute a set of numerical benchmarks than can be used to evaluate the accuracy of the techniques or formalisms that are included in any reactor physics codes. Examples of such evaluations, for the new assembly code APOLLO2 and the slowing-down code SECOL, are given for cases of 238 U and 232 Th fuel elements

Efforts to transform computers reach milestone

CERN Multimedia

Johnson, G

2001-01-01

Scientists in San Jose, Californina, have performed the most complex calculation ever using a quantum computer - factoring the number 15. In contast to the switches in conventional computers, which although tiny consist of billions of atoms, quantum computations are carried out by manipulating single atoms. The laws of quantum mechanics which govern these actions in fact mean that multiple computations could be done in parallel, this would drastically cut down the time needed to carry out very complex calculations.

Study of high-performance canonical molecular orbitals calculation for proteins

Science.gov (United States)

Hirano, Toshiyuki; Sato, Fumitoshi

2017-11-01

The canonical molecular orbital (CMO) calculation can help to understand chemical properties and reactions in proteins. However, it is difficult to perform the CMO calculation of proteins because of its self-consistent field (SCF) convergence problem and expensive computational cost. To certainly obtain the CMO of proteins, we work in research and development of high-performance CMO applications and perform experimental studies. We have proposed the third-generation density-functional calculation method of calculating the SCF, which is more advanced than the FILE and direct method. Our method is based on Cholesky decomposition for two-electron integrals calculation and the modified grid-free method for the pure-XC term evaluation. By using the third-generation density-functional calculation method, the Coulomb, the Fock-exchange, and the pure-XC terms can be given by simple linear algebraic procedure in the SCF loop. Therefore, we can expect to get a good parallel performance in solving the SCF problem by using a well-optimized linear algebra library such as BLAS on the distributed memory parallel computers. The third-generation density-functional calculation method is implemented to our program, ProteinDF. To achieve computing electronic structure of the large molecule, not only overcoming expensive computation cost and also good initial guess for safe SCF convergence are required. In order to prepare a precise initial guess for the macromolecular system, we have developed the quasi-canonical localized orbital (QCLO) method. The QCLO has the characteristics of both localized and canonical orbital in a certain region of the molecule. We have succeeded in the CMO calculations of proteins by using the QCLO method. For simplified and semi-automated calculation of the QCLO method, we have also developed a Python-based program, QCLObot.

Numerical method for three dimensional steady-state two-phase flow calculations

International Nuclear Information System (INIS)

Raymond, P.; Toumi, I.

1992-01-01

This paper presents the numerical scheme which was developed for the FLICA-4 computer code to calculate three dimensional steady state two phase flows. This computer code is devoted to steady state and transient thermal hydraulics analysis of nuclear reactor cores 1,3 . The first section briefly describes the FLICA-4 flow modelling. Then in order to introduce the numerical method for steady state computations, some details are given about the implicit numerical scheme based upon an approximate Riemann solver which was developed for calculation of flow transients. The third section deals with the numerical method for steady state computations, which is derived from this previous general scheme and its optimization. We give some numerical results for steady state calculations and comparisons on required CPU time and memory for various meshing and linear system solvers

Shielding calculation for treatment rooms of high energy linear accelerator

International Nuclear Information System (INIS)

Elleithy, M.A.

2006-01-01

A review of German Institute of Standardization (DIN) scheme of the shielding calculation and the essential data required has been done for X-rays and electron beam in the energy range from 1 MeV to 50 MeV. Shielding calculation was done for primary and secondary radiations generated during X-ray operation of Linac. In addition, shielding was done against X-rays generated (Bremsstrahlung) by useful electron beams. The calculations also covered the neutrons generated from the interactions of useful X-rays (at energies above 8 MeV) with the surrounding. The present application involved the computation of shielding against the double scattered components of X-rays and neutrons in the maze area and the thickness of the paraffin wax of the room door. A new developed computer program was designed to assist shielding thickness calculations for a new Linac installation or in replacing an existing machine. The program used a combination of published tables and figures in computing the shielding thickness at different locations for all possible radiation situations. The DIN published data of 40 MeV accelerator room was compared with the program calculations. It was found that there is good agreement between both calculations. The developed program improved the accuracy and speed of calculation

Development of 2-D/1-D fusion method for three-dimensional whole-core heterogeneous neutron transport calculations

International Nuclear Information System (INIS)

Lee, Gil Soo

2006-02-01

To describe power distribution and multiplication factor of a reactor core accurately, it is necessary to perform calculations based on neutron transport equation considering heterogeneous geometry and scattering angles. These calculations require very heavy calculations and were nearly impossible with computers of old days. From the limitation of computing power, traditional approach of reactor core design consists of heterogeneous transport calculation in fuel assembly level and whole core diffusion nodal calculation with assembly homogenized properties, resulting from fuel assembly transport calculation. This approach may be effective in computation time, but it gives less accurate results for highly heterogeneous problems. As potential for whole core heterogeneous transport calculation became more feasible owing to rapid development of computing power during last several years, the interests in two and three dimensional whole core heterogeneous transport calculations by deterministic method are increased. For two dimensional calculation, there were several successful approaches using even parity transport equation with triangular meshes, S N method with refined rectangular meshes, the method of characteristics (MOC) with unstructured meshes, and so on. The work in this thesis originally started from the two dimensional whole core heterogeneous transport calculation by using MOC. After successful achievement in two dimensional calculation, there were efforts in three-dimensional whole-core heterogeneous transport calculation using MOC. Since direct extension to three dimensional calculation of MOC requires too much computing power, indirect approach to three dimensional calculation was considered.Thus, 2D/1D fusion method for three dimensional heterogeneous transport calculation was developed and successfully implemented in a computer code. The 2D/1D fusion method is synergistic combination of the MOC for radial 2-D calculation and S N -like methods for axial 1

A guide to the use of TIRION. A computer programme for the calculation of the consequences of releasing radioactive material to the atmosphere

International Nuclear Information System (INIS)

Kaiser, G.D.

1976-11-01

A brief description is given of the contents of TIRION, which is a computer program that has been written for use in calculations of the consequences of releasing radioactive material to the atmosphere. This is followed by a section devoted to an account of the control and data cards that make up the input to TIRION. (author)

Finite difference program for calculating hydride bed wall temperature profiles

International Nuclear Information System (INIS)

Klein, J.E.

1992-01-01

A QuickBASIC finite difference program was written for calculating one dimensional temperature profiles in up to two media with flat, cylindrical, or spherical geometries. The development of the program was motivated by the need to calculate maximum temperature differences across the walls of the Tritium metal hydrides beds for thermal fatigue analysis. The purpose of this report is to document the equations and the computer program used to calculate transient wall temperatures in stainless steel hydride vessels. The development of the computer code was motivated by the need to calculate maximum temperature differences across the walls of the hydrides beds in the Tritium Facility for thermal fatigue analysis

Study on availability of GPU for scientific and engineering calculations

International Nuclear Information System (INIS)

Sakamoto, Kensaku; Kobayashi, Seiji

2009-07-01

Recently, the number of scientific and engineering calculations on GPUs (Graphic Processing Units) is increasing. It is said that GPUs have much higher peak floating-point processing power and memory bandwidth than CPUs (Central Processing Units). We have studied the effectiveness of GPUs by applying them to fundamental scientific and engineering calculations with CUDA (Compute Unified Device Architecture) development tools. The results have shown as follows: 1) Computations on GPUs are effective for such calculations as matrix operation, FFT (Fast Fourier Transform) and CFD (Computational Fluid Dynamics) in nuclear research region. 2) Highly-advanced programming is required for bringing out high performance of GPUs. 3) Double-precision performance is low and ECC (Error Correction Code) in graphic memory systems supports are lacking. (author)

42 CFR 403.253 - Calculation of benefits.

Science.gov (United States)

2010-10-01

... the gross premiums are computed to provide coverage. (iv) Reserve for future contingent benefits means... the loss ratio calculation period. (iii) Net premium means the level portion of the gross premium used... period, to— (B) The total policy reserve at the last day of the loss ratio calculation period: and (ii...

GPU based acceleration of first principles calculation

International Nuclear Information System (INIS)

Tomono, H; Tsumuraya, K; Aoki, M; Iitaka, T

2010-01-01

We present a Graphics Processing Unit (GPU) accelerated simulations of first principles electronic structure calculations. The FFT, which is the most time-consuming part, is about 10 times accelerated. As the result, the total computation time of a first principles calculation is reduced to 15 percent of that of the CPU.

Computer-aided design system for a complex of problems on calculation and analysis of engineering and economical indexes of NPP power units

International Nuclear Information System (INIS)

Stepanov, V.I.; Koryagin, A.V.; Ruzankov, V.N.

1988-01-01

Computer-aided design system for a complex of problems concerning calculation and analysis of engineering and economical indices of NPP power units is described. In the system there are means for automated preparation and debugging of data base software complex, which realizes th plotted algorithm in the power unit control system. Besides, in the system there are devices for automated preparation and registration of technical documentation

A computationally efficient software application for calculating vibration from underground railways

International Nuclear Information System (INIS)

Hussein, M F M; Hunt, H E M

2009-01-01

The PiP model is a software application with a user-friendly interface for calculating vibration from underground railways. This paper reports about the software with a focus on its latest version and the plans for future developments. The software calculates the Power Spectral Density of vibration due to a moving train on floating-slab track with track irregularity described by typical values of spectra for tracks with good, average and bad conditions. The latest version accounts for a tunnel embedded in a half space by employing a toolbox developed at K.U. Leuven which calculates Green's functions for a multi-layered half-space.

SARTEMP2 - A computer program to calculate power and temperatures in a transport flask during a criticality accident

International Nuclear Information System (INIS)

Shaw, P.M.

1983-04-01

The computer code SARTEMP2, an extended version of the original SARTEMP program, which calculates the power and temperatures in a transport flask during a hypothetical criticality accident is described. The accident arises, it is assumed, during the refilling of the flask with water, bringing the system to delayed critical. As the water level continues to rise, reactivity is added causing the power to rise, and thus temperatures in the fuel, clad and water to increase. The point kinetics equations are coupled to the one-dimensional heat conduction equation. The model used, the method of solution of the equations and the input data required are given. (author)

Calculating additional shielding requirements in diagnostics X-ray departments by computer

International Nuclear Information System (INIS)

Rahimi, A.

2004-01-01

This report provides an extension of an existing method for the calculation of the barrier thickness required to reduce the three types of radiation exposure emitted from the source, the primary, secondary and leakage radiation, to a specified weekly design limit (MPD). Since each of these three types of radiation are of different beam quality, having different shielding requirements, NCRP 49 has provided means to calculate the necessary protective barrier thickness for each type of radiation individually. Additionally, barrier requirements specified using the techniques stated at NCRP 49, show enormous variations among users. Part of the variations is due to different assumptions made regarding the use of the examined room and the characteristics of adjoining space. Many of the differences result from the difficulty of accurately relating information from the calculations to graphs and tables involved in the calculation process specified by this report. Moreover, the latest technological developments such as mammography are not addressed and attenuation data for three-phase generators, that are most widely used today, is not provided. The design of shielding barriers in diagnostic X-ray departments generally follow the ALARA principle. That means that, in practice, the exposure levels are kept 'as low as reasonably achievable', taking into account economical and technical factors. Additionally, the calculation of barrier requirements includes many uncertainties (e.g. the workload, the actual kVp used etc.). (author)

Direct Computation on the Kinetic Spectrophotometry

DEFF Research Database (Denmark)

Hansen, Jørgen-Walther; Broen Pedersen, P.

1974-01-01

This report describes an analog computer designed for calculations of transient absorption from photographed recordings of the oscilloscope trace of the transmitted light intensity. The computer calculates the optical density OD, the natural logarithm of OD, and the natural logarithm of the diffe...

Improvements in EBR-2 core depletion calculations

International Nuclear Information System (INIS)

Finck, P.J.; Hill, R.N.; Sakamoto, S.

1991-01-01

The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs

Computer code for double beta decay QRPA based calculations

Energy Technology Data Exchange (ETDEWEB)

Barbero, C. A.; Mariano, A. [Departamento de Física, Facultad de Ciencias Exactas, Universidad Nacional de La Plata, La Plata, Argentina and Instituto de Física La Plata, CONICET, La Plata (Argentina); Krmpotić, F. [Instituto de Física La Plata, CONICET, La Plata, Argentina and Instituto de Física Teórica, Universidade Estadual Paulista, São Paulo (Brazil); Samana, A. R.; Ferreira, V. dos Santos [Departamento de Ciências Exatas e Tecnológicas, Universidade Estadual de Santa Cruz, BA (Brazil); Bertulani, C. A. [Department of Physics, Texas A and M University-Commerce, Commerce, TX (United States)

2014-11-11

The computer code developed by our group some years ago for the evaluation of nuclear matrix elements, within the QRPA and PQRPA nuclear structure models, involved in neutrino-nucleus reactions, muon capture and β{sup ±} processes, is extended to include also the nuclear double beta decay.

COMPUTATIONAL EFFICIENCY OF A MODIFIED SCATTERING KERNEL FOR FULL-COUPLED PHOTON-ELECTRON TRANSPORT PARALLEL COMPUTING WITH UNSTRUCTURED TETRAHEDRAL MESHES

Directory of Open Access Journals (Sweden)

JONG WOON KIM

2014-04-01

In this paper, we introduce a modified scattering kernel approach to avoid the unnecessarily repeated calculations involved with the scattering source calculation, and used it with parallel computing to effectively reduce the computation time. Its computational efficiency was tested for three-dimensional full-coupled photon-electron transport problems using our computer program which solves the multi-group discrete ordinates transport equation by using the discontinuous finite element method with unstructured tetrahedral meshes for complicated geometrical problems. The numerical tests show that we can improve speed up to 17∼42 times for the elapsed time per iteration using the modified scattering kernel, not only in the single CPU calculation but also in the parallel computing with several CPUs.

Method for dose calculation in intracavitary irradiation of endometrical carcinoma

International Nuclear Information System (INIS)

Zevrieva, I.F.; Ivashchenko, N.T.; Musapirova, N.A.; Fel'dman, S.Z.; Sajbekov, T.S.

1979-01-01

A method for dose calculation for the conditions of intracavitary gamma therapy of endometrial carcinoma using spherical and linear 60 Co sources was elaborated. Calculations of dose rates for different amount and orientation of spherical radiation sources and for different planes were made with the aid of BEhSM-4M computer. Dosimet were made with the aid of BEhSM-4M computer. Dosimetric study of dose fields was made using a phantom imitating the real conditions of irradiation. Discrepancies between experimental and calculated values are within the limits of the experiment accuracy

Quantum Computations: Fundamentals and Algorithms

International Nuclear Information System (INIS)

Duplij, S.A.; Shapoval, I.I.

2007-01-01

Basic concepts of quantum information theory, principles of quantum calculations and the possibility of creation on this basis unique on calculation power and functioning principle device, named quantum computer, are concerned. The main blocks of quantum logic, schemes of quantum calculations implementation, as well as some known today effective quantum algorithms, called to realize advantages of quantum calculations upon classical, are presented here. Among them special place is taken by Shor's algorithm of number factorization and Grover's algorithm of unsorted database search. Phenomena of decoherence, its influence on quantum computer stability and methods of quantum errors correction are described

A user`s guide to LUGSAN II. A computer program to calculate and archive lug and sway brace loads for aircraft-carried stores

Energy Technology Data Exchange (ETDEWEB)

Dunn, W.N. [Sandia National Labs., Albuquerque, NM (United States). Mechanical and Thermal Environments Dept.

1998-03-01

LUG and Sway brace ANalysis (LUGSAN) II is an analysis and database computer program that is designed to calculate store lug and sway brace loads for aircraft captive carriage. LUGSAN II combines the rigid body dynamics code, SWAY85, with a Macintosh Hypercard database to function both as an analysis and archival system. This report describes the LUGSAN II application program, which operates on the Macintosh System (Hypercard 2.2 or later) and includes function descriptions, layout examples, and sample sessions. Although this report is primarily a user`s manual, a brief overview of the LUGSAN II computer code is included with suggested resources for programmers.

Development of computer aided engineering system for TRAC applications

International Nuclear Information System (INIS)

Arai, Kenji; Itoya, Seihiro; Uematsu, Hitoshi; Tsunoyama, Shigeaki

1990-01-01

An advanced best estimate computer program for nuclear reactor transient analysis, TRAC has been extensively used to carry out various thermal hydraulic calculations in the nuclear engineering field, because of its versatility. To perform efficiently a wide variety of TRAC calculation, the efficient utilization of computers and the convenient environment for input and output processing is necessary. We have applied a computer network comprising a super-computer, engineering work stations and personal computers to TRAC calculations and have assigned the appropriate functions to each computer. We have also been developing an interactive graphics system for input and output processing on an EWS. This hardware and software environment can improve the effectiveness of TRAC utilization for various thermal hydraulic calculations. (author)

Monte Carlo calculations on a parallel computer using MORSE-C.G

International Nuclear Information System (INIS)

Wood, J.

1995-01-01

The general purpose particle transport Monte Carlo code, MORSE-C.G., is implemented on a parallel computing transputer-based system having MIMD architecture. Example problems are solved which are representative of the 3-principal types of problem that can be solved by the original serial code, namely, fixed source, eigenvalue (k-eff) and time-dependent. The results from the parallelized version of the code are compared in tables with the serial code run on a mainframe serial computer, and with an independent, deterministic transport code. The performance of the parallel computer as the number of processors is varied is shown graphically. For the parallel strategy used, the loss of efficiency as the number of processors is increased, is investigated. (author)

Criticality calculations for safety analysis

International Nuclear Information System (INIS)

Vellozo, S.O.

1981-01-01

Criticality studies in uranium nitrate and plutonium nitrate aqueous solutions were done. For uranium compound three basic computer codes are used: GAMTEC-II, DTF-IV, KENO-IV. Water was used as refletor and the results obtained with the different computer codes were analyzed and compared with the 'Handbuck zur Kriticalitat'. The cross sections and the cylindrical geometry were generated by Gamtec-II computer code. In the second compound the thickness of the recipient with plutonium nitrate are used with rectangular geometry and concret reflector. The effective multiplication constant was calculated with the Gamtec-II and Keno-IV library. The results show many differences. (E.G) [pt

Radiation protection calculations for diagnostic medical equipment

International Nuclear Information System (INIS)

Klueter, R.

1992-01-01

The standards DIN 6812 and DIN 6844 define the radiation protection requirements to be met by biomedical radiography equipment or systems for nuclear medicine. The paper explains the use of a specific computer program for radiation protection calculations. The program offers menu-controlled calculation, with free choice of the relevant nuclides. (DG) [de

Computational Model of D-Region Ion Production Caused by Energetic Electron Precipitations Based on General Monte Carlo Transport Calculations

Science.gov (United States)

Kouznetsov, A.; Cully, C. M.

2017-12-01

During enhanced magnetic activities, large ejections of energetic electrons from radiation belts are deposited in the upper polar atmosphere where they play important roles in its physical and chemical processes, including VLF signals subionospheric propagation. Electron deposition can affect D-Region ionization, which are estimated based on ionization rates derived from energy depositions. We present a model of D-region ion production caused by an arbitrary (in energy and pitch angle) distribution of fast (10 keV - 1 MeV) electrons. The model relies on a set of pre-calculated results obtained using a general Monte Carlo approach with the latest version of the MCNP6 (Monte Carlo N-Particle) code for the explicit electron tracking in magnetic fields. By expressing those results using the ionization yield functions, the pre-calculated results are extended to cover arbitrary magnetic field inclinations and atmospheric density profiles, allowing ionization rate altitude profile computations in the range of 20 and 200 km at any geographic point of interest and date/time by adopting results from an external atmospheric density model (e.g. NRLMSISE-00). The pre-calculated MCNP6 results are stored in a CDF (Common Data Format) file, and IDL routines library is written to provide an end-user interface to the model.

Calculation of resonance integral for fuel cluster

International Nuclear Information System (INIS)

Remsak, S.

1969-01-01

The procedure for calculating the shielding correction, formulated in the previous paper [6], was broadened and applied for a cluster of cylindrical rods. The sam analytical method as in the previous paper was applied. A combination of Gauss method with the method of Almgren and Porn used for solving the same type of integral was used to calculate the geometry functions. CLUSTER code was written for ZUSE-Z-23 computer to calculate the shielding corrections for pairs of fuel rods in the cluster. Computing time for one pair of fuel rods depends on the number of closely placed rod, and for two closely placed rods it is about 3 hours. Calculations were done for clusters containing 7 and 19 UO 2 rods. results show that calculated values of resonance integrals are somewhat higher than the values obtained by Helstrand empirical formula. Taking into account the results for two rods from the previous paper it can be noted that the calculated and empirical values for clusters with 2 and 7 rods are in agreement since the deviations do not exceed the limits of experimental error (±2%). In case of larger cluster with 19 rods deviations are higher than the experimental error. Most probably the calculated values exceed the experimental ones result from the fact that in this paper the shielding correction is calculated only in the region up to 1 keV [sr

Massively parallel sparse matrix function calculations with NTPoly

Science.gov (United States)

Dawson, William; Nakajima, Takahito

2018-04-01

We present NTPoly, a massively parallel library for computing the functions of sparse, symmetric matrices. The theory of matrix functions is a well developed framework with a wide range of applications including differential equations, graph theory, and electronic structure calculations. One particularly important application area is diagonalization free methods in quantum chemistry. When the input and output of the matrix function are sparse, methods based on polynomial expansions can be used to compute matrix functions in linear time. We present a library based on these methods that can compute a variety of matrix functions. Distributed memory parallelization is based on a communication avoiding sparse matrix multiplication algorithm. OpenMP task parallellization is utilized to implement hybrid parallelization. We describe NTPoly's interface and show how it can be integrated with programs written in many different programming languages. We demonstrate the merits of NTPoly by performing large scale calculations on the K computer.

When computers were human

CERN Document Server

Grier, David Alan

2013-01-01

Before Palm Pilots and iPods, PCs and laptops, the term ""computer"" referred to the people who did scientific calculations by hand. These workers were neither calculating geniuses nor idiot savants but knowledgeable people who, in other circumstances, might have become scientists in their own right. When Computers Were Human represents the first in-depth account of this little-known, 200-year epoch in the history of science and technology. Beginning with the story of his own grandmother, who was trained as a human computer, David Alan Grier provides a poignant introduction to the wider wo

A Scientific Calculator for Exact Real Number Computation Based on LRT, GMP and FC++

Directory of Open Access Journals (Sweden)

J. A. Hernández

2012-03-01

Full Text Available Language for Redundant Test (LRT is a programming language for exact real number computation. Its lazy evaluation mechanism (also called call-by-need and its infinite list requirement, make the language appropriate to be implemented in a functional programming language such as Haskell. However, a direction translation of the operational semantics of LRT into Haskell as well as the algorithms to implement basic operations (addition subtraction, multiplication, division and trigonometric functions (sin, cosine, tangent, etc. makes the resulting scientific calculator time consuming and so inefficient. In this paper, we present an alternative implementation of the scientific calculator using FC++ and GMP. FC++ is a functional C++ library while GMP is a GNU multiple presicion library. We show that a direct translation of LRT in FC++ results in a faster scientific calculator than the one presented in Haskell.El lenguaje de verificación redundante (LRT, por sus siglas en inglés es un lenguaje de programación para el cómputo con números reales exactos. Su método de evaluación lazy (o mejor conocido como llamada por necesidad y el manejo de listas infinitas requerido, hace que el lenguaje sea apropiado para su implementación en un lenguaje funcional como Haskell. Sin embargo, la implementación directa de la semántica operacional de LRT en Haskell así como los algoritmos para funciones básicas (suma, resta, multiplicación y división y funciones trigonométricas (seno, coseno, tangente, etc hace que la calculadora científica resultante sea ineficiente. En este artículo, presentamos una implementación alternativa de la calculadora científica usando FC++ y GMP. FC++ es una librería que utiliza el paradigma Funcional en C++ mientras que GMP es una librería GNU de múltiple precisión. En el artículo mostramos que la implementación directa de LRT en FC++ resulta en una librería más eficiente que la implementada en Haskell.

Coil protection calculator for TFTR

International Nuclear Information System (INIS)

Marsala, R.J.; Lawson, J.E.; Persing, R.G.; Senko, T.R.; Woolley, R.D.

1989-01-01

A new coil protection system (CPS) is being developed to replace the existing TFTR magnetic coil fault detector. The existing fault detector sacrifices TFTR operating capability for simplicity. The new CPS, when installed in October of 1988, will permit operation up to the actual coil stress limits parameters in real-time. The computation will be done in a microprocessor based Coil Protection Calculator (CPC) currently under construction at PPL. THe new CPC will allow TFTR to operate with higher plasma currents and will permit the optimization of pulse repetition rates. The CPC will provide real-time estimates of critical coil and bus temperatures and stresses based on real-time redundant measurements of coil currents, coil cooling water inlet temperature, and plasma current. The critical parameter calculations are compared to prespecified limits. If these limits are reached or exceeded, protection action will be initiated to a hard wired control system (HCS), which will shut down the power supplies. The CPC consists of a redundant VME based microprocessor system which will sample all input data and compute all stress quantities every ten milliseconds. Thermal calculations will be approximated every 10ms with an exact solution occurring every second. The CPC features continuous cross-checking of redundant input signal, automatic detection of internal failure modes, monitoring and recording of calculated results, and a quick, functional verification of performance via an internal test system. (author)

A computer code for fault tree calculations: PATREC

International Nuclear Information System (INIS)

Blin, A.; Carnino, A.; Koen, B.V.; Duchemin, B.; Lanore, J.M.; Kalli, H.

1978-01-01

A computer code for evaluating the reliability of complex system by fault tree is described in this paper. It uses pattern recognition approach and programming techniques from IBM PL1 language. It can take account of many of the present day problems: multi-dependencies treatment, dispersion in the reliability data parameters, influence of common mode failures. The code is running currently since two years now in Commissariat a l'Energie Atomique Saclay center and shall be used in a future extension for automatic fault trees construction

On the kinetics of the aluminum-water reaction during exposure in high-heat flux test loops: 1, A computer program for oxidation calculations

International Nuclear Information System (INIS)

Pawel, R.E.

1988-01-01

The ''Griess Correlation,'' in which the thickness of the corrosion product on aluminum alloy surfaces is expressed as a function of time and temperature for high-flux-reactor conditions, was rewritten in the form of a simple, general rate equation. Based on this equation, a computer program that calculates oxide-layer thickness for any given time-temperature transient was written. 4 refs

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.

Advances in computational methods for Quantum Field Theory calculations

NARCIS (Netherlands)

Ruijl, B.J.G.

2017-01-01

In this work we describe three methods to improve the performance of Quantum Field Theory calculations. First, we simplify large expressions to speed up numerical integrations. Second, we design Forcer, a program for the reduction of four-loop massless propagator integrals. Third, we extend the R*

Configuration space Faddeev calculations

International Nuclear Information System (INIS)

Payne, G.L.; Klink, W.H.; Polyzou, W.N.

1989-01-01

The detailed study of few-body systems provides one of the most effective means for studying nuclear physics at subnucleon distance scales. For few-body systems the model equations can be solved numerically with errors less than the experimental uncertainties. We have used such systems to investigate the size of relativistic effects, the role of meson-exchange currents, and the importance of quark degrees of freedom in the nucleus. Complete calculations for momentum-dependent potentials have been performed, and the properties of the three-body bound state for these potentials have been studied. Few-body calculations of the electromagnetic form factors of the deuteron and pion have been carried out using a front-form formulation of relativistic quantum mechanics. The decomposition of the operators transforming convariantly under the Poincare group into kinematical and dynamical parts has been studies. New ways for constructing interactions between particles, as well as interactions which lead to the production of particles, have been constructed in the context of a relativistic quantum mechanics. To compute scattering amplitudes in a nonperturbative way, classes of operators have been generated out of which the phase operator may be constructed. Finally, we have worked out procedures for computing Clebsch-Gordan and Racah coefficients on a computer, as well as giving procedures for dealing with the multiplicity problem

Argosy 4 - A programme for lattice calculations

International Nuclear Information System (INIS)

MacDougall, J.D.

1965-07-01

This report contains a detailed description of the methods of calculation used in the Argosy 4 computer programme, and of the input requirements and printed results produced by the programme. An outline of the physics of the Argosy method is given. Section 2 describes the lattice calculation, including the burn up calculation, section 3 describes the control rod calculation and section 4 the reflector calculation. In these sections the detailed equations solved by the programme are given. In section 5 input requirements are given, and in section 6 the printed output obtained from an Argosy calculation is described. In section 7 are noted the principal differences between Argosy 4 and earlier versions of the Argosy programme

Reactor calculations for improving utilization of TRIGA reactor

International Nuclear Information System (INIS)

Ravnik, M.

1986-01-01

A brief review of our work on reactor calculations of 250 kW TRIGA with mixed core (standard + FLIP fuel) will be presented. The following aspects will be treated: - development of computer programs; - optimization of in-core fuel management with respect to fuel costs and irradiation channels utilization. TRIGAP programme package will be presented as an example of computer programs. It is based on 2-group 1-D diffusion approximation and besides calculations offers possibilities for operational data logging and fuel inventory book-keeping as well. It is developed primarily for the research reactor operators as a tool for analysing reactor operation and fuel management. For this reason it is arranged for a small (PC) computer. Second part will be devoted to reactor physics properties of the mixed cores. Results of depletion calculations will be presented together with measured data to confirm some general guidelines for optimal mixed core fuel management. As the results are obtained using TRIGAP program package results can be also considered as an illustration and qualification for its application. (author)

Cloud Computing Quality

Directory of Open Access Journals (Sweden)

Anamaria Şiclovan

2013-02-01

Full Text Available Cloud computing was and it will be a new way of providing Internet services and computers. This calculation approach is based on many existing services, such as the Internet, grid computing, Web services. Cloud computing as a system aims to provide on demand services more acceptable as price and infrastructure. It is exactly the transition from computer to a service offered to the consumers as a product delivered online. This paper is meant to describe the quality of cloud computing services, analyzing the advantages and characteristics offered by it. It is a theoretical paper.Keywords: Cloud computing, QoS, quality of cloud computing

Computers and data processing

CERN Document Server

Deitel, Harvey M

1985-01-01

Computers and Data Processing provides information pertinent to the advances in the computer field. This book covers a variety of topics, including the computer hardware, computer programs or software, and computer applications systems.Organized into five parts encompassing 19 chapters, this book begins with an overview of some of the fundamental computing concepts. This text then explores the evolution of modern computing systems from the earliest mechanical calculating devices to microchips. Other chapters consider how computers present their results and explain the storage and retrieval of

Charged-particle calculations using Boltzmann transport methods

International Nuclear Information System (INIS)

Hoffman, T.J.; Dodds, H.L. Jr.; Robinson, M.T.; Holmes, D.K.

1981-01-01

Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of charged particle range distributions, reflection coefficients, and sputtering yields. The Boltzmann transport approach can be implemented, with minor changes, in standard neutral particle computer codes. With the multigroup discrete ordinates code, ANISN, determination of ion and target atom distributions as functions of position, energy, and direction can be obtained without the stochastic error associated with atomistic computer codes such as MARLOWE and TRIM. With the multigroup Monte Carlo code, MORSE, charged particle effects can be obtained for problems associated with very complex geometries. Results are presented for several charged particle problems. Good agreement is obtained between quantities calculated with the multigroup approach and those obtained experimentally or by atomistic computer codes

Assessment of effectiveness of geologic isolation systems. ARRRG and FOOD: computer programs for calculating radiation dose to man from radionuclides in the environment

International Nuclear Information System (INIS)

Napier, B.A.; Roswell, R.L.; Kennedy, W.E. Jr.; Strenge, D.L.

1980-06-01

The computer programs ARRRG and FOOD were written to facilitate the calculation of internal radiation doses to man from the radionuclides in the environment and external radiation doses from radionuclides in the environment. Using ARRRG, radiation doses to man may be calculated for radionuclides released to bodies of water from which people might obtain fish, other aquatic foods, or drinking water, and in which they might fish, swim or boat. With the FOOD program, radiation doses to man may be calculated from deposition on farm or garden soil and crops during either an atmospheric or water release of radionuclides. Deposition may be either directly from the air or from irrigation water. Fifteen crop or animal product pathways may be chosen. ARRAG and FOOD doses may be calculated for either a maximum-exposed individual or for a population group. Doses calculated are a one-year dose and a committed dose from one year of exposure. The exposure is usually considered as chronic; however, equations are included to calculate dose and dose commitment from acute (one-time) exposure. The equations for calculating internal dose and dose commitment are derived from those given by the International Commission on Radiological Protection (ICRP) for body burdens and Maximum Permissible Concentration (MPC) of each radionuclide. The radiation doses from external exposure to contaminated farm fields or shorelines are calculated assuming an infinite flat plane source of radionuclides. A factor of two is included for surface roughness. A modifying factor to compensate for finite extent is included in the shoreline calculations

Assessment of effectiveness of geologic isolation systems. ARRRG and FOOD: computer programs for calculating radiation dose to man from radionuclides in the environment

Energy Technology Data Exchange (ETDEWEB)

Napier, B.A.; Roswell, R.L.; Kennedy, W.E. Jr.; Strenge, D.L.

1980-06-01

The computer programs ARRRG and FOOD were written to facilitate the calculation of internal radiation doses to man from the radionuclides in the environment and external radiation doses from radionuclides in the environment. Using ARRRG, radiation doses to man may be calculated for radionuclides released to bodies of water from which people might obtain fish, other aquatic foods, or drinking water, and in which they might fish, swim or boat. With the FOOD program, radiation doses to man may be calculated from deposition on farm or garden soil and crops during either an atmospheric or water release of radionuclides. Deposition may be either directly from the air or from irrigation water. Fifteen crop or animal product pathways may be chosen. ARRAG and FOOD doses may be calculated for either a maximum-exposed individual or for a population group. Doses calculated are a one-year dose and a committed dose from one year of exposure. The exposure is usually considered as chronic; however, equations are included to calculate dose and dose commitment from acute (one-time) exposure. The equations for calculating internal dose and dose commitment are derived from those given by the International Commission on Radiological Protection (ICRP) for body burdens and Maximum Permissible Concentration (MPC) of each radionuclide. The radiation doses from external exposure to contaminated farm fields or shorelines are calculated assuming an infinite flat plane source of radionuclides. A factor of two is included for surface roughness. A modifying factor to compensate for finite extent is included in the shoreline calculations.

Burnup calculation for a tokamak commercial hybrid reactor

International Nuclear Information System (INIS)

Feng Kaiming; Xie Zhongyou

1990-08-01

A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

Validation of dose calculation programmes for recycling

International Nuclear Information System (INIS)

Menon, Shankar; Brun-Yaba, Christine; Yu, Charley; Cheng, Jing-Jy; Williams, Alexander

2002-12-01

This report contains the results from an international project initiated by the SSI in 1999. The primary purpose of the project was to validate some of the computer codes that are used to estimate radiation doses due to the recycling of scrap metal. The secondary purpose of the validation project was to give a quantification of the level of conservatism in clearance levels based on these codes. Specifically, the computer codes RESRAD-RECYCLE and CERISE were used to calculate radiation doses to individuals during the processing of slightly contaminated material, mainly in Studsvik, Sweden. Calculated external doses were compared with measured data from different steps of the process. The comparison of calculations and measurements shows that the computer code calculations resulted in both overestimations and underestimations of the external doses for different recycling activities. The SSI draws the conclusion that the accuracy is within one order of magnitude when experienced modellers use their programmes to calculate external radiation doses for a recycling process involving material that is mainly contaminated with cobalt-60. No errors in the codes themselves were found. Instead, the inaccuracy seems to depend mainly on the choice of some modelling parameters related to the receptor (e.g., distance, time, etc.) and simplifications made to facilitate modelling with the codes (e.g., object geometry). Clearance levels are often based on studies on enveloping scenarios that are designed to cover all realistic exposure pathways. It is obvious that for most practical cases, this gives a margin to the individual dose constraint (in the order of 10 micro sievert per year within the EC). This may be accentuated by the use of conservative assumptions when modelling the enveloping scenarios. Since there can obviously be a fairly large inaccuracy in the calculations, it seems reasonable to consider some degree of conservatism when establishing clearance levels based on

Validation of dose calculation programmes for recycling

Energy Technology Data Exchange (ETDEWEB)

Menon, Shankar [Menon Consulting, Nykoeping (Sweden); Brun-Yaba, Christine [Inst. de Radioprotection et Securite Nucleaire (France); Yu, Charley; Cheng, Jing-Jy [Argonne National Laboratory, IL (United States). Environmental Assessment Div.; Bjerler, Jan [Studsvik Stensand, Nykoeping (Sweden); Williams, Alexander [Dept. of Energy (United States). Office of Environmental Management

2002-12-01

This report contains the results from an international project initiated by the SSI in 1999. The primary purpose of the project was to validate some of the computer codes that are used to estimate radiation doses due to the recycling of scrap metal. The secondary purpose of the validation project was to give a quantification of the level of conservatism in clearance levels based on these codes. Specifically, the computer codes RESRAD-RECYCLE and CERISE were used to calculate radiation doses to individuals during the processing of slightly contaminated material, mainly in Studsvik, Sweden. Calculated external doses were compared with measured data from different steps of the process. The comparison of calculations and measurements shows that the computer code calculations resulted in both overestimations and underestimations of the external doses for different recycling activities. The SSI draws the conclusion that the accuracy is within one order of magnitude when experienced modellers use their programmes to calculate external radiation doses for a recycling process involving material that is mainly contaminated with cobalt-60. No errors in the codes themselves were found. Instead, the inaccuracy seems to depend mainly on the choice of some modelling parameters related to the receptor (e.g., distance, time, etc.) and simplifications made to facilitate modelling with the codes (e.g., object geometry). Clearance levels are often based on studies on enveloping scenarios that are designed to cover all realistic exposure pathways. It is obvious that for most practical cases, this gives a margin to the individual dose constraint (in the order of 10 micro sievert per year within the EC). This may be accentuated by the use of conservative assumptions when modelling the enveloping scenarios. Since there can obviously be a fairly large inaccuracy in the calculations, it seems reasonable to consider some degree of conservatism when establishing clearance levels based on

Study on the acceleration of the neutronics calculation based on GPGPU

International Nuclear Information System (INIS)

Ohoka, Y.; Tatsumi, M.

2007-01-01

The cost of the reactor physics calculation tends to become higher with more detail treatment in the physics models and computational algorithms. For example, SCOPE2 requires considerably high computational costs for multi-group transport calculation in 3-D pin-by-pin geometry. In this paper, applicability of GPGPU to acceleration of neutronics calculation is discussed. At first, performance and accuracy of the basic matrix calculations with fundamental arithmetic operators and the exponential, function are studied. The calculation was performed on a machine with Pentium 4 of 3.2 MHz and GPU of nVIDIA GeForce7800GTX using a test program written in C++, OpenGL and GLSL on Linux. When matrix size becomes large, the calculation on GPU is 10-50 times faster than that on CPU for fundamental arithmetic operators. For the exponential function, calculation on GPU is 270-370 times faster than that on CPU. The precision of all the cases are equivalent to that on CPU, which is less than the criterion of IEEE754 (10 -6 as single precision). Next, the GPGPU is applied to a functional module in SCOPE2. In the present study, as the first step of GPGPU application, calculations in. small geometry are tested. Performance gain, by GPGPU in this application was relatively modest, approximately 15%, compared to the feasibility study. This is because the part in which GPGPU was applied had appropriate structure for GPGPU implementation but had only small fraction of computational load. For much advanced acceleration, it is important to consider various factors such as easiness of implementation, fraction of computational load and bottleneck in data transfer between GPU and CPU. (authors)

An EGS4-ready tomographic computational model of a 14-year-old female torso for calculating organ doses from CT examinations

International Nuclear Information System (INIS)

Caon, M.; School of Physics and Electronic Systems Engineering, University of South Australia, The Levels Campus, Mawson Lakes, South Australia, 5095; Pattison, J.

1999-01-01

Fifty-four consecutive CT scans have been used to construct a tomographic computational model of a 14-year-old female torso suitable for the determination of organ doses from CT. The model, known as ADELAIDE, is in the form of an input file compatible with user codes based on XYZDOS.MOR from the readily available EGS4 Monte Carlo radiation transport code. ADELAIDE's dimensions are close to the Australian averages for her age so the model is representative of a 14-year-old girl. The realistic anatomy in the model differs considerably from that in Cristy's 15-year-old mathematical computational model by having realistically shaped organs that are appropriately located within a real external contour. Average absorbed dose to organs from simulated CT examinations of the chest and abdomen have been calculated for ADELAIDE using EGS4 within a geometry specific to the General Electric Hi-Speed Advantage CT scanner and using an x-ray spectrum calculated using data from the scanner's x-ray tube. The simulations include the scanner's beam shaping filter and patient table. It is suggested that the resulting values have fewer possible sources of uncertainty than organ doses derived from dose coefficients calculated for a MIRD style model with mathematical anatomy and a spectrum that may not match that of the scanner. The organ doses were normalized using the scanner's CTDI measured free-in-air and an EGS4 simulation of the CTDI measurement. Effective dose to the torso from 26-slice chest and 24-slice abdomen examinations (at 120 kV, 200 mAs, 7 mm slices) is 4.6±0.1mSv and 4.3±0.1mSv respectively. (author)

Computing in plasma physics

International Nuclear Information System (INIS)

Nuehrenberg, J.

1986-01-01

These proceedings contain the articles presented at the named conference. These concern numerical methods for astrophysical plasmas, the numerical simulation of reversed-field pinch dynamics, methods for numerical simulation of ideal MHD stability of axisymmetric plasmas, calculations of the resistive internal m=1 mode in tokamaks, parallel computing and multitasking, particle simulation methods in plasma physics, 2-D Lagrangian studies of symmetry and stability of laser fusion targets, computing of rf heating and current drive in tokamaks, three-dimensional free boundary calculations using a spectral Green's function method, as well as the calculation of three-dimensional MHD equilibria with islands and stochastic regions. See hints under the relevant topics. (HSI)

Implatation of MC2 computer code

International Nuclear Information System (INIS)

Seehusen, J.; Nair, R.P.K.; Becceneri, J.C.

1981-01-01

The implantation of MC2 computer code in the CDC system is presented. The MC2 computer code calculates multigroup cross sections for tipical compositions of fast reactors. The multigroup constants are calculated using solutions of PI or BI approximations for determined buckling value as weighting function. (M.C.K.) [pt

Program realization of mathematical model of kinetostatical calculation of flat lever mechanisms

Directory of Open Access Journals (Sweden)

M. A. Vasechkin

2016-01-01

Full Text Available Global computerization determined the dominant position of the analytical methods for the study of mechanisms. As a result, kinetostatics analysis of mechanisms using software packages is an important part of scientific and practical activities of engineers and designers. Therefore, software implementation of mathematical models kinetostatical calculating mechanisms is of practical interest. The mathematical model obtained in [1]. In the language of Turbo Pascal developed a computer procedure that calculates the forces in kinematic pairs in groups Assur (GA and a balancing force at the primary level. Before use appropriate computational procedures it is necessary to know all external forces and moments acting on the GA and to determine the inertial forces and moments of inertia forces. The process of calculations and constructions of the provisions of the mechanism can be summarized as follows. Organized cycle in which to calculate the position of an initial link of the mechanism. Calculate the position of the remaining links of the mechanism by referring to relevant procedures module DIADA in GA [2,3]. Using the graphics mode of the computer displaying on the display the position of the mechanism. The computed inertial forces and moments of inertia forces. Turning to the corresponding procedures of the module, calculated all the forces in kinematic pairs and the balancing force at the primary level. In each kinematic pair build forces and their direction with the help of simple graphical procedures. The magnitude of these forces and their direction are displayed in a special window with text mode. This work contains listings of the test programs MyTеst, is an example of using computing capabilities of the developed module. As a check on the calculation procedures of module in the program is reproduced an example of calculating the balancing forces according to the method of Zhukovsky (Zhukovsky lever.

Dose Distribution Calculation Using MCNPX Code in the Gamma-ray Irradiation Cell

International Nuclear Information System (INIS)

Kim, Yong Ho

1991-02-01

60 Co-gamma irradiators have long been used for foods sterilization, plant mutation and development of radio-protective agents, radio-sensitizers and other purposes. The Applied Radiological Science Research Institute of Cheju National University has a multipurpose gamma irradiation facility loaded with a MDS Nordin standard 60 Co source (C188), of which the initial activity was 400 TBq (10,800 Ci) on February 19, 2004. This panoramic gamma irradiator is designed to irradiate in all directions various samples such as plants, cultured cells and mice to administer given radiation doses. In order to give accurate doses to irradiation samples, appropriate methods of evaluating, both by calculation and measurement, the radiation doses delivered to the samples should be set up. Computational models have been developed to evaluate the radiation dose distributions inside the irradiation chamber and the radiation doses delivered to typical biolological samples which are frequently irradiated in the facility. The computational models are based on using the MCNPX code. The horizontal and vertical dose distributions has been calculated inside the irradiation chamber and compared the calculated results with measured data obtained with radiation dosimeters to verify the computational models. The radiation dosimeters employed are a Famer's type ion chamber and MOSFET dosimeters. Radiation doses were calculated by computational models, which were delivered to cultured cell samples contained in test tubes and to a mouse fixed in a irradiation cage, and compared the calculated results with the measured data. The computation models are also tested to see if they can accurately simulate the case where a thick lead shield is placed between the source and detector. Three tally options of the MCNPX code, F4, F5 and F6, are alternately used to see which option produces optimum results. The computation models are also used to calculate gamma ray energy spectra of a BGO scintillator at

Methods for Melting Temperature Calculation

Science.gov (United States)

Hong, Qi-Jun

Melting temperature calculation has important applications in the theoretical study of phase diagrams and computational materials screenings. In this thesis, we present two new methods, i.e., the improved Widom's particle insertion method and the small-cell coexistence method, which we developed in order to capture melting temperatures both accurately and quickly. We propose a scheme that drastically improves the efficiency of Widom's particle insertion method by efficiently sampling cavities while calculating the integrals providing the chemical potentials of a physical system. This idea enables us to calculate chemical potentials of liquids directly from first-principles without the help of any reference system, which is necessary in the commonly used thermodynamic integration method. As an example, we apply our scheme, combined with the density functional formalism, to the calculation of the chemical potential of liquid copper. The calculated chemical potential is further used to locate the melting temperature. The calculated results closely agree with experiments. We propose the small-cell coexistence method based on the statistical analysis of small-size coexistence MD simulations. It eliminates the risk of a metastable superheated solid in the fast-heating method, while also significantly reducing the computer cost relative to the traditional large-scale coexistence method. Using empirical potentials, we validate the method and systematically study the finite-size effect on the calculated melting points. The method converges to the exact result in the limit of a large system size. An accuracy within 100 K in melting temperature is usually achieved when the simulation contains more than 100 atoms. DFT examples of Tantalum, high-pressure Sodium, and ionic material NaCl are shown to demonstrate the accuracy and flexibility of the method in its practical applications. The method serves as a promising approach for large-scale automated material screening in which

Parallel SN transport calculations on a transputer network

International Nuclear Information System (INIS)

Kim, Yong Hee; Cho, Nam Zin

1994-01-01

A parallel computing algorithm for the neutron transport problems has been implemented on a transputer network and two reactor benchmark problems (a fixed-source problem and an eigenvalue problem) are solved. We have shown that the parallel calculations provided significant reduction in execution time over the sequential calculations

Direct calculation of wind turbine tip loss

DEFF Research Database (Denmark)

Wood, D.H.; Okulov, Valery; Bhattacharjee, D.

2016-01-01

. We develop three methods for the direct calculation of the tip loss. The first is the computationally expensive calculation of the velocities induced by the helicoidal wake which requires the evaluation of infinite sums of products of Bessel functions. The second uses the asymptotic evaluation......The usual method to account for a finite number of blades in blade element calculations of wind turbine performance is through a tip loss factor. Most analyses use the tip loss approximation due to Prandtl which is easily and cheaply calculated but is known to be inaccurate at low tip speed ratio...

MO-G-17A-04: Internal Dosimetric Calculations for Pediatric Nuclear Imaging Applications, Using Monte Carlo Simulations and High-Resolution Pediatric Computational Models

Energy Technology Data Exchange (ETDEWEB)

Papadimitroulas, P; Kagadis, GC [University of Patras, Rion, Ahaia (Greece); Loudos, G [Technical Educational Institute of Athens, Aigaleo, Attiki (Greece)

2014-06-15

Purpose: Our purpose is to evaluate the administered absorbed dose in pediatric, nuclear imaging studies. Monte Carlo simulations with the incorporation of pediatric computational models can serve as reference for the accurate determination of absorbed dose. The procedure of the calculated dosimetric factors is described, while a dataset of reference doses is created. Methods: Realistic simulations were executed using the GATE toolkit and a series of pediatric computational models, developed by the “IT'IS Foundation”. The series of the phantoms used in our work includes 6 models in the range of 5–14 years old (3 boys and 3 girls). Pre-processing techniques were applied to the images, to incorporate the phantoms in GATE simulations. The resolution of the phantoms was set to 2 mm3. The most important organ densities were simulated according to the GATE “Materials Database”. Several used radiopharmaceuticals in SPECT and PET applications are being tested, following the EANM pediatric dosage protocol. The biodistributions of the several isotopes used as activity maps in the simulations, were derived by the literature. Results: Initial results of absorbed dose per organ (mGy) are presented in a 5 years old girl from the whole body exposure to 99mTc - SestaMIBI, 30 minutes after administration. Heart, kidney, liver, ovary, pancreas and brain are the most critical organs, in which the S-factors are calculated. The statistical uncertainty in the simulation procedure was kept lower than 5%. The Sfactors for each target organ are calculated in Gy/(MBq*sec) with highest dose being absorbed in kidneys and pancreas (9.29*10{sup 10} and 0.15*10{sup 10} respectively). Conclusion: An approach for the accurate dosimetry on pediatric models is presented, creating a reference dosage dataset for several radionuclides in children computational models with the advantages of MC techniques. Our study is ongoing, extending our investigation to other reference models and

MO-G-17A-04: Internal Dosimetric Calculations for Pediatric Nuclear Imaging Applications, Using Monte Carlo Simulations and High-Resolution Pediatric Computational Models

International Nuclear Information System (INIS)

Papadimitroulas, P; Kagadis, GC; Loudos, G

2014-01-01

Purpose: Our purpose is to evaluate the administered absorbed dose in pediatric, nuclear imaging studies. Monte Carlo simulations with the incorporation of pediatric computational models can serve as reference for the accurate determination of absorbed dose. The procedure of the calculated dosimetric factors is described, while a dataset of reference doses is created. Methods: Realistic simulations were executed using the GATE toolkit and a series of pediatric computational models, developed by the “IT'IS Foundation”. The series of the phantoms used in our work includes 6 models in the range of 5–14 years old (3 boys and 3 girls). Pre-processing techniques were applied to the images, to incorporate the phantoms in GATE simulations. The resolution of the phantoms was set to 2 mm3. The most important organ densities were simulated according to the GATE “Materials Database”. Several used radiopharmaceuticals in SPECT and PET applications are being tested, following the EANM pediatric dosage protocol. The biodistributions of the several isotopes used as activity maps in the simulations, were derived by the literature. Results: Initial results of absorbed dose per organ (mGy) are presented in a 5 years old girl from the whole body exposure to 99mTc - SestaMIBI, 30 minutes after administration. Heart, kidney, liver, ovary, pancreas and brain are the most critical organs, in which the S-factors are calculated. The statistical uncertainty in the simulation procedure was kept lower than 5%. The Sfactors for each target organ are calculated in Gy/(MBq*sec) with highest dose being absorbed in kidneys and pancreas (9.29*10 10 and 0.15*10 10 respectively). Conclusion: An approach for the accurate dosimetry on pediatric models is presented, creating a reference dosage dataset for several radionuclides in children computational models with the advantages of MC techniques. Our study is ongoing, extending our investigation to other reference models and evaluating the

Recent skyshine calculations at Jefferson Lab

International Nuclear Information System (INIS)

Degtyarenko, P.

1997-01-01

New calculations of the skyshine dose distribution of neutrons and secondary photons have been performed at Jefferson Lab using the Monte Carlo method. The dose dependence on neutron energy, distance to the neutron source, polar angle of a source neutron, and azimuthal angle between the observation point and the momentum direction of a source neutron have been studied. The azimuthally asymmetric term in the skyshine dose distribution is shown to be important in the dose calculations around high-energy accelerator facilities. A parameterization formula and corresponding computer code have been developed which can be used for detailed calculations of the skyshine dose maps

Accuracy of WWR-M criticality calculations with code MCU-RFFI

International Nuclear Information System (INIS)

Petrov, Yu.V.; Erykalov, A.N.; Onegin, M.S.

1999-01-01

The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about Δρ≅±0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, α) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

Accuracy of WWR-M criticality calculations with code MCU-RFFI

Energy Technology Data Exchange (ETDEWEB)

Petrov, Yu V [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation); Erykalov, A N; Onegin, M S [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation)

1999-10-01

The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about {delta}{rho}{approx_equal}{+-}0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, {alpha}) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

Processing computed tomography images by using personal computer

International Nuclear Information System (INIS)

Seto, Kazuhiko; Fujishiro, Kazuo; Seki, Hirofumi; Yamamoto, Tetsuo.

1994-01-01

Processing of CT images was attempted by using a popular personal computer. The program for image-processing was made with C compiler. The original images, acquired with CT scanner (TCT-60A, Toshiba), were transferred to the computer by 8-inch flexible diskette. Many fundamental image-processing, such as displaying image to the monitor, calculating CT value and drawing the profile curve. The result showed that a popular personal computer had ability to process CT images. It seemed that 8-inch flexible diskette was still useful medium of transferring image data. (author)

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

International Nuclear Information System (INIS)

Rittmann, P.D.

1990-12-01

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

Modifications of the SEPHIS computer code for calculating the Purex solvent extraction system

International Nuclear Information System (INIS)

Watson, S.B.; Rainey, R.H.

1975-12-01

The SEPHIS computer program was developed to simulate the countercurrent solvent extraction. This report gives modifications in the program which result in improved fit to experimental data, a decrease in computer storage requirements, and a decrease in execution time. Methods for applying the computer program to practical solvent extraction problems are explained

Rotor calculations for neutron spectroscopy; Calculs des rotors de spectrometres a neutrons

Energy Technology Data Exchange (ETDEWEB)

Gobert, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

The determination of stress in a rotating disk plane of symmetry normal to the axis of rotation has been studied by a number of investigators. In a recent paper Reich gives an operating process for an analytical solution in an asymmetric rotating disk. In the report we give the calculation of finite difference stress solutions applicable to the two rotating disks. The equations are then programmed for the 360.75 computer by Fortran methods concerning the rotors of choppers. (author) [French] La determination des contraintes dans les disques symetriques, en rotation a ete etudiee par de nombreux auteurs. Dans un recent rapport, Reich donne une solution pour le calcul des disques asymetriques. Ce rapport concerne l'application du calcul des contraintes par differences finies aux deux types de rotors. Les equations ecrites en langage Fortran pour l'ordinateur 360.75 concerne les rotors de choppers. (auteur)

Independent calculation of monitor units for VMAT and SPORT

International Nuclear Information System (INIS)

Chen, Xin; Bush, Karl; Ding, Aiping; Xing, Lei

2015-01-01

Purpose: Dose and monitor units (MUs) represent two important facets of a radiation therapy treatment. In current practice, verification of a treatment plan is commonly done in dose domain, in which a phantom measurement or forward dose calculation is performed to examine the dosimetric accuracy and the MU settings of a given treatment plan. While it is desirable to verify directly the MU settings, a computational framework for obtaining the MU values from a known dose distribution has yet to be developed. This work presents a strategy to calculate independently the MUs from a given dose distribution of volumetric modulated arc therapy (VMAT) and station parameter optimized radiation therapy (SPORT). Methods: The dose at a point can be expressed as a sum of contributions from all the station points (or control points). This relationship forms the basis of the proposed MU verification technique. To proceed, the authors first obtain the matrix elements which characterize the dosimetric contribution of the involved station points by computing the doses at a series of voxels, typically on the prescription surface of the VMAT/SPORT treatment plan, with unit MU setting for all the station points. An in-house Monte Carlo (MC) software is used for the dose matrix calculation. The MUs of the station points are then derived by minimizing the least-squares difference between doses computed by the treatment planning system (TPS) and that of the MC for the selected set of voxels on the prescription surface. The technique is applied to 16 clinical cases with a variety of energies, disease sites, and TPS dose calculation algorithms. Results: For all plans except the lung cases with large tissue density inhomogeneity, the independently computed MUs agree with that of TPS to within 2.7% for all the station points. In the dose domain, no significant difference between the MC and Eclipse Anisotropic Analytical Algorithm (AAA) dose distribution is found in terms of isodose contours

Independent calculation of monitor units for VMAT and SPORT

Energy Technology Data Exchange (ETDEWEB)

Chen, Xin; Bush, Karl; Ding, Aiping; Xing, Lei, E-mail: lei@stanford.edu [Department of Radiation Oncology, Stanford University, Stanford, California 94305 (United States)

2015-02-15

Purpose: Dose and monitor units (MUs) represent two important facets of a radiation therapy treatment. In current practice, verification of a treatment plan is commonly done in dose domain, in which a phantom measurement or forward dose calculation is performed to examine the dosimetric accuracy and the MU settings of a given treatment plan. While it is desirable to verify directly the MU settings, a computational framework for obtaining the MU values from a known dose distribution has yet to be developed. This work presents a strategy to calculate independently the MUs from a given dose distribution of volumetric modulated arc therapy (VMAT) and station parameter optimized radiation therapy (SPORT). Methods: The dose at a point can be expressed as a sum of contributions from all the station points (or control points). This relationship forms the basis of the proposed MU verification technique. To proceed, the authors first obtain the matrix elements which characterize the dosimetric contribution of the involved station points by computing the doses at a series of voxels, typically on the prescription surface of the VMAT/SPORT treatment plan, with unit MU setting for all the station points. An in-house Monte Carlo (MC) software is used for the dose matrix calculation. The MUs of the station points are then derived by minimizing the least-squares difference between doses computed by the treatment planning system (TPS) and that of the MC for the selected set of voxels on the prescription surface. The technique is applied to 16 clinical cases with a variety of energies, disease sites, and TPS dose calculation algorithms. Results: For all plans except the lung cases with large tissue density inhomogeneity, the independently computed MUs agree with that of TPS to within 2.7% for all the station points. In the dose domain, no significant difference between the MC and Eclipse Anisotropic Analytical Algorithm (AAA) dose distribution is found in terms of isodose contours

SOILD: A computer model for calculating the effective dose equivalent from external exposure to distributed gamma sources in soil

International Nuclear Information System (INIS)

Chen, S.Y.; LePoire, D.; Yu, C.; Schafetz, S.; Mehta, P.

1991-01-01

The SOLID computer model was developed for calculating the effective dose equivalent from external exposure to distributed gamma sources in soil. It is designed to assess external doses under various exposure scenarios that may be encountered in environmental restoration programs. The models four major functional features address (1) dose versus source depth in soil, (2) shielding of clean cover soil, (3) area of contamination, and (4) nonuniform distribution of sources. The model is also capable of adjusting doses when there are variations in soil densities for both source and cover soils. The model is supported by a data base of approximately 500 radionuclides. 4 refs

Computational efficiency for the surface renewal method

Science.gov (United States)

Kelley, Jason; Higgins, Chad

2018-04-01

Measuring surface fluxes using the surface renewal (SR) method requires programmatic algorithms for tabulation, algebraic calculation, and data quality control. A number of different methods have been published describing automated calibration of SR parameters. Because the SR method utilizes high-frequency (10 Hz+) measurements, some steps in the flux calculation are computationally expensive, especially when automating SR to perform many iterations of these calculations. Several new algorithms were written that perform the required calculations more efficiently and rapidly, and that tested for sensitivity to length of flux averaging period, ability to measure over a large range of lag timescales, and overall computational efficiency. These algorithms utilize signal processing techniques and algebraic simplifications that demonstrate simple modifications that dramatically improve computational efficiency. The results here complement efforts by other authors to standardize a robust and accurate computational SR method. Increased speed of computation time grants flexibility to implementing the SR method, opening new avenues for SR to be used in research, for applied monitoring, and in novel field deployments.

Parallel computer calculation of quantum spin lattices; Calcul de chaines de spins quantiques sur ordinateur parallele

Energy Technology Data Exchange (ETDEWEB)

Lamarcq, J. [Service de Physique Theorique, CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France)

1998-07-10

Numerical simulation allows the theorists to convince themselves about the validity of the models they use. Particularly by simulating the spin lattices one can judge about the validity of a conjecture. Simulating a system defined by a large number of degrees of freedom requires highly sophisticated machines. This study deals with modelling the magnetic interactions between the ions of a crystal. Many exact results have been found for spin 1/2 systems but not for systems of other spins for which many simulation have been carried out. The interest for simulations has been renewed by the Haldane`s conjecture stipulating the existence of a energy gap between the ground state and the first excited states of a spin 1 lattice. The existence of this gap has been experimentally demonstrated. This report contains the following four chapters: 1. Spin systems; 2. Calculation of eigenvalues; 3. Programming; 4. Parallel calculation 14 refs., 6 figs.

Symbolic math for computation of radiation shielding

International Nuclear Information System (INIS)

Suman, Vitisha; Datta, D.; Sarkar, P.K.; Kushwaha, H.S.

2010-01-01

Radiation transport calculations for shielding studies in the field of accelerator technology often involve intensive numerical computations. Traditionally, radiation transport equation is solved using finite difference scheme or advanced finite element method with respect to specific initial and boundary conditions suitable for the geometry of the problem. All these computations need CPU intensive computer codes for accurate calculation of scalar and angular fluxes. Computation using symbols of the analytical expression representing the transport equation as objects is an enhanced numerical technique in which the computation is completely algorithm and data oriented. Algorithm on the basis of symbolic math architecture is developed using Symbolic math toolbox of MATLAB software. Present paper describes the symbolic math algorithm and its application as a case study in which shielding calculation of rectangular slab geometry is studied for a line source of specific activity. Study of application of symbolic math in this domain evolves a new paradigm compared to the existing computer code such as DORT. (author)

Two dimensional burn-up calculation of TRIGA core

International Nuclear Information System (INIS)

Persic, A.; Ravnik, M.; Slavic, S.

1996-01-01

TRIGLAV is a new computer program for burn-up calculation of mixed core of research reactors. The code is based on diffusion model in two dimensions and iterative procedure is applied for its solution. The material data used in the model are calculated with the transport program WIMS. In regard to fission density distribution and energy produced by the reactor the burn-up increment of fuel elements is determined. In this paper the calculation model of diffusion constants and burn-up calculation are described and some results of calculations for TRIGA MARK II reactor are presented. (author)

Structure problems in the analog computation; Problemes de structure dans le calcul analogique

Energy Technology Data Exchange (ETDEWEB)

Braffort, P.L. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The recent mathematical development showed the importance of elementary structures (algebraic, topological, etc.) in abeyance under the great domains of classical analysis. Such structures in analog computation are put in evidence and possible development of applied mathematics are discussed. It also studied the topological structures of the standard representation of analog schemes such as additional triangles, integrators, phase inverters and functions generators. The analog method gives only the function of the variable: time, as results of its computations. But the course of computation, for systems including reactive circuits, introduces order structures which are called 'chronological'. Finally, it showed that the approximation methods of ordinary numerical and digital computation present the same structure as these analog computation. The structure analysis permits fruitful comparisons between the several domains of applied mathematics and suggests new important domains of application for analog method. (M.P.)

Structure problems in the analog computation; Problemes de structure dans le calcul analogique

Energy Technology Data Exchange (ETDEWEB)

Braffort, P L [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The recent mathematical development showed the importance of elementary structures (algebraic, topological, etc.) in abeyance under the great domains of classical analysis. Such structures in analog computation are put in evidence and possible development of applied mathematics are discussed. It also studied the topological structures of the standard representation of analog schemes such as additional triangles, integrators, phase inverters and functions generators. The analog method gives only the function of the variable: time, as results of its computations. But the course of computation, for systems including reactive circuits, introduces order structures which are called 'chronological'. Finally, it showed that the approximation methods of ordinary numerical and digital computation present the same structure as these analog computation. The structure analysis permits fruitful comparisons between the several domains of applied mathematics and suggests new important domains of application for analog method. (M.P.)

The role of desk calculators in nuclear data evaluation

International Nuclear Information System (INIS)

Motta, M.

1980-01-01

The performances of the modern Desk Calculators are more and more increasing. Consequently, the best feature for the definition of a Desk Calculator seems to be prices and volume occupation. The interactive operating mode and the low installation and maintaining costs make the use of these computing machines very likely and economical. A list of tasks which are profitably performed in nuclear data preparation are presented here. A lot of practical applications are given through the formulas and the list of codes written in BASIC language, which is commonly adopted for these small computers. (author)

Automated procedure for calculating time-dependent sensitivities in ORIGEN2

International Nuclear Information System (INIS)

Worley, B.A.; Wright, R.Q.

1985-10-01

ORIGEN2 is a widely used point-depletion and radioactive-decay computer code for use in simulating nuclear fuel cycles and/or spent fuel characteristics. This paper presents the application of the GRESS procedure to the ORIGEN2 code for performing a sensitivity analysis of a high-level waste disposal problem. The GRESS procedure uses computer calculus and the GRESS precompiler to automate the generation and calculation of gradients in a computer code. The GRESS version of ORIGEN2 is used to calculate the nuclide-dependent sensitivities of the decay heat and radioactivity of 1008 nuclides comprising reprocessed high-level waste to changes in data and input parameters. The sensitivities are calculated in a single execution of the revised code as compared to the conventional method of rerunning the code numerous times. The availability of sensitivity data as an option in ORIGEN2 reveals relationships not easily recognized even with reruns

Linear augmented plane wave method for self-consistent calculations

International Nuclear Information System (INIS)

Takeda, T.; Kuebler, J.

1979-01-01

O.K. Andersen has recently introduced a linear augmented plane wave method (LAPW) for the calculation of electronic structure that was shown to be computationally fast. A more general formulation of an LAPW method is presented here. It makes use of a freely disposable number of eigenfunctions of the radial Schroedinger equation. These eigenfunctions can be selected in a self-consistent way. The present formulation also results in a computationally fast method. It is shown that Andersen's LAPW is obtained in a special limit from the present formulation. Self-consistent test calculations for copper show the present method to be remarkably accurate. As an application, scalar-relativistic self-consistent calculations are presented for the band structure of FCC lanthanum. (author)

Interactions of model biomolecules. Benchmark CC calculations within MOLCAS

Energy Technology Data Exchange (ETDEWEB)

Urban, Miroslav [Slovak University of Technology in Bratislava, Faculty of Materials Science and Technology in Trnava, Institute of Materials Science, Bottova 25, SK-917 24 Trnava, Slovakia and Department of Physical and Theoretical Chemistry, Faculty of Natural Scie (Slovakia); Pitoňák, Michal; Neogrády, Pavel; Dedíková, Pavlína [Department of Physical and Theoretical Chemistry, Faculty of Natural Sciences, Comenius University, Mlynská dolina, SK-842 15 Bratislava (Slovakia); Hobza, Pavel [Institute of Organic Chemistry and Biochemistry and Center for Complex Molecular Systems and biomolecules, Academy of Sciences of the Czech Republic, Prague (Czech Republic)

2015-01-22

We present results using the OVOS approach (Optimized Virtual Orbitals Space) aimed at enhancing the effectiveness of the Coupled Cluster calculations. This approach allows to reduce the total computer time required for large-scale CCSD(T) calculations about ten times when the original full virtual space is reduced to about 50% of its original size without affecting the accuracy. The method is implemented in the MOLCAS computer program. When combined with the Cholesky decomposition of the two-electron integrals and suitable parallelization it allows calculations which were formerly prohibitively too demanding. We focused ourselves to accurate calculations of the hydrogen bonded and the stacking interactions of the model biomolecules. Interaction energies of the formaldehyde, formamide, benzene, and uracil dimers and the three-body contributions in the cytosine – guanine tetramer are presented. Other applications, as the electron affinity of the uracil affected by solvation are also shortly mentioned.

A computer-based matrix for rapid calculation of pulmonary hemodynamic parameters in congenital heart disease

International Nuclear Information System (INIS)

Lopes, Antonio Augusto; Miranda, Rogerio dos Anjos; Goncalves, Rilvani Cavalcante; Thomaz, Ana Maria

2009-01-01

In patients with congenital heart disease undergoing cardiac catheterization for hemodynamic purposes, parameter estimation by the indirect Fick method using a single predicted value of oxygen consumption has been a matter of criticism. We developed a computer-based routine for rapid estimation of replicate hemodynamic parameters using multiple predicted values of oxygen consumption. Using Microsoft Excel facilities, we constructed a matrix containing 5 models (equations) for prediction of oxygen consumption, and all additional formulas needed to obtain replicate estimates of hemodynamic parameters. By entering data from 65 patients with ventricular septal defects, aged 1 month to 8 years, it was possible to obtain multiple predictions for oxygen consumption, with clear between-age groups ( P <.001) and between-methods ( P <.001) differences. Using these predictions in the individual patient, it was possible to obtain the upper and lower limits of a likely range for any given parameter, which made estimation more realistic. The organized matrix allows for rapid obtainment of replicate parameter estimates, without error due to exhaustive calculations. (author)

Comparison between different computational schemes for variational calculations in nuclear structure

International Nuclear Information System (INIS)

Puddu, G.

2009-01-01

We compare several iteration methods for angular-momentum- and parity-projected Hartree-Fock calculations. We used the Anderson update, the modified Broyden method, newly introduced in nuclear-structure calculations, and variants of the Broyden-Fletcher-Goldhaber-Shanno methods (BFGS). We performed ground-state calculations for 18 C and 6 Li using the two-body Hamiltonian obtained from the CDBonn-2000 potential via the Lee-Suzuki renormalization method. We found that BFGS methods are superior to both the Anderson update and to the modified Broyden method. In the case of 6 Li we found that the Anderson update and modified Broyden method do not converge to the angular-momentum- and parity-projected Hartree-Fock minimum. The reason is traced back to the lack of a mechanism that guarantees a decrease of the energy from one iteration to the next and to the fact that these methods guarantee a stationary solution rather than a minimum of the energy. (orig.)

Calculation of wind turbine aeroelastic behaviour. The Garrad Hassan approach

Energy Technology Data Exchange (ETDEWEB)

Quarton, D C [Garrad Hassan and Partners Ltd., Bristol (United Kingdom)

1996-09-01

The Garrad Hassan approach to the prediction of wind turbine loading and response has been developed over the last decade. The goal of this development has been to produce calculation methods that contain realistic representation of the wind, include sensible aerodynamic and dynamic models of the turbine and can be used to predict fatigue and extreme loads for design purposes. The Garrad Hassan calculation method is based on a suite of four key computer programs: WIND3D for generation of the turbulent wind field; EIGEN for modal analysis of the rotor and support structure; BLADED for time domain calculation of the structural loads; and SIGNAL for post-processing of the BLADED predictions. The interaction of these computer programs is illustrated. A description of the main elements of the calculation method will be presented. (au)

Dynamic Thermal Loads and Cooling Requirements Calculations for V ACs System in Nuclear Fuel Processing Facilities Using Computer Aided Energy Conservation Models

International Nuclear Information System (INIS)

EL Fawal, M.M.; Gadalla, A.A.; Taher, B.M.

2010-01-01

In terms of nuclear safety, the most important function of ventilation air conditioning (VAC) systems is to maintain safe ambient conditions for components and structures important to safety inside the nuclear facility and to maintain appropriate working conditions for the plant's operating and maintenance staff. As a part of a study aimed to evaluate the performance of VAC system of the nuclear fuel cycle facility (NFCF) a computer model was developed and verified to evaluate the thermal loads and cooling requirements for different zones of fuel processing facility. The program is based on transfer function method (TFM) and it is used to calculate the dynamic heat gain by various multilayer walls constructions and windows hour by hour at any orientation of the building. The developed model was verified by comparing the obtained calculated results of the solar heat gain by a given building with the corresponding calculated values using finite difference method (FDM) and total equivalent temperature different method (TETD). As an example the developed program is used to calculate the cooling loads of the different zones of a typical nuclear fuel facility the results showed that the cooling capacities of the different cooling units of each zone of the facility meet the design requirements according to safety regulations in nuclear facilities.

Three dimensional diffusion calculations of nuclear reactors

International Nuclear Information System (INIS)

Caspo, N.

1981-07-01

This work deals with the three dimensional calculation of nuclear reactors using the code TRITON. The purposes of the work were to perform three-dimensional computations of the core of the Soreq nuclear reactor and of the power reactor ZION and to validate the TRITON code. Possible applications of the TRITON code in Soreq reactor calculations and in power reactor research are suggested. (H.K.)

Dosimetric comparison of lung stereotactic body radiotherapy treatment plans using averaged computed tomography and end-exhalation computed tomography images: Evaluation of the effect of different dose-calculation algorithms and prescription methods