SLAROM: a code for cell homogenization calculation of fast reactor
International Nuclear Information System (INIS)
Nakagawa, Masayuki; Tsuchihashi, Keiichiro
1984-12-01
A revised version of the SLAROM code has been developed. The main function of SLAROM is to perform the cell homogenization calculation of a fast power reactor and a fast critical assembly. The code uses the JFS2 or JFS3 type cross section set as a multi-group cross section library. The region dependent effective cross sections are calculated by taking account of the heterogeneity effect of resonance shielding for heavy nuclides. The integral transport equations are solved by using the collision probability method. SLAROM installs collision probability calculation routines for various geometries encountered in a fast reactor analysis. The effective multiplication factor (ksub(eff)) calculation or buckling search mode is available. The cell homogenized cross sections are obtained by weighting with the fine structure flux and volumes. The calculation of anisotropic diffusion coefficient is based on the Benoist's definition with use of the directional collision probability. The averaged macroscopic and microscopic cross sections are saved on the Partitioned Data Set file with a unified format. In addition to the cell calculation, another module is equipped to solve one dimensional diffusion equations in normal and adjoint modes. The fluxes obtained by this module can be used to collapse the fine group cross sections into the broad group structure. The perturbation calculation is also available. This report describes the calculational method adopted in the SLAROM code, input data and job control statements instructions, structure of the code, file requirement and sample input and output data. Since the input data are punched in a free format, users will be easy to prepare them. The description of auxiliary programs is given in Appendix for a help of the data handling on the PDS file. (author)
LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory
International Nuclear Information System (INIS)
Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.
1976-04-01
The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented
Kinetic parameters evaluation of PWRs using static cell and core calculation codes
International Nuclear Information System (INIS)
Jahanbin, Ali; Malmir, Hessam
2012-01-01
Highlights: ► In this study, we have calculated effective delayed neutron fraction and prompt neutron lifetime in PWRs. ► New software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. ► This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. ► The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. - Abstract: In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs, using static cell and core calculation codes, is reported. A new software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. Using the WIMS cell calculation code, multigroup microscopic cross-sections and number densities of different materials can be generated in a binary file. By the use of BORGES code, these binary-form cross-sections and number densities are converted to a format readable by the CITATION core calculation code, by which the kinetic parameters can be finally obtained. This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. Benchmarking of the results against the final safety analysis report (FSAR) of the aforementioned reactors shows very good agreements with these published documents.
Application Of WIMS Code To Calculation Kartini Reactor Parameters By Pin-Cell And Cluster Method
International Nuclear Information System (INIS)
Sumarsono, Bambang; Tjiptono, T.W.
1996-01-01
Analysis UZrH fuel element parameters calculation in Kartini Reactor by WIMS Code has been done. The analysis is done by pin cell and cluster method. The pin cell method is done as a function percent burn-up and by 8 group 3 region analysis and cluster method by 8 group 12 region analysis. From analysis and calculation resulted K ∼ = 1.3687 by pin cell method and K ∼ = 1.3162 by cluster method and so deviation is 3.83%. By pin cell analysis as a function percent burn-up at the percent burn-up greater than 59.50%, the multiplication factor is less than one (k ∼ < 1) it is mean that the fuel element reactivity is negative
Vector calculation of particle code
International Nuclear Information System (INIS)
Nishiguchi, A.; Yabe, T.; Orii, S.
1985-01-01
The development of vector computer requires the modification of the algorithm into a suitable form for vector calculation. Among many algorithms, the particle code is a typical example which has suffered a damage in the calculation on supercomputer owing to its possibility of recurrent data access in collecting cell-wise quantities from particle's quartities. In this article, we report a new method to liberate the particle code from recurrent calculations. It should be noticed, however, that the method may depend on the architecture of supercomputer, and works well on FACOM VP-100 and VP-200: the indirect data accessing must be vectorized and its speed should be fast. (Mori, K.)
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
Energy Technology Data Exchange (ETDEWEB)
Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de, E-mail: zelmolima@yahoo.com.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2017-07-01
This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)
Directory of Open Access Journals (Sweden)
Diego Ferraro
2011-01-01
Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.
Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code
Directory of Open Access Journals (Sweden)
Ahmed Amin El Said Abd El Hameed
2015-08-01
Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code. We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon
The Flutter Shutter Code Calculator
Directory of Open Access Journals (Sweden)
Yohann Tendero
2015-08-01
Full Text Available The goal of the flutter shutter is to make uniform motion blur invertible, by a"fluttering" shutter that opens and closes on a sequence of well chosen sub-intervals of the exposure time interval. In other words, the photon flux is modulated according to a well chosen sequence calledflutter shutter code. This article provides a numerical method that computes optimal flutter shutter codes in terms of mean square error (MSE. We assume that the observed objects follow a known (or learned random velocity distribution. In this paper, Gaussian and uniform velocity distributions are considered. Snapshots are also optimized taking the velocity distribution into account. For each velocity distribution, the gain of the optimal flutter shutter code with respectto the optimal snapshot in terms of MSE is computed. This symmetric optimization of theflutter shutter and of the snapshot allows to compare on an equal footing both solutions, i.e. camera designs. Optimal flutter shutter codes are demonstrated to improve substantially the MSE compared to classic (patented or not codes. A numerical method that permits to perform a reverse engineering of any existing (patented or not flutter shutter codes is also describedand an implementation is given. In this case we give the underlying velocity distribution fromwhich a given optimal flutter shutter code comes from. The combination of these two numerical methods furnishes a comprehensive study of the optimization of a flutter shutter that includes a forward and a backward numerical solution.
Two-dimensional sensitivity calculation code: SENSETWO
International Nuclear Information System (INIS)
Yamauchi, Michinori; Nakayama, Mitsuo; Minami, Kazuyoshi; Seki, Yasushi; Iida, Hiromasa.
1979-05-01
A SENSETWO code for the calculation of cross section sensitivities with a two-dimensional model has been developed, on the basis of first order perturbation theory. It uses forward neutron and/or gamma-ray fluxes and adjoint fluxes obtained by two-dimensional discrete ordinates code TWOTRAN-II. The data and informations of cross sections, geometry, nuclide density, response functions, etc. are transmitted to SENSETWO by the dump magnetic tape made in TWOTRAN calculations. The required input for SENSETWO calculations is thus very simple. The SENSETWO yields as printed output the cross section sensitivities for each coarse mesh zone and for each energy group, as well as the plotted output of sensitivity profiles specified by the input. A special feature of the code is that it also calculates the reaction rate with the response function used as the adjoint source in TWOTRAN adjoint calculation and the calculated forward flux from the TWOTRAN forward calculation. (author)
Theoretical calculation possibilities of the computer code HAMMER
International Nuclear Information System (INIS)
Onusic Junior, J.
1978-06-01
With the aim to know the theoretical calculation possibilities of the computer code HAMMER, developed at Savanah River Laboratory, a analysis of the crytical cells assembly of the kind utilized in PWR reactors is made. (L.F.S.) [pt
Benchmark calculation of subchannel analysis codes
International Nuclear Information System (INIS)
1996-02-01
In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)
SPINLIE - new computer code for polarization calculation
International Nuclear Information System (INIS)
Eidelman, Yu.; Yakimenko, V.
1993-01-01
The application of the analytical method of spin calculation is described. The calculation of both orbital and spin motion is based on Lie operators technique. The computer code SPINLIE realizing this method is discussed. The input language of SPINLIE is compatible with that of MAD. In SPINLIE elements are described as open-quotes thick lensesclose quotes for spin motion as well as for orbital calculations. The explicit expressions of Lie operators were found for orbital and spin motion for elements of different type (bending magnets, quadrupoles, sextupoles, RF-cavities, solenoids, kickers). The rules of addition of spin transformations were obtained for the beam passing the collider structure. Good agreement was found for SPINLIE results for the linear spin resonances in comparison with the other codes (SITF, SMILE). The first results are presented for the calculation of nonlinear spin resonances
TEA: A CODE CALCULATING THERMOCHEMICAL EQUILIBRIUM ABUNDANCES
Energy Technology Data Exchange (ETDEWEB)
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver, E-mail: jasmina@physics.ucf.edu [Planetary Sciences Group, Department of Physics, University of Central Florida, Orlando, FL 32816-2385 (United States)
2016-07-01
We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. and Eriksson. It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature–pressure pairs. We tested the code against the method of Burrows and Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows and Sharp. Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is available under a reproducible-research, open-source license via https://github.com/dzesmin/TEA.
TEA: A CODE CALCULATING THERMOCHEMICAL EQUILIBRIUM ABUNDANCES
International Nuclear Information System (INIS)
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver
2016-01-01
We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. and Eriksson. It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature–pressure pairs. We tested the code against the method of Burrows and Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows and Sharp. Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is available under a reproducible-research, open-source license via https://github.com/dzesmin/TEA.
Code ATOM for calculation of atomic characteristics
International Nuclear Information System (INIS)
Vainshtein, L.A.
1990-01-01
In applying atomic physics to problems of plasma diagnostics, it is necessary to determine some atomic characteristics, including energies and transition probabilities, for very many atoms and ions. Development of general codes for calculation of many types of atomic characteristics has been based on general but comparatively simple approximate methods. The program ATOM represents an attempt at effective use of such a general code. This report gives a brief description of the methods used, and the possibilities of and limitations to the code are discussed. Characteristics of the following processes can be calculated by ATOM: radiative transitions between discrete levels, radiative ionization and recombination, collisional excitation and ionization by electron impact, collisional excitation and ionization by point heavy particle (Born approximation only), dielectronic recombination, and autoionization. ATOM explores Born (for z=1) or Coulomb-Born (for z>1) approximations. In both cases exchange and normalization can be included. (N.K.)
Burnup calculation code system COMRAD96
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu
1997-06-01
COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, `Cross Section Treatment`, `Generation and Depletion Calculation`, and `Post Process`. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the {gamma} Spectrum on a terminal. This report is the general description and user`s manual of COMRAD96. (author)
Burnup calculation code system COMRAD96
International Nuclear Information System (INIS)
Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.
1997-06-01
COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)
Data calculation program for RELAP 5 code
Energy Technology Data Exchange (ETDEWEB)
Silvestre, Larissa J.B.; Sabundjian, Gaiane, E-mail: larissajbs@usp.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2015-07-01
As the criteria and requirements for a nuclear power plant are extremely rigid, computer programs for simulation and safety analysis are required for certifying and licensing a plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors. A major difficulty in the simulation using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data leads to a very large number of mathematical operations for calculating the geometry of the components. Therefore, a mathematical friendly preprocessor was developed in order to perform these calculations and prepare RELAP5 input data. The Visual Basic for Application (VBA) combined with Microsoft EXCEL demonstrated to be an efficient tool to perform a number of tasks in the development of the program. Due to the absence of necessary information about some RELAP5 components, this work aims to make improvements to the Mathematic Preprocessor for RELAP5 code (PREREL5). For the new version of the preprocessor, new screens of some components that were not programmed in the original version were designed; moreover, screens of pre-existing components were redesigned to improve the program. In addition, an English version was provided for the new version of the PREREL5. The new design of PREREL5 contributes for saving time and minimizing mistakes made by users of the RELAP5 code. The final version of this preprocessor will be applied to Angra 2. (author)
Integrated burnup calculation code system SWAT
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya; Hirakawa, Naohiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwasaki, Tomohiko
1997-11-01
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user`s manual of SWAT. (author)
Calculation code MIXSET for Purex process
International Nuclear Information System (INIS)
Gonda, Kozo; Fukuda, Shoji.
1977-09-01
MIXSET is a FORTRAN IV calculation code for Purex process that simulate the dynamic behavior of solvent extraction processes in mixer-settlers. Two options permit terminating dynamic phase by time or by achieving steady state. These options also permit continuing calculation successively using new inputs from a arbitrary phase. A third option permits artificial rapid close to steady state and a fourth option permits searching optimum input to satisfy both of specification and recovery rate of product. MIXSET handles maximum chemical system of eight components with or without mutual dependence of the distribution of the components. The chemical system in MIXSET includes chemical reactions and/or decaying reaction. Distribution data can be supplied by third-power polynominal equations or tables, and kinetic data by tables or given constants. The fluctuation of the interfacial level height in settler is converted into the flow rate changes of organic and aqueous stream to follow dynamic behavior of extraction process in detail. MIXSET can be applied to flowsheet study, start up and/or shut down procedure study and real time process management in countercurrent solvent extraction processes. (auth.)
Code system BCG for gamma-ray skyshine calculation
International Nuclear Information System (INIS)
Ryufuku, Hiroshi; Numakunai, Takao; Miyasaka, Shun-ichi; Minami, Kazuyoshi.
1979-03-01
A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)
Development of the code package KASKAD for calculations of WWERs
International Nuclear Information System (INIS)
Bolobov, P.A.; Lazarenko, A.P.; Tomilov, M.Ju.
2008-01-01
The new version of software package for neutron calculation of WWER cores KASKAD 2007 consists of some calculating and service modules, which are integrated in the common framework. The package is based on the old version, which was expanded with some new functions and the new calculating modules, such as: -the BIPR-2007 code is the new one which performs calculation of power distribution in three-dimensional geometry for 2-group neutron diffusion calculation. This code is based on the BIPR-8KN model, provides all possibilities of BIPR-7A code and uses the same input data; -the PERMAK-2007 code is pin-by-pin few-group multilayer and 3-D code for neutron diffusion calculation; -graphical user interface for input data preparation of the TVS-M code. The report also includes some calculation results obtained with modified version of the KASKAD 2007 package. (Authors)
Lattice cell burnup calculation
International Nuclear Information System (INIS)
Pop-Jordanov, J.
1977-01-01
Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics
International Nuclear Information System (INIS)
Shindo, R.; Yamashita, K.; Murata, I.
1991-01-01
The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs
Development of Reactivity Calculation Code for HANARO
Energy Technology Data Exchange (ETDEWEB)
Park, B. G.; Kim, M. S. [KAERI, Daejeon (Korea, Republic of)
2016-05-15
Reactivity of the reactor core is measured by a multi-channel wide range reactivity computer (or called reactivity meter), which uses current signals from the compensated ion chamber (CIC) mounted on the courtside wall of the reflector tank in the pool [1]. Because there were a few difficulties in operating the reactivity meter in the MS-DOS environment, some researches have been carried out to improve and upgrade it on the Windows environment [2]. Nevertheless, it is still hard for reactor operators to immediately check the time-dependent reactivity in case of power excursion because of some limitations such as aging of devices and compatibility issues. In this study, a simple off-line tool which can estimate the timedependent reactivity by using the fission chamber signals has been developed, and utilized to the case of loose parts of dummy rod in the 86-2th cycles of HANARO. In order to check the reactivity quickly for reactor operators, the inverse kinetic quations have been incorporated, and several useful functions have been implemented to the code. in the case of 86-2th cycles of HANARO, the developed code showed good performance to estimate time dependent reactivity. In the future, the on line analysis modules will be implanted to the code with upgrade of the measrement equipment such as current meters and data acquisition devices. Additionally, reactivity will be estimated by using the reactivity meter in the MS-DOS enviroment, and the new Windows version, for the verification of the developed code.
Thermal hydraulic calculation of STORM facility using GOTHIC code
International Nuclear Information System (INIS)
Pevec, D.; Grgic, D.; Prah, M.
1995-01-01
Benchmark calculation CTI defined in frame of STORM experimental programme is used to prove that the GOTHIC code is capable to predict behaviour of experimental facility with reasonable accuracy. GOTHIC code is developed mainly for containment calculation. In this situation it is successfully used for calculation of one dimensional flow of steam and noncondensable mixture. Steady state distributions of pressure, temperature and the velocity of gas along facility are consistent with results obtained by other benchmark participants. (author)
Development and validation of a nodal code for core calculation
International Nuclear Information System (INIS)
Nowakowski, Pedro Mariano
2004-01-01
The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency
SRAC2006: A comprehensive neutronics calculation code system
International Nuclear Information System (INIS)
Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro
2007-02-01
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, S N transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)
MCOR - Monte Carlo depletion code for reference LWR calculations
Energy Technology Data Exchange (ETDEWEB)
Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)
2011-04-15
Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally
MCOR - Monte Carlo depletion code for reference LWR calculations
International Nuclear Information System (INIS)
Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan
2011-01-01
Research highlights: → Introduction of a reference Monte Carlo based depletion code with extended capabilities. → Verification and validation results for MCOR. → Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations
Use of the Apollo-II multigroup transport code for criticality calculations
International Nuclear Information System (INIS)
Coste, M.; Mathonniere, G.; Sanchez, R.; Stankovski, Z.; Van der Gucht, C.; Zmijarevic, I.
1992-01-01
APPOLO-II is a new-generation multigroup transport code for assembly calculation. The code has been designed to be used as a tool for reactor design as well as for the analysis and interpretation of small nuclear facilities. As the first step in a criticality calculation, the collision probability module of the APPOLO-II code can be used to generate cell or assembly homogenized reaction-rate preserving cross sections that account for self-shielding effects as well as for the fine-energy within cell flux spectral variations. These cross section data can then be used either directly within the APPOLO-II code in a direct discrete ordinate multigroup transport calculation of a small nuclear facility or, more generally, be formatted by a post-processing module to be used by the multigroup diffusion code CRONOS-II or by the multigroup Monte Carlo code TRIMARAN
Research on Primary Shielding Calculation Source Generation Codes
Zheng, Zheng; Mei, Qiliang; Li, Hui; Shangguan, Danhua; Zhang, Guangchun
2017-09-01
Primary Shielding Calculation (PSC) plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF) for the source particle sample code of the J Monte Carlo Transport (JMCT) code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three dimensional (3D) power distributions in xyz geometry to source distributions in r θ z geometry for the J Discrete Ordinate Transport (JSNT) code. Validation on PSC model of Qinshan No.1 nuclear power plant (NPP), CAP1400 and CAP1700 reactors are performed. Numerical results show that the theoretical model and the codes are both correct.
Multi-group diffusion perturbation calculation code. PERKY (2002)
Energy Technology Data Exchange (ETDEWEB)
Iijima, Susumu; Okajima, Shigeaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-12-01
Perturbation calculation code based on the diffusion theory ''PERKY'' is designed for nuclear characteristic analyses of fast reactor. The code calculates reactivity worth on the multi-group diffusion perturbation theory in two or three dimensional core model and kinetics parameters such as effective delayed neutron fraction, prompt neutron lifetime and absolute reactivity scale factor ({rho}{sub 0} {delta}k/k) for FCA experiments. (author)
Direct calculation of current drive efficiency in FISIC code
International Nuclear Information System (INIS)
Wright, J.C.; Phillips, C.K.; Bonoli, P.T.
1996-01-01
Two-dimensional RF modeling codes use a parameterization (1) of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform quasi-linear diffusion coefficient and requires a priori knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by the full wave code, FISIC (2). Current profiles are calculated using the adjoint formulation (3). Comparisons between the two formulations are presented. copyright 1996 American Institute of Physics
Direct calculation of current drive efficiency in FISIC code
Wright, J. C.; Phillips, C. K.; Bonoli, P. T.
1996-02-01
Two-dimensional RF modeling codes use a parameterization (1) of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform quasi-linear diffusion coefficient and requires a priori knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by the full wave code, FISIC (2). Current profiles are calculated using the adjoint formulation (3). Comparisons between the two formulations are presented.
Direct calculation of current drive efficiency in FISIC code
Energy Technology Data Exchange (ETDEWEB)
Wright, J.C.; Phillips, C.K. [Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543-0451 (United States); Bonoli, P.T. [Plasma Fusion Center, MIT Cambridge, Massachusetts 02139 (United States)
1996-02-01
Two-dimensional RF modeling codes use a parameterization (1) of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform quasi-linear diffusion coefficient and requires {ital a} {ital priori} knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by the full wave code, FISIC (2). Current profiles are calculated using the adjoint formulation (3). Comparisons between the two formulations are presented. {copyright} {ital 1996 American Institute of Physics.}
Description of the CAREM Reactor Neutronic Calculation Codes
International Nuclear Information System (INIS)
Villarino, Eduardo; Hergenreder, Daniel
2000-01-01
In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included
Hamor-2: a computer code for LWR inventory calculation
International Nuclear Information System (INIS)
Guimaraes, L.N.F.; Marzo, M.A.S.
1985-01-01
A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt
Energy Technology Data Exchange (ETDEWEB)
NONE
2003-07-01
These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)
A computer code for calculating a γ-external dose from a randomly distributed radioactive cloud
International Nuclear Information System (INIS)
Kai, Michiaki
1984-02-01
A computer code ( CIDE ) has been developed to calculate a γ-external dose from a randomly distributed radioactive cloud. Atmospheric dispersion of radioactive materials accidentally released from a nuclear reactor needs to be estimated considering time-dependent meteorological data and terrain heights. Particle-in-Cell model is useful for that purpose, but it is not easy to calculate the dose from the randomly distributed concentration by numerical integration. In this study the mean concentration in a cell evaluated by PIC model was assumed to be uniformly distributed over that cell, which was integrated as a constant concentration by a point kernel method. The dose was obtained by summing the attributable cell doses. When the concentration of plume had a Gaussian distribution, the results of CIDE code well agreed with those of GAMPLE, which was the code for calculating the dose from the Gaussian distribution. The choice of cell sizes affecting the accuracy of the calculated results was discussed. (author)
Exposure calculation code module for reactor core analysis: BURNER
International Nuclear Information System (INIS)
Vondy, D.R.; Cunningham, G.W.
1979-02-01
The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules
Exposure calculation code module for reactor core analysis: BURNER
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Cunningham, G.W.
1979-02-01
The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.
Application of fuel management calculation codes for CANDU reactor
International Nuclear Information System (INIS)
Ju Haitao; Wu Hongchun
2003-01-01
Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III
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
A neutron-reflecting conical tube calculation code: FOCUS
International Nuclear Information System (INIS)
Shimooke, Takanori; Levine, M.M.
1976-02-01
A focalizer, i.e. the neutron-reflecting conical tube is a device to collimate and intensify a beam of neutrons by means of the total reflection. This report describes the code FOCUS for calculating the neutron flux intensity and spectrum emerging from such a device. Complete ''input and output'' procedures are written for the code. The equations and methods used in the code are also explained. In the introduction, a brief description is given on the total reflection of neutrons and on the several devices of the neutron reflecting and conducting tubes. (auth.)
User effects on the transient system code calculations. Final report
International Nuclear Information System (INIS)
Aksan, S.N.; D'Auria, F.
1995-01-01
Large thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are accomplished by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies has the influence of the code user on the calculated results been directly addressed. In this paper, the results of the investigations on the user effects for the thermal-hydraulic transient system codes is presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects are discussed in detail and general recommendations and conclusions are presented to control and limit them
The MCEF code for nuclear evaporation and fission calculations
International Nuclear Information System (INIS)
Deppman, A.; Pina, S.R. de; Likhachev, V.P.; Mesa, J.; Arruda-Neto, J.D.T.; Rodriguez, O.; Goncalves, M.
2001-11-01
We present an object oriented algorithm, written in the Java programming language, which performs a Monte Carlo calculation of the evaporation-fission process taking place inside an excited nucleus. We show that this nuclear physics problem is very suited for the object oriented programming by constructing two simple objects: one that handles all nuclear properties and another that takes care of the nuclear reaction. The MCEF code was used to calculate important results for nuclear reactions, and here we show examples of possible uses for this code. (author)
Development of throughflow calculation code for axial flow compressors
International Nuclear Information System (INIS)
Kim, Ji Hwan; Kim, Hyeun Min; No, Hee Cheon
2005-01-01
The power conversion systems of the current HTGRs are based on closed Brayton cycle and major concern is thermodynamic performance of the axial flow helium gas turbines. Particularly, the helium compressor has some unique design challenges compared to the air-breathing compressor such as high hub-to-tip ratios throughout the machine and a large number of stages due to the physical property of the helium and thermodynamic cycle. Therefore, it is necessary to develop a design and analysis code for helium compressor that can estimate the design point and off-design performance accurately. KAIST nuclear system laboratory has developed a compressor design and analysis code by means of throughflow calculation and several loss models. This paper presents the outline of the development of a throughflow calculation code and its verification results
WIPP Benchmark calculations with the large strain SPECTROM codes
Energy Technology Data Exchange (ETDEWEB)
Callahan, G.D.; DeVries, K.L. [RE/SPEC, Inc., Rapid City, SD (United States)
1995-08-01
This report provides calculational results from the updated Lagrangian structural finite-element programs SPECTROM-32 and SPECTROM-333 for the purpose of qualifying these codes to perform analyses of structural situations in the Waste Isolation Pilot Plant (WIPP). Results are presented for the Second WIPP Benchmark (Benchmark II) Problems and for a simplified heated room problem used in a parallel design calculation study. The Benchmark II problems consist of an isothermal room problem and a heated room problem. The stratigraphy involves 27 distinct geologic layers including ten clay seams of which four are modeled as frictionless sliding interfaces. The analyses of the Benchmark II problems consider a 10-year simulation period. The evaluation of nine structural codes used in the Benchmark II problems shows that inclusion of finite-strain effects is not as significant as observed for the simplified heated room problem, and a variety of finite-strain and small-strain formulations produced similar results. The simplified heated room problem provides stratigraphic complexity equivalent to the Benchmark II problems but neglects sliding along the clay seams. The simplified heated problem does, however, provide a calculational check case where the small strain-formulation produced room closures about 20 percent greater than those obtained using finite-strain formulations. A discussion is given of each of the solved problems, and the computational results are compared with available published results. In general, the results of the two SPECTROM large strain codes compare favorably with results from other codes used to solve the problems.
Verification calculations for the WWER version of the TRANSURANUS code
International Nuclear Information System (INIS)
Elenkov, D.; Boneva, S.; Georgieva, M.; Georgiev, S.; Schubert, A.; Van Uffelen, P.
2006-01-01
The paper presents part of the work performed in the study project 'Research and Development for Licensing of Nuclear Fuel in Bulgaria'. The main objective of the project is to provide assistance for solving technical questions of the fuel licensing process in Bulgaria. One important issue is the extension of the predictive capabilities of fuel performance codes for Russian-type WWER reactors. In the last decade, a series of international projects has been based on the TRANSURANUS fuel performance code: Specific models for WWER fuel have been developed and implemented in the code in the late 90's. In 2000-2003, basic verification work was done by using experimental data of nuclear fuel irradiated in WWER-440 reactors. While the present paper focuses on the analysis of WWER-1000 standard fuel under normal operating conditions, the above study project covers additional tasks: 1) Post-irradiation calculations of ramp tests performed in the DR3 test reactor of the Risoe National Laboratory (five instrumented fuel rods of the Risoe 3 dataset contained in the IFPE database) using the TRANSURANUS code; 2) Compilation of cross-section libraries for isotope evolution calculations in WWER-440 and WWER-1000 fuel assemblies using the ORIGEN-S code; 3) Analysis of current situation and needs for an extension of the curriculum in Nuclear Engineering at the Technical University of Sofia. In this paper the post-irradiation calculations of steady-state irradiation experiments with nuclear fuel for Russian-type WWER-1000 reactors, using the latest release of the TRANSURANUS code (v1m1j03)are presented. Regarding a comprehensive verification of modern fuel performance codes, the burn-up region above 40 MWd/kgU is of increasing importance. A number of new phenomena emerge at high fuel burn-up, implying the need for enlarged databases of postirradiation examinations (PIE). For one fuel assembly irradiated in a WWER-1000 reactor with a rod discharge burn-up between 50 and 55 MWd
The Calculation of Flooding Level using CFX Code
International Nuclear Information System (INIS)
Oh, Seo Bin; Kim, Keon Yeop; Lee, Hyung Ho
2015-01-01
The plant design should consider internal flooding by postulated pipe ruptures, component failures, actuation of spray systems, and improper system alignment. The flooding causes failure of safety-related equipment and affects the integrity of the structure. The safety-related equipment should be installed above the flood level for protection against flooding effects. Conservative estimates of the flood level are important when a DBA occurs. The flooding level can be calculated simply applying Bernoulli's equation. However, in this study, a realistic calculation is performed with ANSYS CFX code. In calculation with CFX, air-core vortex phenomena, and turbulent flow can be simulated, which cannot be calculated analytically. The flooding level is evaluated by analytical calculation and CFX analysis for an assumed condition. The flood level is calculated as 0.71m and 1.1m analytically and with CFX simulation, respectively. Comparing the analytical calculation and simulation, they are similar, but the analytical calculation is not conservative. There are many factors reducing the drainage capacity such as air-core vortex, intake of air, and turbulent flow. Therefore, in case of flood level evaluation by analytical calculation, a sufficient safety margin should be considered
Opacity calculations for extreme physical systems: code RACHEL
Drska, Ladislav; Sinor, Milan
1996-08-01
Computer simulations of physical systems under extreme conditions (high density, temperature, etc.) require the availability of extensive sets of atomic data. This paper presents basic information on a self-consistent approach to calculations of radiative opacity, one of the key characteristics of such systems. After a short explanation of general concepts of the atomic physics of extreme systems, the structure of the opacity code RACHEL is discussed and some of its applications are presented.
Calculation code PULCO for Purex process in pulsed column
International Nuclear Information System (INIS)
Gonda, Kozo; Matsuda, Teruo
1982-03-01
The calculation code PULCO, which can simulate the Purex process using a pulsed column as an extractor, has been developed. The PULCO is based on the fundamental concept of mass transfer that the mass transfer within a pulsed column occurs through the interface of liquid drops and continuous phase fluid, and is the calculation code different from conventional ones, by which various phenomena such as the generation of liquid drops, their rising and falling, and the unification of liquid drops actually occurring in a pulsed column are exactly reflected and can be correctly simulated. In the PULCO, the actually measured values of the fundamental quantities representing the extraction behavior of liquid drops in a pulsed column are incorporated, such as the mass transfer coefficient of each component, the diameter and velocity of liquid drops in a pulsed column, the holdup of dispersed phase, and axial turbulent flow diffusion coefficient. The verification of the results calculated with the PULCO was carried out by installing a pulsed column of 50 mm inside diameter and 2 m length with 40 plate stage in a glove box for unirradiated uranium-plutonium mixed system. The results of the calculation and test were in good agreement, and the validity of the PULCO was confirmed. (Kako, I.)
Direct Calculations of Current Drive with a Full Wave Code
Wright, John C.; Phillips, Cynthia K.
1997-11-01
We have developed a current drive package that evaluates the current driven by fast magnetosonic waves in arbitrary flux geometry. An expression for the quasilinear flux has been derived which accounts for coupling between modes in the spectrum of waves launched from the antenna. The field amplitudes are calculated in the full wave code, FISIC, and the current response function, \\chi, also known as the Spitzer function, is determined with Charles Karney's Fokker-Planck code, adj.f. Both codes have been modified to incorporate the same numerical equilibria. To model the effects of a trapped particle population, the bounce averaged equations for current and power are used, and the bounce averaged flux is calculated. The computer model is benchmarked against the homogenous equations for a high aspect ratio case in which the expected agreement is confirmed. Results from cases for TFTR, NSTX and CDX-U are contrasted with the predictions of the Ehst-Karney parameterization of current drive for circular equilibria. For theoretical background, please see the authors' archive of papers. (http://w3.pppl.gov/ ~jwright/Publications)
Calculation of reactor pressure vessel fluence using TORT code
International Nuclear Information System (INIS)
Shin, Chul Ho; Kim, Jong Kyung
1998-01-01
TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Unit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) for all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library, BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The maximum fast neutron fluence calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effective full power days is 1.784x10 18 n/cm 2 . The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60 cm below the midplane at zero degree
PCRELAP5: data calculation program for RELAP 5 code
International Nuclear Information System (INIS)
Silvestre, Larissa Jacome Barros
2016-01-01
Nuclear accidents in the world led to the establishment of rigorous criteria and requirements for nuclear power plant operations by the international regulatory bodies. By using specific computer programs, simulations of various accidents and transients likely to occur at any nuclear power plant are required for certifying and licensing a nuclear power plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most widely used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors in Brazil and worldwide. A major difficulty in the simulation by using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data requires a great number of mathematical operations to calculate the geometry of the components. Thus, for those calculations performance and preparation of RELAP5 input data, a friendly mathematical preprocessor was designed. The Visual Basic for Application (VBA) for Microsoft Excel demonstrated to be an effective tool to perform a number of tasks in the development of the program. In order to meet the needs of RELAP5 users, the RELAP5 Calculation Program (Programa de Calculo do RELAP5 - PCRELAP5) was designed. The components of the code were codified; all entry cards including the optional cards of each one have been programmed. In addition, an English version for PCRELAP5 was provided. Furthermore, a friendly design was developed in order to minimize the time of preparation of input data and errors committed by users. In this work, the final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra 2. (author)
Energy Technology Data Exchange (ETDEWEB)
Jan, S.; Laedermann, J.P.; Bochud, F.; Ferragut, A.; Bordy, J.M.; Parisi, L.L.; Abou-Khalil, R.; Longeot, M.; Kitsos, S.; Groetz, J.E.; Villagrasa, C.; Daures, J.; Martin, E.; Henriet, J.; Tsilanizara, A.; Farah, J.; Uyttenhove, W.; Perrot, Y.; De Carlan, L.; Vivier, A.; Kodeli, I.; Sayah, R.; Hadid, L.; Courageot, E.; Fritsch, P.; Davesne, E.; Michel, X.
2010-07-01
This document gathers the slides of the available presentations given during these conference days. Twenty seven presentations are assembled in the document and deal with: 1 - GATE: calculation code for medical imaging, radiotherapy and dosimetry (S. Jan); 2 - estimation of conversion factors for the measurement of the ambient dose equivalent rate by in-situ spectroscopy (J.P. Laedermann); 3 - geometry specific calibration factors for nuclear medicine activity meters (F. Bochud); 4 - Monte Carlo simulation of a rare gases measurement system - calculation and validation, ASGA/VGM system (A. Ferragut); 5 - design of a realistic radiation field for the calibration of the dosemeters used in interventional radiology/cardiology (medical personnel dosimetry) (J.M. Bordy); 6 - determination of the position and height of the KALINA facility chimney at CEA Cadarache (L.L. Parisi); 7 - MERCURAD{sup TM} - 3D simulation software for dose rates calculation (R. Abou-Khalil); 8 - PANTHERE - 3D software for gamma dose rates simulation of complex nuclear facilities (M. Longeot); 9 - radioprotection, from the design to the exploitation of radioactive materials transportation containers (S. Kitsos); 10 - post-simulation processing of MCNPX responses in neutron spectroscopy (J.E. Groetz); 11 - last developments of the Geant4 Monte Carlo code for trace amounts simulation in liquid water at the molecular scale (C. Villagrasa); 12 - Calculation of H{sub p}(3)/K{sub air} conversion coefficients using PENELOPE Monte-Carlo code and comparison with MCNP calculation results (J. Daures); 13 - artificial neural networks, a new alternative to Monte Carlo calculations for radiotherapy (E. Martin); 14 - use of case-based reasoning for the reconstruction and handling of voxelized fantoms (J. Henriet); 15 - resolution of the radioactive decay inverse problem for dose calculation in radioprotection (A. Tsilanizara); 16 - use of NURBS-type fantoms for the study of the morphological factors influencing
Development and application of critical calculation code KENO
International Nuclear Information System (INIS)
Li Sumei; Jin Wenmian
1995-01-01
KENO is a large code of Monte Carlo critical calculation, compiled by Oak Ridge National laboratory of U.S.A. and widely used in the world. It is improved all the time from the first appearing at 1969. Now it is up to KENO IV, KENO Va. They have great capacity of geometry, can be used in any problems of complex 3-dimension geometric figure, meanwhile, they can use cross section library of Hansen Roach 16 groups or AMPX type with their better ability of cross section. In addition, KENO Va has the characteristic in deal with large groups, which can save computer RAM. The program gives K eff as the main calculation. The paper introduces the major features of KENO program, its improvement, development, micro-computerization and popularized application
Burnup calculations using serpent code in accelerator driven thorium reactors
International Nuclear Information System (INIS)
Korkmaz, M.E.; Agar, O.; Yigit, M.
2013-01-01
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232 Th and mixed 233 U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
Burnup calculations using serpent code in accelerator driven thorium reactors
Energy Technology Data Exchange (ETDEWEB)
Korkmaz, M.E.; Agar, O. [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Physics Dept.; Yigit, M. [Aksaray Univ. (Turkey). Physics Dept.
2013-07-15
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed {sup 232}Th and mixed {sup 233}U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
Calculation of local boron dilution accidents with the Hextran code
International Nuclear Information System (INIS)
Kyrki-Rajamaeki, R.; Stenius, T.
1995-01-01
Possibilities of Reactivity Initiated Accidents (RIA) due to local boron dilution slugs entering the core of PWRs have been widely studied in recent years. In Finland the main analysis tool for reactor dynamics RIA calculations has been the three dimensional HEXTRAN code which also includes full circuit models. Reliable calculation of propagating boron fronts is very difficult with standard numerical algorithms because numerical diffusion tends to smoothen the front. Thus the reactivity effect of the boron dilution can be significantly lowered and conservatism of the analyses cannot be guaranteed. In normal flow conditions this problem has been avoided in HEXTRAN analyses by simulating the dilution front directly to the core inlet. In natural circulation conditions there occurs significant numerical diffusion even during the propagation of boron front inside the core. Therefore a new hydraulics solution method PLIM (Piecewise Linear Interpolation Method) has been applied to HEXTRAN. Examples are given of analyses made with HEXTRAN in both flow conditions
The PHREEQE Geochemical equilibrium code data base and calculations
International Nuclear Information System (INIS)
Andersoon, K.
1987-01-01
Compilation of a thermodynamic data base for actinides and fission products for use with PHREEQE has begun and a preliminary set of actinide data has been tested for the PHREEQE code in a version run on an IBM XT computer. The work until now has shown that the PHREEQE code mostly gives satisfying results for specification of actinides in natural water environment. For U and Np under oxidizing conditions, however, the code has difficulties to converge with pH and Eh conserved when a solubility limit is applied. For further calculations of actinide and fission product specification and solubility in a waste repository and in the surrounding geosphere, more data are needed. It is necessary to evaluate the influence of the large uncertainties of some data. A quality assurance and a check on the consistency of the data base is also needed. Further work with data bases should include: an extension to fission products, an extension to engineering materials, an extension to other ligands than hydroxide and carbonate, inclusion of more mineral phases, inclusion of enthalpy data, a control of primary references in order to decide if values from different compilations are taken from the same primary reference and contacts and discussions with other groups, working with actinide data bases, e.g. at the OECD/NEA and at the IAEA. (author)
Evaluation and validation of criticality codes for fuel dissolver calculations
International Nuclear Information System (INIS)
Santamarina, A.; Smith, H.J.; Whitesides, G.E.
1991-01-01
During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)
International Nuclear Information System (INIS)
Asai, Kiyoshi; Shinozawa, Naohisa; Ishikawa, Hirohiko; Chino, Masamichi; Hayashi, Takashi
1983-02-01
Three computer codes MATHEW, ADPIC of LLNL and GAMPUL of JAERI for prediction of wind field, concentration and external exposure rate of airborne radioactive materials are vectorized and the results are presented. Using the continuous equation of incompressible flow as a constraint, the MATHEW calculates the three dimensional wind field by a variational method. Using the particle-in -cell method, the ADPIC calculates the advection and diffusion of radioactive materials in three dimensional wind field and terrain, and gives the concentration of the materials in each cell of the domain. The GAMPUL calculates the external exposure rate assuming Gaussian plume type distribution of concentration. The vectorized code MATHEW attained 7.8 times speedup by a vector processor FACOM230-75 APU. The ADPIC and GAMPUL are estimated to attain 1.5 and 4 times speedup respectively on CRAY-1 type vector processor. (author)
PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation
International Nuclear Information System (INIS)
Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.
1995-01-01
1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the
Quasiparticle GW calculations within the GPAW electronic structure code
DEFF Research Database (Denmark)
Hüser, Falco
properties are to a large extent governed by the physics on the atomic scale, that means pure quantum mechanics. For many decades, Density Functional Theory has been the computational method of choice, since it provides a fairly easy and yet accurate way of determining electronic structures and related...... properties. However, it has several drawbacks. A conceptual problem is the diculty of interpreting the calculated results with respect to experimentally measured quantities, resulting in, for example, the “band gap problem” in semiconductors. A practical issue is the necessity of adapting the method......The GPAW electronic structure code, developed at the physics department at the Technical University of Denmark, is used today by researchers all over the world to model the structural, electronic, optical and chemical properties of materials. They address fundamental questions in material science...
DRAGON, Reactor Cell Calculation System with Burnup
International Nuclear Information System (INIS)
2007-01-01
1 - Description of program or function: DRAGON is a collection of models to simulate the neutronic behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations which can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. The user must supply cross sections. DRAGON can access directly standard microscopic cross-section libraries in the following formats: DRAGON, MATXS (TRANSX-CTR), WIMSD4, WIMS-AECL, and APOLLO. It has the capability of exchanging macroscopic and microscopic cross-section libraries with a code such as PSR-0206/TRANSX-CTR or PSR-0317/TRANSX-2 by the use of the GOXS and ISOTXS format files. Macroscopic cross sections can also be read in DRAGON via the input data stream. 2 - Method of solution: DRAGON contains a multigroup iterator conceived to control a number of different algorithms for the solution of the neutron transport equation. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are included in a source term. The current version, DRAGON 9 71124 (Release 3.02), which was released in January 1998, contains three such algorithms. The JPM option solves the integral transport equation using the interface current method applied to homogeneous blocks; the SYBIL option solves the integral transport equation using the collision probability method for simple one-dimensional (1-D) or two-dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; and the
Relative efficiency calculation of a HPGe detector using MCNPX code
International Nuclear Information System (INIS)
Medeiros, Marcos P.C.; Rebello, Wilson F.; Lopes, Jose M.; Silva, Ademir X.
2015-01-01
High-purity germanium detectors (HPGe) are mandatory tools for spectrometry because of their excellent energy resolution. The efficiency of such detectors, quoted in the list of specifications by the manufacturer, frequently refers to the relative full-energy peak efficiency, related to the absolute full-energy peak efficiency of a 7.6 cm x 7.6 cm (diameter x height) NaI(Tl) crystal, based on the 1.33 MeV peak of a 60 Co source positioned 25 cm from the detector. In this study, we used MCNPX code to simulate a HPGe detector (Canberra GC3020), from Real-Time Neutrongraphy Laboratory of UFRJ, to survey the spectrum of a 60 Co source located 25 cm from the detector in order to calculate and confirm the efficiency declared by the manufacturer. Agreement between experimental and simulated data was achieved. The model under development will be used for calculating and comparison purposes with the detector calibration curve from software Genie2000™, also serving as a reference for future studies. (author)
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
Transport calculations with the BALDUR code. Pt. 1
International Nuclear Information System (INIS)
Lackner, K.; Wunderlich, R.
1979-12-01
1-d transport calculations with the BALDUR-code are described for predicting the performance of ZEPHYR under D-T operation. Results presented in this report refer to the impurity-free case, and ion and electron heat conduction losses described by CHIsub(i) = neoclassical and CHIsub(e) = 6.25 x 10 17 /nsub(e) (cgs-units). A simple refuelling scenario taking account of the density limit for the ohmic heating phase, the contribution of neutral injection to the refuelling rate and the need for an approximately balanced D-T mixture at the instance of ignition is adopted. The heating scenario assumes a neutral injection beam with 160 keV particle energy in the main component, with a duration of 1.1 sec. Major radius compression by a factor of 1.5 starts 1 sec after the onset of neutral injection and lasts 100 msec. For this standard scenario the performance is studied in different density regimes and for different neutral injection powers. Under the above assumption ignition is predicted for total neutral injection powers < approx. 16 MW (9.6 MW in the main energy component) and average total β-values < 2.8%. Results including impurities, alternative scaling laws, and deviations from the standard scenario will be presented in another report. (orig.) 891 GG/orig. 892 HIS
Comparison of computer codes for calculating dynamic loads in wind turbines
Spera, D. A.
1978-01-01
The development of computer codes for calculating dynamic loads in horizontal axis wind turbines was examined, and a brief overview of each code was given. The performance of individual codes was compared against two sets of test data measured on a 100 KW Mod-0 wind turbine. All codes are aeroelastic and include loads which are gravitational, inertial and aerodynamic in origin.
Expanding of reactor power calculation model of RELAP5 code
International Nuclear Information System (INIS)
Lin Meng; Yang Yanhua; Chen Yuqing; Zhang Hong; Liu Dingming
2007-01-01
For better analyzing of the nuclear power transient in rod-controlled reactor core by RELAP5 code, a nuclear reactor thermal-hydraulic best-estimate system code, it is expected to get the nuclear power using not only the point neutron kinetics model but also one-dimension neutron kinetics model. Thus an existing one-dimension nuclear reactor physics code was modified, to couple its neutron kinetics model with the RELAP5 thermal-hydraulic model. The detailed example test proves that the coupling is valid and correct. (authors)
A FACSIMILE code for calculating void swelling, version VS1
International Nuclear Information System (INIS)
Windsor, M.; Bullough, R.; Wood, M.H.
1979-11-01
VS1 is the first of a series of FACSIMILE codes that are being made available to predict the swelling of materials under irradiation at different temperatures, using chemical rate equations for the point defect losses to voids, interstitial loops, dislocation network, grain boundaries and foil surfaces. In this report the rate equations used in the program are given together with a detailed description of the code and directions for its use. (author)
Calculation of neutron spectra produced in neutron generator target: Code testing.
Gaganov, V V
2018-03-01
DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.
Computer codes used in the calculation of high-temperature thermodynamic properties of sodium
Energy Technology Data Exchange (ETDEWEB)
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.
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
SRAC2006; A Comprehensive neutronics calculation code system
奥村 啓介; 久語 輝彦; 金子 邦男; 土橋 敬一郎
2007-01-01
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five ele...
Revised SWAT. The integrated burnup calculation code system
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)
2000-07-01
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)
Revised SWAT. The integrated burnup calculation code system
International Nuclear Information System (INIS)
Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide
2000-07-01
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)
International Nuclear Information System (INIS)
Benmansour, L.
1992-01-01
The present work shows a group of results, obtained by a neutronic study, concerning the TRIGA MARK II reactor and LIGHT WATER reactors. These studies aim to make cell and diffusion calculations. WIMS D-4 with extended library and DIXY programs are used and tested for those purposes. We also have proceeded to a qualification of WIMS code based on the fuel temperature coefficient calculations. 33 refs.; 23 figs.; 30 tabs. (author)
Calculation of power spectra for block coded signals
DEFF Research Database (Denmark)
Justesen, Jørn
2001-01-01
We present some improvements in the procedure for calculating power spectra of signals based on finite state descriptions and constant block size. In addition to simplified calculations, our results provide some insight into the form of the closed expressions and to the relation between the spectra...
Development of the multistep compound process calculation code
Energy Technology Data Exchange (ETDEWEB)
Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan)
1998-03-01
A program `cmc` has been developed to calculate the multistep compound (MSC) process by Feshback-Kerman-Koonin. A radial overlap integral in the transition matrix element is calculated microscopically, and comparisons are made for neutron induced {sup 93}Nb reactions. Strengths of the two-body interaction V{sub 0} are estimated from the total MSC cross sections. (author)
High speed coding for velocity by archerfish retinal ganglion cells
Directory of Open Access Journals (Sweden)
Kretschmer Viola
2012-06-01
Full Text Available Abstract Background Archerfish show very short behavioural latencies in response to falling prey. This raises the question, which response parameters of retinal ganglion cells to moving stimuli are best suited for fast coding of stimulus speed and direction. Results We compared stimulus reconstruction quality based on the ganglion cell response parameters latency, first interspike interval, and rate. For stimulus reconstruction of moving stimuli using latency was superior to using the other stimulus parameters. This was true for absolute latency, with respect to stimulus onset, as well as for relative latency, with respect to population response onset. Iteratively increasing the number of cells used for reconstruction decreased the calculated error close to zero. Conclusions Latency is the fastest response parameter available to the brain. Therefore, latency coding is best suited for high speed coding of moving objects. The quantitative data of this study are in good accordance with previously published behavioural response latencies.
International Nuclear Information System (INIS)
Heo, B. G.; Jung, C. H.; Yoon, H. Y.; Yeo, D. J.; Song, C. H.
2002-01-01
A multidimensional numerical code for solving incompressible two-fluid is presented based on the Finite Volume Method (FVM) and the Simplified Marker And Cell (SMAC) method. Details of the present method and comparisons between the calculation and experiment are described for two-dimensional flow patterns of bubbly flow which show good agreement. Further implementations of the interfacial correlations are required for the application of the present code to various two-phase problems
MUXS: a code to generate multigroup cross sections for sputtering calculations
Energy Technology Data Exchange (ETDEWEB)
Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.
1982-10-01
This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc.
Improvement and test calculation on basic code or sodium-water reaction jet
International Nuclear Information System (INIS)
Saito, Yoshinori; Itooka, Satoshi; Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo
1999-03-01
In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)
Fuel management and core design code systems for pressurized water reactor neutronic calculations
International Nuclear Information System (INIS)
Ahnert, C.; Arayones, J.M.
1985-01-01
A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions
Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)
International Nuclear Information System (INIS)
Maiorino, J.R.
1990-01-01
The user manual of the ANISN computer code describing input and output subroutines is presented. ANISN code was developed to solve one-dimensional transport equation for neutron or gamma rays in slab, sphere or cylinder geometry with general anisotropic scattering. The solution technique is the discrete ordinate method. (M.C.K.)
ARTEMIS: a 3D transport code for shielding calculations
Energy Technology Data Exchange (ETDEWEB)
Varin, E.; Samba, G. [Commissariat a l' Energie Atomique, Bruyeres-Le-Chatels (France); Roy, R. [Ecole Polytechnique de Montreal, Montreal, Quebec (Canada)
2002-07-01
In radiation transport problems, as shielding applications, the solution of the Boltzmann transport equation is usually obtained by the discrete ordinates deterministic method. An alternative methodology has been developed in three dimensions into the code ARTEMIS. A Spherical Harmonics expansion of the angular flux has been chosen to guaranty solutions free of ray-effects. A least squares approach is applied over the linear transport equation; this approach leads to well-defined symmetric positive definite systems which allows the use of finite element spatial discretization. This paper presents the basic derivation of the discrete equations and provides examples on the use of this technique to solve different transport problems. (author)
Computational fluid mechanics qualification calculations for the code TEACH
International Nuclear Information System (INIS)
DeGrazia, M.C.; Fitzsimmons, L.B.; Reynolds, J.T.
1979-11-01
KAPL is developing a predictive method for three-dimensional (3-D) turbulent fluid flow configurations typically encountered in the thermal-hydraulic design of a nuclear reactor. A series of experiments has been selected for analysis to investigate the adequacy of the two-equation turbulence model developed at Imperial College, London, England for predicting the flow patterns in simple geometries. The analysis of these experiments is described with the two-dimensional (2-D) turbulent fluid flow code TEACH. This work qualifies TEACH for a variety of geometries and flow conditions
Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation
International Nuclear Information System (INIS)
Keco, M.; Debrecin, N.; Grgic, D.
2005-01-01
Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)
Code-B-1 for stress/strain calculation for TRISO fuel particle (Contract research)
International Nuclear Information System (INIS)
Aihara, Jun; Ueta, Shohei; Shibata, Taiju; Sawa, Kazuhiro
2011-12-01
We have developed Code-B-1 for the prediction of the failure probabilities of the coated fuel particles for the high temperature gas-cooled reactors (HTGRs) under operation by modification of an existing code. A finite element method (FEM) is employed for the stress calculation part and Code-B-1 can treat the plastic deformation of the coating layer of the coated fuel particles which the existing code cannot treat. (author)
Uncertainty analysis of RFSP-IST code flux calculation after cobalt adjuster rod replacement
International Nuclear Information System (INIS)
Tang Chuntao; Yang Bo; Bi Guangwen; Wang Jun
2014-01-01
In physical analysis of the cobalt adjuster rods design change for Qinshan Phase Ⅲ, the incremental cross section used in RFSP-IST code for the cobalt adjuster rod was calculated by DRAGON, whose methodology was not exactly the same as the MULTICELL used for Qinshan Phase Ⅲ RFSAR (2007), hence it was necessary to do the uncertainty analysis for RFSP-IST code flux calculation after cobalt adjuster rod replacement. Based on the historical data from Unit 1 and Unit 2 of Qinshan Phase Ⅲ, the uncertainty analysis of RFSP-IST code flux calculation was done with 95/95 upper tolerance limit method. Numerical results show that the cobalt adjuster rod design change and different incremental cross section calculation codes do not impact the uncertainty of RFSP-IST code flux calculation. (authors)
Equivalence of computer codes for calculation of coincidence summing correction factors
International Nuclear Information System (INIS)
Vidmar, T.; Capogni, M.; Hult, M.; Hurtado, S.; Kastlander, J.; Lutter, G.; Lépy, M.-C.; Martinkovič, J.; Ramebäck, H.; Sima, O.; Tzika, F.; Vidmar, G.
2014-01-01
The aim of the study was to check for equivalence of computer codes that can perform calculations of true coincidence summing correction factors. All calculations were performed for a set of well-defined detector and sample parameters, without any reference to empirical data. For a p-type detector model the application of different codes resulted in satisfactory agreement in the calculated correction factors. For high-efficiency geometries in combination with an n-type detector and a radionuclide emitting abundant X-rays the results were scattered. - Highlights: • Measured and calculated true coincidence corrections differ in recent study. • We checked for equivalence of computer codes that can perform such calculations. • Codes compared with each other for a set of well-defined detector and sample models. • Satisfactory agreement between the codes for a p-type detector model. • Large differences for an n-type detector model and X-ray emitting radionuclides
Siragusa, Mattia; Baiocco, Giorgio; Fredericia, Pil M; Friedland, Werner; Groesser, Torsten; Ottolenghi, Andrea; Jensen, Mikael
2017-08-01
COmputation Of Local Electron Release (COOLER), a software program has been designed for dosimetry assessment at the cellular/subcellular scale, with a given distribution of administered low-energy electron-emitting radionuclides in cellular compartments, which remains a critical step in risk/benefit analysis for advancements in internal radiotherapy. The software is intended to overcome the main limitations of the medical internal radiation dose (MIRD) formalism for calculations of cellular S-values (i.e., dose to a target region in the cell per decay in a given source region), namely, the use of the continuous slowing down approximation (CSDA) and the assumption of a spherical cell geometry. To this aim, we developed an analytical approach, entrusted to a MATLAB-based program, using as input simulated data for electron spatial energy deposition directly derived from full Monte Carlo track structure calculations with PARTRAC. Results from PARTRAC calculations on electron range, stopping power and residual energy versus traveled distance curves are presented and, when useful for implementation in COOLER, analytical fit functions are given. Example configurations for cells in different culture conditions (V79 cells in suspension or adherent culture) with realistic geometrical parameters are implemented for use in the tool. Finally, cellular S-value predictions by the newly developed code are presented for different cellular geometries and activity distributions (uniform activity in the nucleus, in the entire cell or on the cell surface), validated against full Monte Carlo calculations with PARTRAC, and compared to MIRD standards, as well as results based on different track structure calculations (Geant4-DNA). The largest discrepancies between COOLER and MIRD predictions were generally found for electrons between 25 and 30 keV, where the magnitude of disagreement in S-values can vary from 50 to 100%, depending on the activity distribution. In calculations for
The use of the codes from MCU family for calculations of WWER type reactors
International Nuclear Information System (INIS)
Abagijan, L.P.; Alexeyev, N.I.; Bryzgalov, V.I.; Gomin, E.A.; Glushkov, A.E.; Gorodkov, S.S.; Gurevich, M.I.; Kalugin, M.A.; Marin, S.V.; Shkarovsky, D.A.; Yudkevich, M.S.
2000-01-01
The MCU-RFFI/A and MCU-REA codes developed within the framework of the long term MCU project are widely used for calculations of neutron physic characteristics of WWER type reactors. Complete descriptions of the codes are available in both Russian and English. The codes are verified and validated by means of the comparison of calculated results with experimental data and mathematical benchmarks. The codes are licensed by Russian Nuclear and Criticality Safety Regulatory Body (Gosatomnadzor RF) (Code Passports: N 61 of 17.10.1966 and N 115 of 02.03.2000 accordingly)). The report gives examples of WWER reactor physic tasks important for practice solved using the codes from the MCU family. Some calculational results are given too. (Authors)
SAMDIST A Computer Code for Calculating Statistical Distributions for R-Matrix Resonance Parameters
Leal, L C
1995-01-01
The: SAMDIST computer code has been developed to calculate distribution of resonance parameters of the Reich-Moore R-matrix type. The program assumes the parameters are in the format compatible with that of the multilevel R-matrix code SAMMY. SAMDIST calculates the energy-level spacing distribution, the resonance width distribution, and the long-range correlation of the energy levels. Results of these calculations are presented in both graphic and tabular forms.
Calculation code; REPROSY-P for process studies on Pu purification with TBP
International Nuclear Information System (INIS)
Tsujino, Takeshi; Kohsaka, Atsuo; Aochi, Tetsuo
1975-10-01
To evaluate the plutonium purification process with tributhyl phosphate (TBP), the calculation code; REPROSY-P for TBP-HNO 3 -Pu system, has been prepared on the basis of a batchwise counter-current extraction cascade model. With the code, the concentration profiles under transient or steady state condition and the number of theoretical stages required, can be calculated for a given process condition. Empirical distribution equations, distribution data of plutonium and the detailed calculation program, are described. (auth.)
International Nuclear Information System (INIS)
Ferraro, V.; Igawa, N.; Marinelli, V.
2010-01-01
A calculation code, named INLUX-DBR, is presented, which is a modified version of INLUX code, able to predict the illuminance distribution on the inside surfaces of a room with six walls and a window, and on the work plane. At each desired instant the code solves the system of the illuminance equations of each surface element, characterized by the latter's reflection coefficient and its view factors toward the other elements. In the model implemented in the code, the sky-diffuse luminance distribution, the sun beam light and the light reflected from the ground toward the room are considered. The code was validated by comparing the calculated values of illuminance with the experimental values measured inside a scale model (1:5) of a building room, in various sky conditions of overcast, clear and intermediate days. The validation is performed using the sky luminance data measured by a sky scanner and the measured beam illuminance of the sun as input data. A comparative analysis of some of the well-known calculation models of sky luminance, namely Perez, Igawa and CIE models was also carried out, comparing the code predictions and the measured values of inside illuminance in the scale model.
Dose calculation on voxels phantoms using the GEANT4 code
International Nuclear Information System (INIS)
Martins, Maximiano C.; Santos, Denison S.; Queiroz Filho, Pedro P.; Begalli, Marcia
2009-01-01
This work implemented an anthropomorphic phantom of voxels on the structure of Monte Carlo GEANT4, for utilization by professionals from the radioprotection, external dosimetry and medical physics. This phantom allows the source displacement that can be isotropic punctual, plain beam, linear or radioactive gas, in order to obtain diverse irradiation geometries. In them, the radioactive sources exposure is simulated viewing the determination of effective dose or the dose in each organ of the human body. The Zubal head and body trunk phantom was used, and we can differentiate the organs and tissues by the chemical constitution in soft tissue, lung tissue, bone tissue, water and air. The calculation method was validated through the comparison with other well established method, the Visual Monte Carlo (VMC). Besides, a comparison was done with the international recommendation for the evaluation of dose by exposure to punctual sources, described in the document TECDOC - 1162- Generic Procedures for Assessment and Response During a Radiological Emergency, where analytical expressions for this calculation are given. Considerations are made on the validity limits of these expressions for various irradiation geometries, including linear sources, immersion into clouds and contaminated soils
Code Calculations for Local Stability of Shaft Guides
Directory of Open Access Journals (Sweden)
Fiołek Przemysław
2017-09-01
Full Text Available Steel structures for a conveyance guiding system are subjected to prolonged, intense corrosion during their operation leading to a considerable loss of material and structure capacity reduction. Shaft guides are made of closed profiles welded from hot-rolled channel sections. These profiles are categorized as class 1 cross-sections according to Eurocode 3, which means that they are resistant to local instability upon bending [1]. With an increase in the corrosion loss of the guides, the inertia moment of the cross-section is reduced. The resistance of profiles to local buckling is also reduced. However, calculations for local stability in guides upon bending are not required by the local Polish regulations on the operation of conveyance in shafts [2]. The question is whether this constitutes a shortcoming and risk for safe operation. Calculations according to steel construction standards [1] supported by numerical simulation were used to evaluate shaft steelwork guides resistance to buckling and their sensitivity to corrosion loss. It was shown that the guides of corrosion loss of 52–63%, depending on profile size, are prone to local buckling.
International Nuclear Information System (INIS)
Daures, J.; Gouriou, J.; Bordy, J.M.
2010-01-01
The authors report calculations performed using the MNCP and PENELOPE codes to determine the Hp(3)/K air conversion coefficient which allows the Hp(3) dose equivalent to be determined from the measured value of the kerma in the air. They report the definition of the phantom, a 20 cm diameter and 20 cm high cylinder which is considered as representative of a head. Calculations are performed for an energy range corresponding to interventional radiology or cardiology (20 keV-110 keV). Results obtained with both codes are compared
Determination of Solution Accuracy of Numerical Schemes as Part of Code and Calculation Verification
Energy Technology Data Exchange (ETDEWEB)
Blottner, F.G.; Lopez, A.R.
1998-10-01
This investigation is concerned with the accuracy of numerical schemes for solving partial differential equations used in science and engineering simulation codes. Richardson extrapolation methods for steady and unsteady problems with structured meshes are presented as part of the verification procedure to determine code and calculation accuracy. The local truncation error de- termination of a numerical difference scheme is shown to be a significant component of the veri- fication procedure as it determines the consistency of the numerical scheme, the order of the numerical scheme, and the restrictions on the mesh variation with a non-uniform mesh. Genera- tion of a series of co-located, refined meshes with the appropriate variation of mesh cell size is in- vestigated and is another important component of the verification procedure. The importance of mesh refinement studies is shown to be more significant than just a procedure to determine solu- tion accuracy. It is suggested that mesh refinement techniques can be developed to determine con- sistency of numerical schemes and to determine if governing equations are well posed. The present investigation provides further insight into the conditions and procedures required to effec- tively use Richardson extrapolation with mesh refinement studies to achieve confidence that sim- ulation codes are producing accurate numerical solutions.
Verification of RRC Ki code package for neutronic calculations of WWER core with GD
International Nuclear Information System (INIS)
Aleshin, S.S.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Pavlov, V.I.; Pavlovitchev, A.M.; Sidorenko, V.D.; Tsvetkov, V.M.
2001-01-01
The report presented is concerned with verification results of TVS-M/PERMAK-A/BIPR-7A code package for WWERs neutronic calculation as applied to calculation of systems containing U-GD pins. The verification is based on corresponded benchmark calculations, data critical experiments and on operation data obtained WWER units with Gd. The comparison results are discussed (Authors)
Verification of 3-D generation code package for neutronic calculations of WWERs
International Nuclear Information System (INIS)
Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.
2000-01-01
Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)
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.
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
Energy Technology Data Exchange (ETDEWEB)
Perfetti, Christopher M [ORNL; Martin, William R [University of Michigan; Rearden, Bradley T [ORNL; Williams, Mark L [ORNL
2012-01-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the SHIFT Monte Carlo code within the Scale code package. The methods were used for several simple test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods.
International Nuclear Information System (INIS)
Young, P.G.; Arthur, E.D.
1977-11-01
A new multistep Hauser--Feshbach code that includes corrections for preequilibrium effects is described. The code can calculate up to 60 decay reactions (cross sections and energy spectra) in one computation, and thereby provide considerable flexibility for handling processes with complicated reaction chains. Input parameter setup, problem output, and subroutine descriptions are given along with a sample problem calculation. A brief theoretical description is also included. 8 figures, 3 tables
Energy Technology Data Exchange (ETDEWEB)
Strenge, D.L.; Peloquin, R.A.
1981-04-01
The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested.
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)
Refuelling design and core calculations at NPP Paks: codes and methods
International Nuclear Information System (INIS)
Pos, I.; Nemes, I.; Javor, E.; Korpas, L.; Szecsenyi, Z.; Patai-Szabo, S.
2001-01-01
This article gives a brief review of the computer codes used in the fuel management practice at NPP Paks. The code package consist of the HELIOS neutron and gamma transport code for preparation of few-group cross section library, the CERBER code to determine the optimal core loading patterns and the C-PORCA code for detailed reactor physical analysis of different reactor states. The last two programs have been developed at the NPP Paks. HELIOS gives sturdy basis for our neutron physical calculation, CERBER and C-PORCA programs have been enhanced in great extent for last years. Methods and models have become more detailed and accurate as regards the calculated parameters and space resolution. Introduction of a more advanced data handling algorithm arbitrary move of fuel assemblies can be followed either in the reactor core or storage pool. The new interactive WINDOWS applications allow easier and more reliable use of codes. All these computer code developments made possible to handle and calculate new kind of fuels as profiled Russian and BNFL fuel with burnable poison or to support the reliable reuse of fuel assemblies stored in the storage pool. To extend thermo-hydraulic capability, with KFKI contribution the COBRA code will also be coupled to the system (Authors)
Boltzmann-Fokker-Planck calculations using standard discrete-ordinates codes
International Nuclear Information System (INIS)
Morel, J.E.
1987-01-01
The Boltzmann-Fokker-Planck (BFP) equation can be used to describe both neutral and charged-particle transport. Over the past several years, the author and several collaborators have developed methods for representing Fokker-Planck operators with standard multigroup-Legendre cross-section data. When these data are input to a standard S/sub n/ code such as ONETRAN, the code actually solves the Boltzmann-Fokker-Planck equation rather than the Boltzmann equation. This is achieved wihout any modification to the S/sub n/ codes. Because BFP calculations can be more demanding from a numerical viewpoint than standard neutronics calculations, we have found it useful to implement new quadrature methods ad convergence acceleration methods in the standard discrete-ordinates code, ONETRAN. We discuss our BFP cross-section representation techniques, our improved quadrature and acceleration techniques, and present results from BFP coupled electron-photon transport calculations performed with ONETRAN. 19 refs., 7 figs
Code Betal to calculation Alpha/Beta activities in environmental samples
International Nuclear Information System (INIS)
Romero, L.; Travesi, A.
1983-01-01
A codes, BETAL, was developed, written in FORTRAN IV, to automatize calculations and presentations of the result of the total alpha-beta activities measurements in environmental samples. This code performs the necessary calculations for transformation the activities measured in total counts, to pCi/1., bearing in mind the efficiency of the detector used and the other necessary parameters. Further more, it appraise the standard deviation of the result, and calculus the Lower limit of detection for each measurement. This code is written in iterative way by screen-operator dialogue, and asking the necessary data to perform the calculation of the activity in each case by a screen label. The code could be executed through any screen and keyboard terminal, (whose computer accepts Fortran IV) with a printer connected to the said computer. (Author) 5 refs
Gaganov, V. V.
2017-12-01
The correctness of calculations performed with the SRIANG code for modeling the spectra of DT neutrons is estimated by comparing the obtained spectra to the results of calculations carried out with five different codes based on the Monte Carlo method.
International Nuclear Information System (INIS)
Ioki, Kimihiro; Harada, Yuhei; Asami, Naoto.
1976-03-01
The computer code ACTIVE has been prepared for calculation of the transmutation rate, induced activity and decay heat. Calculations are carried out with activation chain and spatial distribution of neutron energy spectrum. The spatial distribution of secondary gamma-ray source due to the unstable nuclides is also obtainable. Special attension is paid to the short life decays. (auth.)
MFE/ACT: a TRS-80 code for calculating neutron activation
International Nuclear Information System (INIS)
Dorn, D.W.
1982-10-01
The MFE/ACT code, written to run on the TRS-80, can be used to calculate the neutron activation of materials used in fission and fusion reactors. Input data include the specific isotopes to be calculated, the neutron fluxes, the neutron cross sections, and the nuclear decay half-lives
Audit calculations of accidents analysis for second unit of Ignalina NPP with ATHLET code
International Nuclear Information System (INIS)
Adomavicius, A.; Belousov, A.; Ognerubov, V.
2004-01-01
Background of thermo hydraulic processes audit calculations in the frame of RSR-2 project is presented. Assumptions for the design based accident - RBMK-1500 group distributor header break analysis and modeling are presented. Audit calculations by ATHLET code and evaluation of results were provided. (author)
Manual of Nucost 1.0 - code for calculation of nuclear power generation costs
International Nuclear Information System (INIS)
Mascarenhas, H.A.
1989-01-01
Nucost is a computer code developed at CDTN to perform cost calculation of electric power generated in PWR nuclear power plants, based on present worth cost method. The Nucost version 1.0 performs calculations of nuclear fuel cost cycle by cycle during the time life of the power plant. That calculation is performed with enough details permitting optimization and minimization. The code is also a tool to aid reload projects and economic operation of PWR reactors. This manual presents a description of Nucost version 1.0, instruction to enter data preparation and description of the Nucost output. (M.I.)
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.
DNBR calculation in digital core protection system by a subchannel analysis code
International Nuclear Information System (INIS)
In, W. K.; Yoo, Y. J.; Hwang, T. H.; Ji, S. K.
2001-01-01
The DNBR calculation uncertainty and DNBR margin were evaluated in digital core protection system by a thermal-hydrualic subchannel analysis code MATRA. A simplified thermal-hydraulic code CETOP is used to calculate on-line DNBR in core protection system at a digital PWR. The DNBR tuning process against a best-estimate subchannel analysis code is required for CETOP to ensure accurate and conservative DNBR calculation but not necessary for MATRA. The DNBR calculations by MATRA and CETOP were performed for a large number of operating condition in Yonggwang nulcear units 3-4 where the digitial core protection system is initially implemented in Korea. MATRA resulted in a less negative mean value (i.e., reduce the overconservatism) and a somewhat larger standard deviation of the DNBR error. The uncertainty corrected minimum DNBR by MATRA was shown to be higher by 1.8% -9.9% that the CETOP DNBR
International Nuclear Information System (INIS)
Kim, Jung-Ha; Hill, Robin; Kuncic, Zdenka
2012-01-01
The Monte Carlo (MC) method has proven invaluable for radiation transport simulations to accurately determine radiation doses and is widely considered a reliable computational measure that can substitute a physical experiment where direct measurements are not possible or feasible. In the EGSnrc/BEAMnrc MC codes, there are several user-specified parameters and customized transport algorithms, which may affect the calculation results. In order to fully utilize the MC methods available in these codes, it is essential to understand all these options and to use them appropriately. In this study, the effects of the electron transport algorithms in EGSnrc/BEAMnrc, which are often a trade-off between calculation accuracy and efficiency, were investigated in the buildup region of a homogeneous water phantom and also in a heterogeneous phantom using the DOSRZnrc user code. The algorithms and parameters investigated include: boundary crossing algorithm (BCA), skin depth, electron step algorithm (ESA), global electron cutoff energy (ECUT) and electron production cutoff energy (AE). The variations in calculated buildup doses were found to be larger than 10% for different user-specified transport parameters. We found that using BCA = EXACT gave the best results in terms of accuracy and efficiency in calculating buildup doses using DOSRZnrc. In addition, using the ESA = PRESTA-I option was found to be the best way of reducing the total calculation time without losing accuracy in the results at high energies (few keV ∼ MeV). We also found that although choosing a higher ECUT/AE value in the beam modelling can dramatically improve computation efficiency, there is a significant trade-off in surface dose uncertainty. Our study demonstrates that a careful choice of user-specified transport parameters is required when conducting similar MC calculations. (note)
Method for calculating internal radiation and ventilation with the ADINAT heat-flow code
International Nuclear Information System (INIS)
Butkovich, T.R.; Montan, D.N.
1980-01-01
One objective of the spent fuel test in Climax Stock granite (SFTC) is to correctly model the thermal transport, and the changes in the stress field and accompanying displacements from the application of the thermal loads. We have chosen the ADINA and ADINAT finite element codes to do these calculations. ADINAT is a heat transfer code compatible to the ADINA displacement and stress analysis code. The heat flow problem encountered at SFTC requires a code with conduction, radiation, and ventilation capabilities, which the present version of ADINAT does not have. We have devised a method for calculating internal radiation and ventilation with the ADINAT code. This method effectively reproduces the results from the TRUMP multi-dimensional finite difference code, which correctly models radiative heat transport between drift surfaces, conductive and convective thermal transport to and through air in the drifts, and mass flow of air in the drifts. The temperature histories for each node in the finite element mesh calculated with ADINAT using this method can be used directly in the ADINA thermal-mechanical calculation
Development of sump model for containment hydrogen distribution calculations using CFD code
Energy Technology Data Exchange (ETDEWEB)
Ravva, Srinivasa Rao, E-mail: srini@aerb.gov.in [Indian Institute of Technology-Bombay, Mumbai (India); Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India); Iyer, Kannan N. [Indian Institute of Technology-Bombay, Mumbai (India); Gaikwad, A.J. [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)
2015-12-15
Highlights: • Sump evaporation model was implemented in FLUENT using three different approaches. • Validated the implemented sump evaporation models against TOSQAN facility. • It was found that predictions are in good agreement with the data. • Diffusion based model would be able to predict both condensation and evaporation. - Abstract: Computational Fluid Dynamics (CFD) simulations are necessary for obtaining accurate predictions and local behaviour for carrying out containment hydrogen distribution studies. However, commercially available CFD codes do not have all necessary models for carrying out hydrogen distribution analysis. One such model is sump or suppression pool evaporation model. The water in the sump may evaporate during the accident progression and affect the mixture concentrations in the containment. Hence, it is imperative to study the sump evaporation and its effect. Sump evaporation is modelled using three different approaches in the present work. The first approach deals with the calculation of evaporation flow rate and sump liquid temperature and supplying these quantities through user defined functions as boundary conditions. In this approach, the mean values of the domain are used. In the second approach, the mass, momentum, energy and species sources arise due to the sump evaporation are added to the domain through user defined functions. Cell values adjacent to the sump interface are used in this. Heat transfer between gas and liquid is calculated automatically by the code itself. However, in these two approaches, the evaporation rate was computed using an experimental correlation. In the third approach, the evaporation rate is directly estimated using diffusion approximation. The performance of these three models is compared with the sump behaviour experiment conducted in TOSQAN facility.Classification: K. Thermal hydraulics.
Calculation code of mass and heat transfer in a pulsed column for Purex process
International Nuclear Information System (INIS)
Tsukada, Takeshi; Takahashi, Keiki
1993-01-01
A calculation code for extraction behavior analysis in a pulsed column employed at an extraction process of a reprocessing plant was developed. This code was also combined with our previously developed calculation code for axial temperature profiles in a pulsed column. The one-dimensional dispersion model was employed for both of the extraction behavior analysis and the axial temperature profile analysis. The reported values of the fluid characteristics coefficient, the transfer coefficient and the diffusivities in the pulsed column were used. The calculated concentration profiles of HNO 3 , U and Pu for the steady state have a good agreement with the reported experimental results. The concentration and temperature profiles were calculated under the operation conditions which induce the abnormal U extraction behavior, i.e. U extraction zone is moved to the bottom of the column. Thought there is slight difference between calculated and experimental value, it is appeared that our developed code can be applied to the simulation under the normal operation condition and the relatively slowly transient condition. Pu accumulation phenomena was analyzed with this code and the accumulation tendency is similar to the reported analysis results. (author)
Code accuracy evaluation of ISP 35 calculations based on NUPEC M-7-1 test
International Nuclear Information System (INIS)
Auria, F.D.; Oriolo, F.; Leonardi, M.; Paci, S.
1995-01-01
Quantitative evaluation of code uncertainties is a necessary step in the code assessment process, above all if best-estimate codes are utilised for licensing purposes. Aiming at quantifying the code accuracy, an integral methodology based on the Fast Fourier Transform (FFT) has been developed at the University of Pisa (DCMN) and has been already applied to several calculations related to primary system test analyses. This paper deals with the first application of the FFT based methodology to containment code calculations based on a hydrogen mixing and distribution test performed in the NUPEC (Nuclear Power Engineering Corporation) facility. It is referred to pre-test and post-test calculations submitted for the International Standard Problem (ISP) n. 35. This is a blind exercise, simulating the effects of steam injection and spray behaviour on gas distribution and mixing. The result of the application of this methodology to nineteen selected variables calculated by ten participants are here summarized, and the comparison (where possible) of the accuracy evaluated for the pre-test and for the post-test calculations of a same user is also presented. (author)
Energy Technology Data Exchange (ETDEWEB)
Vitruk, S.G.; Korsun, A.S. [Moscow Engineering Physics Institute (Russian Federation); Ushakov, P.A. [Institute of Physics and Power Engineering, Obninsk (R)] [and others
1995-09-01
The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors.
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
Calculation code evaluating the confinement of a nuclear facility in case of fires
Energy Technology Data Exchange (ETDEWEB)
Laborde, J.C.; Prevost, C.; Vendel, J. [and others
1995-02-01
Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation.
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Tanaka, Shun-ichi
1991-09-01
A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)
A computer code for calculations in the algebraic collective model of the atomic nucleus
Welsh, T. A.; Rowe, D. J.
2014-01-01
A Maple code is presented for algebraic collective model (ACM) calculations. The ACM is an algebraic version of the Bohr model of the atomic nucleus, in which all required matrix elements are derived by exploiting the model's SU(1,1) x SO(5) dynamical group. This paper reviews the mathematical formulation of the ACM, and serves as a manual for the code. The code enables a wide range of model Hamiltonians to be analysed. This range includes essentially all Hamiltonians that are rational functi...
The FLUFF code for calculating finned surface heat transfer -description and user's guide
International Nuclear Information System (INIS)
Fry, C.J.
1985-08-01
FLUFF is a computer code for calculating heat transfer from finned surfaces by convection and radiation. It can also represent heat transfer by radiation to a partially emitting and absorbing medium within the fin cavity. The FLUFF code is useful not only for studying the behaviour of finned surfaces but also for deriving heat fluxes which can be applied as boundary conditions to other heat transfer codes. In this way models of bodies with finned surfaces may be greatly simplified since the fins need not be explicitly represented. (author)
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
Benchmark calculations of the solution-fuel criticality experiments by SRAC code system
International Nuclear Information System (INIS)
Senuma, Ichiro; Miyoshi, Yoshinori; Suzaki, Takenori; Kobayashi, Iwao
1984-06-01
Benchmark calculations were performed by using newly developed SRAC (Standard Reactor Analysis Code) system and nuclear data library based upon JENDL-2. The 34 benchmarks include variety of composition, concentration and configuration of Pu homogeneous and U/Pu homogeneous systems (nitrate, mainly), also include UO 2 /PuO 2 rods in fissile solution: a simplified model of the dissolver process of the fuel reprocessing plant. Calculation results shows good agreement with Monte Carlo method. This code-evaluation work has been done for the the part of the Detailed Design of CSEF (Critical Satety Experimental Facility), which is now in Progress. (author)
Coil tests and superconductor code calculations for the stellarator W7-X coils
Baldzuhn, J.; Ehmler, H.; Hoelting, A.; Hertel, K.; Sborchia, C.; Genini, L.; Schild, T.
2006-07-01
For the stellarator Wendelstein 7-X, a plasma fusion experiment, the performance of the superconducting coils is tested in a cryogenic test facility. Focus is on the quench behaviour of these coils. Some key data of the coils are given here. The coil quench data, obtained during the tests, are compared to GANDALF code calculations. GANDALF is a one-dimensional finite elements code for the simulation of the quench properties of superconducting CICC cables. Good consistency between measurement and calculation is found for the development of the resistive voltage and temperature increase during the quench.
International Nuclear Information System (INIS)
Perfetti, C.; Martin, W.; Rearden, B.; Williams, M.
2012-01-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the Shift Monte Carlo code within the SCALE code package. The methods were used for two small-scale test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods. (authors)
Energy Technology Data Exchange (ETDEWEB)
Perfetti, C.; Martin, W. [Univ. of Michigan, Dept. of Nuclear Engineering and Radiological Sciences, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109-2104 (United States); Rearden, B.; Williams, M. [Oak Ridge National Laboratory, Reactor and Nuclear Systems Div., Bldg. 5700, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)
2012-07-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the Shift Monte Carlo code within the SCALE code package. The methods were used for two small-scale test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods. (authors)
Calculation of static harmonics of a nuclear reactor using CITATION code
International Nuclear Information System (INIS)
Belchior Junior, A.; Moreira, J.M.L.
1989-01-01
The CITATION code, which solves the multigroup diffusion equation by the finite difference method, calculates the fundamental λ-mode (harmonic) for nuclear reactors. In this work, two fission source correction methods are attempted to obtain higher λ-modes through the CITATION code. The two methods are compared, their advantages and disadvantages analysed and verified against analytical solutions. Two dimensional harmonic modes are calculated for the IEA-R1 research reactor and for the ANGRA-I power reactor. The results are shown in graphics and tables. (author) [pt
International Nuclear Information System (INIS)
Devine, R.T.; Hsu, Hsiao-Hua
1994-01-01
The current basis for conversion coefficients for calibrating individual photon dosimeters in terms of dose equivalents is found in the series of papers by Grosswent. In his calculation the collision kerma inside the phantom is determined by calculation of the energy fluence at the point of interest and the use of the mass energy absorption coefficient. This approximates the local absorbed dose. Other Monte Carlo methods can be sued to provide calculations of the conversion coefficients. Rogers has calculated fluence-to-dose equivalent conversion factors with the Electron-Gamma Shower Version 3, EGS3, Monte Carlo program and produced results similar to Grosswent's calculations. This paper will report on calculations using the Integrated TIGER Series Version 3, ITS3, code to calculate the conversion coefficients in ICRU Tissue and in PMMA. A complete description of the input parameters to the program is given and comparison to previous results is included
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
Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels
International Nuclear Information System (INIS)
Daverio, Hernando J.
2003-01-01
Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)
A FACSIMILE code for calculating void swelling and creep, with vacancy loops present: version VS4
International Nuclear Information System (INIS)
Windsor, M.E.; Bullough, R.; Wood, M.H.
1981-10-01
This FACSIMILE code calculates void swelling and creep of irradiated materials, taking into account the effects of cavities, interstitial loops, vacancy loops, dislocation network and either grain boundaries or foil surfaces. The creep calculations are based on SIPA theory (stress induced preferred absorption), with no preferred nucleation. Either interactive or non-interactive options are available for the sink strength equations, but rate limitation is not incorporated. FACSIMILE is a computer program for solving simultaneous differential equations, and this VS4 code is one of a series of codes for calculating void swelling using increasingly complex theories. Other reports describing the VS1 and VS2 codes explain their use under control of the TSO system of the Harwell IBM 3033 computer, and explain the basic organization of the codes as required for use by FACSIMILE. The creep theory assumes that the material is under a constant uniaxial tensile stress during the irradiation. Three directions are considered for network parameters relative to the direction of the stress, and two directions for interstitial and vacancy loops. To give a full picture of these various contributions to the total creep, a large set of output parameter values are printed for each demanded dose value via a FORTRAN subroutine. (author)
GRIMH3: A new reactor calculation code at Savannah River Site
International Nuclear Information System (INIS)
Le, T.T.; Pevey, R.E.
1993-01-01
The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex
Energy Technology Data Exchange (ETDEWEB)
Pavlovichev, A.M.
2001-06-19
The report presents calculation results of isotopic composition of irradiated fuel performed for the Quad Cities-1 reactor bundle with UO{sub 2} and MOX fuel. The MCU-REA code was used for calculations. The code is developed in Kurchatov Institute, Russia. The MCU-REA results are compared with the experimental data and HELIOS code results.
International Nuclear Information System (INIS)
Markova, L.
2001-01-01
As a sensitivity study the impact on the system reactivity was studied in the case that different calculational methodologies of spent fuel isotopic concentrations were used for WWER spent fuel inventory computations. The sets of isotopic concentrations obtained by calculations with different codes and libraries as a result of the CB2 international benchmark focused on WWER-440 burnup credit were used to show the spread of the calculated spent fuel system reactivity. Using the MCNP 4B code and changing the isotopics input data, the multiplication factor of an infinite array of the WWER-440 fuel pin cells was calculated. The evaluation of the results shows the sensitivity of the calculated reactivity to different calculational methodologies used for the spent fuel inventory computation. In the studied cases of the CB2 benchmark, the spread of the reference k-results relative to the mean was found less or about ±1% in spite of the fact that the data of isotopic concentrations were spread much more. (author)
Triplet correlations among similarly tuned cells impact population coding
Directory of Open Access Journals (Sweden)
Natasha Alexandra Cayco Gajic
2015-05-01
Full Text Available Which statistical features of spiking activity matter for how stimuli are encoded in neural populations? A vast body of work has explored how firing rates in individual cells and correlations in the spikes of cell pairs impact coding. Recent experiments have shown evidence for the existence of higher-order spiking correlations, which describe simultaneous firing in triplets and larger ensembles of cells; however, little is known about their impact on encoded stimulus information. Here, we take a first step toward closing this gap. We vary triplet correlations in small (approximately 10 cell neural populations while keeping single cell and pairwise statistics fixed at typically reported values. This connection with empirically observed lower-order statistics important, as it places strong constraints on the level of triplet correlations that can occur. For each value of triplet correlations, we estimate the performance of the neural population on a two-stimulus discrimination task. We find that the allowed changes in the level of triplet correlations can significantly enhance coding, in particular if triplet correlations differ for the two stimuli. In this scenario, triplet correlations must be included in order to accurately quantify the functionality of neural populations. When both stimuli elicit similar triplet correlations, however, pairwise models provide relatively accurate descriptions of coding accuracy. We explain our findings geometrically via the skew that triplet correlations induce in population-wide distributions of neural responses. Finally, we calculate how many samples are necessary to accurately measure spiking correlations of this type, providing an estimate of the necessary recording times in future experiments.
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
KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code
International Nuclear Information System (INIS)
Kim, Young Gyun; Kim, Young Il
2006-12-01
Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006
FEAST: a two-dimensional non-linear finite element code for calculating stresses
International Nuclear Information System (INIS)
Tayal, M.
1986-06-01
The computer code FEAST calculates stresses, strains, and displacements. The code is two-dimensional. That is, either plane or axisymmetric calculations can be done. The code models elastic, plastic, creep, and thermal strains and stresses. Cracking can also be simulated. The finite element method is used to solve equations describing the following fundamental laws of mechanics: equilibrium; compatibility; constitutive relations; yield criterion; and flow rule. FEAST combines several unique features that permit large time-steps in even severely non-linear situations. The features include a special formulation for permitting many finite elements to simultaneously cross the boundary from elastic to plastic behaviour; accomodation of large drops in yield-strength due to changes in local temperature and a three-step predictor-corrector method for plastic analyses. These features reduce computing costs. Comparisons against twenty analytical solutions and against experimental measurements show that predictions of FEAST are generally accurate to ± 5%
A model for bootstrap current calculations with bounce averaged Fokker-Planck codes
Westerhof, E.; Peeters, A.G.
1996-01-01
A model is presented that allows the calculation of the neoclassical bootstrap current originating from the radial electron density and pressure gradients in standard (2+1)D bounce averaged Fokker-Planck codes. The model leads to an electron momentum source located almost exclusively at the
Calculation report of FUMEX-II excercise by using INFRA code
Energy Technology Data Exchange (ETDEWEB)
Yang, Yong Sik; Lee, Chan Bock; Kim, Dae Ho; Kim, Young Min; Kim, Sun Ki; Bang, Je Geon
2005-12-15
Final calculation results of IAEA Coordinated Research Program, FUMEX-II, were reported. By using the selected performance database, INFRA code prediction capability was compared with measured data such as FGR and RIP, fuel temperature, radial fission product distribution, PCMI and cladding creep. INFRA prediction results showed good agreement with measured data.
International Nuclear Information System (INIS)
Zazula, J.M.
1983-01-01
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
SYN3D: a single-channel, spatial flux synthesis code for diffusion theory calculations
Energy Technology Data Exchange (ETDEWEB)
Adams, C. H.
1976-07-01
This report is a user's manual for SYN3D, a computer code which uses single-channel, spatial flux synthesis to calculate approximate solutions to two- and three-dimensional, finite-difference, multigroup neutron diffusion theory equations. SYN3D is designed to run in conjunction with any one of several one- and two-dimensional, finite-difference codes (required to generate the synthesis expansion functions) currently being used in the fast reactor community. The report describes the theory and equations, the use of the code, and the implementation on the IBM 370/195 and CDC 7600 of the version of SYN3D available through the Argonne Code Center.
SYN3D: a single-channel, spatial flux synthesis code for diffusion theory calculations
International Nuclear Information System (INIS)
Adams, C.H.
1976-07-01
This report is a user's manual for SYN3D, a computer code which uses single-channel, spatial flux synthesis to calculate approximate solutions to two- and three-dimensional, finite-difference, multigroup neutron diffusion theory equations. SYN3D is designed to run in conjunction with any one of several one- and two-dimensional, finite-difference codes (required to generate the synthesis expansion functions) currently being used in the fast reactor community. The report describes the theory and equations, the use of the code, and the implementation on the IBM 370/195 and CDC 7600 of the version of SYN3D available through the Argonne Code Center
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
Energy Technology Data Exchange (ETDEWEB)
Urata, Kazuhiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
2003-03-01
In design of the future fusion devises in which low activation ferritic steel is planned to use as the plasma facing material and/or the inserts for ripple reduction, the appreciation of the error field effect against the plasma as well as the optimization of ferritic plate arrangement to reduce the toroidal field ripple require calculation of magnetic field generated by ferritic steel. However iterative calculations concerning the non-linearity in B-H curve of ferritic steel disturbs high-speed calculation required as the design tool. In the strong toroidal magnetic field that is characteristic in the tokamak fusion devices, fully magnetic saturation of ferritic steel occurs. Hence a distribution of magnetic charges as magnetic field source is determined straightforward and any iteration calculation are unnecessary. Additionally objective ferritic steel geometry is limited to the thin plate and ferritic plates are installed along the toroidal magnetic field. Taking these special conditions into account, high-speed calculation code ''FEMAG'' has been developed. In this report, the formalization of 'FEMAG' code, how to use 'FEMAG', and the validity check of 'FEMAG' in comparison with a 3D FEM code, with the measurements of the magnetic field in JFT-2M are described. The presented examples are numerical results of design studies for JT-60 modification. (author)
DYNREL - the reference calculation (coupled code utilization on analysis of RIA-transient)
International Nuclear Information System (INIS)
Strmensky, C.; Darilek, P.
2003-01-01
DYNREL is coupled code, comprising DYN3D and RELAP5 programs. The coupled code has been developed during four years. Now DYNREL is tested on selected RIA and thermo-hydraulic transient calculations. This material describes some results from selected RIA transient calculation (initiated by control rod movement). DYNREL modelled the whole nuclear reactors. The core is modeled as 313 or 349 independent thermo-hydraulic channels with 10 or 20 axial layers. Thermo-hydraulic part contains about 700 components that covered the six loops' model of nuclear power plant in detail. The calculated results are compared with DYN3D/M3, DYN3D/H1.1 results (Authors)
Linear calculations of edge current driven kink modes with BOUT++ code
Li, G. Q.; Xu, X. Q.; Snyder, P. B.; Turnbull, A. D.; Xia, T. Y.; Ma, C. H.; Xi, P. W.
2014-10-01
This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density.
Linear calculations of edge current driven kink modes with BOUT++ code
Energy Technology Data Exchange (ETDEWEB)
Li, G. Q., E-mail: ligq@ipp.ac.cn; Xia, T. Y. [Institute of Plasma Physics, CAS, Hefei, Anhui 230031 (China); Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Xu, X. Q. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Snyder, P. B.; Turnbull, A. D. [General Atomics, San Diego, California 92186 (United States); Ma, C. H.; Xi, P. W. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); FSC, School of Physics, Peking University, Beijing 100871 (China)
2014-10-15
This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density.
The spectral code Apollo2: from lattice to 2D core calculations
Energy Technology Data Exchange (ETDEWEB)
Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I. [CEA Saclay (DEN/DANS/SERMA), 91 - Gif-sur-Yvette (France); Santamarina, A. [CEA Cadarache (CEA/DEN/DER/SPRC), 13 - Saint Paul lez Durance (France)
2005-07-01
Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations.
The spectral code Apollo2: from lattice to 2D core calculations
International Nuclear Information System (INIS)
Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I.; Santamarina, A.
2005-01-01
Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations
International Nuclear Information System (INIS)
Mann, F.M.
1998-01-01
The Tank Waste Remediation System (TWRS) is responsible for the safe storage, retrieval, and disposal of waste currently being held in 177 underground tanks at the Hanford Site. In order to successfully carry out its mission, TWRS must perform environmental analyses describing the consequences of tank contents leaking from tanks and associated facilities during the storage, retrieval, or closure periods and immobilized low-activity tank waste contaminants leaving disposal facilities. Because of the large size of the facilities and the great depth of the dry zone (known as the vadose zone) underneath the facilities, sophisticated computer codes are needed to model the transport of the tank contents or contaminants. This document presents the code selection criteria for those vadose zone analyses (a subset of the above analyses) where the hydraulic properties of the vadose zone are constant in time the geochemical behavior of the contaminant-soil interaction can be described by simple models, and the geologic or engineered structures are complicated enough to require a two-or three dimensional model. Thus, simple analyses would not need to use the fairly sophisticated codes which would meet the selection criteria in this document. Similarly, those analyses which involve complex chemical modeling (such as those analyses involving large tank leaks or those analyses involving the modeling of contaminant release from glass waste forms) are excluded. The analyses covered here are those where the movement of contaminants can be relatively simply calculated from the moisture flow. These code selection criteria are based on the information from the low-level waste programs of the US Department of Energy (DOE) and of the US Nuclear Regulatory Commission as well as experience gained in the DOE Complex in applying these criteria. Appendix table A-1 provides a comparison between the criteria in these documents and those used here. This document does not define the models (that
The concerted calculation of the BN-600 reactor for the deterministic and stochastic codes
Bogdanova, E. V.; Kuznetsov, A. N.
2017-01-01
The solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC “Kurchatov Institute” in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.
Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code
Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina
2018-02-01
The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.
Energy Technology Data Exchange (ETDEWEB)
Sher, R. [Rudolph Sher Associates, Stanford, CA (United States); Li, J. [Polestar Applied Technology, Inc., Los Altos, CA (United States)
1995-02-01
NAUAHYGROS is a computer code to calculate the behavior of fission product and other aerosol particles in the containment of a nuclear reactor following a severe accident. It is an extension of the German code NAUA, which has been in widespread use for many years. Early versions of NAUA treated various aerosol phenomena in dry atmospheres, including aerosol agglomeration, diffusion (plateout), and settling processes. Later versions added treatments of steam condensation on particles in saturated or supersaturated containment atmospheres. The importance of these condensation effects on aerosol removal rates was demonstrated in large scale simulated containment tests. The additional features incorporated in NAUAHYGROS include principally a treatment of steam condensation on hygroscopic aerosols, which can grow as a result of steam condensation even in superheated atmospheres, and improved modelling of steam condensation on the walls of the containment. The code has been validated against the LACE experiments.
The use of the SRIM code for calculation of radiation damage induced by neutrons
Mohammadi, A.; Hamidi, S.; Asadabad, Mohsen Asadi
2017-12-01
Materials subjected to neutron irradiation will being evolve to structural changes by the displacement cascades initiated by nuclear reaction. This study discusses a methodology to compute primary knock-on atoms or PKAs information that lead to radiation damage. A program AMTRACK has been developed for assessing of the PKAs information. This software determines the specifications of recoil atoms (using PTRAC card of MCNPX code) and also the kinematics of interactions. The deterministic method was used for verification of the results of (MCNPX+AMTRACK). The SRIM (formely TRIM) code is capable to compute neutron radiation damage. The PKAs information was extracted by AMTRACK program, which can be used as an input of SRIM codes for systematic analysis of primary radiation damage. Then the Bushehr Nuclear Power Plant (BNPP) radiation damage on reactor pressure vessel is calculated.
A study of power adaptation factor calculation of ONED94 code by using Bayes' theorem
International Nuclear Information System (INIS)
Lee, S. M.; Lee, E. C.; Oh, S. Y.
1999-01-01
The ONED94 code developed by KAERI is a 1-dimensional 2-group diffusion theory code reactor simulation. In this code, the adaptation method for Xenon, Power shape and Control rod worth is further employed to reduce the errors in the collapsing process of cross-section data. The current power adaptation method has turned out to show several problems as not calculating adequate adaptation factors according to core axial power shape. Consequently, the Detector response method is applied to solve this problem. This method, however, also has the difficulty that the detector response matrix must be known first. Subsequently, the new adaptation method using Bayesian theorem is employed to overcome these problems. As a result, the adaptation method using Bayesian theorem could easily and accurately find an adequate adaptation factor within 1% of relative error
Calculation of the RSG-GAS core using computer code citation-3D
International Nuclear Information System (INIS)
Taryo, T.; Rokhmadi
1998-01-01
Since core reactivity is one of the reactor safety parameters, this R and D has been carried out. To carry out the R and D, the code called WIMSD4 was used respectively for generating cross section and diffusion parameters. The code CITATION was then applied to estimate core reactivity in the RSG-GAS core. To verify the result of the calculation, data and information of the RSG-GAS Typical Working Core Were used. To Prove the codes reliably used, the case of all control elements down in the reactor core and that of all control rods up in the core were applied. The result taking into account those cases showed respectively that K eff are less and greater than unity (K eff eff >1)
International Nuclear Information System (INIS)
Mairani, A; Brons, S; Parodi, K; Cerutti, F; Ferrari, A; Sommerer, F; Fasso, A; Kraemer, M; Scholz, M
2010-01-01
Clinical Monte Carlo (MC) calculations for carbon ion therapy have to provide absorbed and RBE-weighted dose. The latter is defined as the product of the dose and the relative biological effectiveness (RBE). At the GSI Helmholtzzentrum fuer Schwerionenforschung as well as at the Heidelberg Ion Therapy Center (HIT), the RBE values are calculated according to the local effect model (LEM). In this paper, we describe the approach followed for coupling the FLUKA MC code with the LEM and its application to dose and RBE-weighted dose calculations for a superimposition of two opposed 12 C ion fields as applied in therapeutic irradiations. The obtained results are compared with the available experimental data of CHO (Chinese hamster ovary) cell survival and the outcomes of the GSI analytical treatment planning code TRiP98. Some discrepancies have been observed between the analytical and MC calculations of absorbed physical dose profiles, which can be explained by the differences between the laterally integrated depth-dose distributions in water used as input basic data in TRiP98 and the FLUKA recalculated ones. On the other hand, taking into account the differences in the physical beam modeling, the FLUKA-based biological calculations of the CHO cell survival profiles are found in good agreement with the experimental data as well with the TRiP98 predictions. The developed approach that combines the MC transport/interaction capability with the same biological model as in the treatment planning system (TPS) will be used at HIT to support validation/improvement of both dose and RBE-weighted dose calculations performed by the analytical TPS.
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)
International Nuclear Information System (INIS)
Ekberg, K.
1980-03-01
The radioactivity induced in PWR cluster control rods during reactor operation has been calculated using the computer programme ORIGEN. Neutron fluxes and spectrum conditions as well as the strongly shielded cross sections for the absorber materials Ag, In and Cd have been obtained by running the cell and assembly code CASMO for a couple of typical cases. The results show that Ag-110m, Fe-55 and Co-60 give the largest activity contributions in the interval 1-10 years after the end of irradiation, and Ni-63 and Cd-113m in a longer time perspective. (author)
Mairani, A; Kraemer, M; Sommerer, F; Parodi, K; Scholz, M; Cerutti, F; Ferrari, A; Fasso, A
2010-01-01
Clinical Monte Carlo (MC) calculations for carbon ion therapy have to provide absorbed and RBE-weighted dose. The latter is defined as the product of the dose and the relative biological effectiveness (RBE). At the GSI Helmholtzzentrum fur Schwerionenforschung as well as at the Heidelberg Ion Therapy Center (HIT), the RBE values are calculated according to the local effect model (LEM). In this paper, we describe the approach followed for coupling the FLUKA MC code with the LEM and its application to dose and RBE-weighted dose calculations for a superimposition of two opposed C-12 ion fields as applied in therapeutic irradiations. The obtained results are compared with the available experimental data of CHO (Chinese hamster ovary) cell survival and the outcomes of the GSI analytical treatment planning code TRiP98. Some discrepancies have been observed between the analytical and MC calculations of absorbed physical dose profiles, which can be explained by the differences between the laterally integrated depth-d...
MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA
International Nuclear Information System (INIS)
Okumura, Keisuke
2015-10-01
MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations. A CD-ROM is attached as an appendix. (J.P.N.)
Bécares, V.; Pérez Martín, S.; Vázquez Antolín, Miriam; Villamarín, D.; Martín Fuertes, Francisco; González Romero, E.M.; Merino Rodríguez, Iván
2014-01-01
The calculation of the effective delayed neutron fraction, beff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for beff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we...
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
Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes
International Nuclear Information System (INIS)
Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.
2016-01-01
To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)
First vapor explosion calculations performed with MC3D thermal-hydraulic code
Energy Technology Data Exchange (ETDEWEB)
Brayer, C.; Berthoud, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires
1998-01-01
This paper presents the first calculations performed with the `explosion` module of the multiphase computer code MC3D, which is devoted to the fine fragmentation and explosion phase of a fuel coolant interaction. A complete description of the physical laws included in this module is given. The fragmentation models, taking into account two fragmentation mechanisms, a thermal one and an hydrodynamic one, are also developed here. Results to some calculations to test the numerical behavior of MC3D and to test the explosion models in 1D or 2D are also presented. (author)
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.
Oliveira, C
2001-01-01
A systematic study of isodose distributions and dose uniformity in sample carriers of the Portuguese Gamma Irradiation Facility was carried out using the MCNP code. The absorbed dose rate, gamma flux per energy interval and average gamma energy were calculated. For comparison purposes, boxes filled with air and 'dummy' boxes loaded with layers of folded and crumpled newspapers to achieve a given value of density were used. The magnitude of various contributions to the total photon spectra, including source-dependent factors, irradiator structures, sample material and other origins were also calculated.
Energy Technology Data Exchange (ETDEWEB)
Mavroulakis, A.; Trombe, A. [INSA - Genie Civl, Laboratoire d`Etudes Thermiques et Mecaniques, 31 - Toulouse (France)
1996-12-31
This paper presents the main processes which allow to determine and to take into account in terms of form factors, a scene seen from an emitter and projected onto a receiver. The elements that compose the emitter have a triangular shape while no subdivision is made on the receiver. The analytical method used for the calculation of the form factors of one element in front of a polygonal receiver is briefly presented. Two cell configurations are presented, the second one having not convex facets with no prerequisite subdivision. The sums of form factors from one given emitter are less than 0.01 away from the unit value. For each configuration, the influence of obstacles is encoded as change rates of individual form factors. Finally, in order to illustrate the interest of these form factor calculations, an example of computerized simulation applied to a complex cavity is presented. (J.S.) 6 refs.
A group of neutronics calculations in the MNSR using the MCNP-4C code
International Nuclear Information System (INIS)
Khattab, K.; Sulieman, I.
2009-11-01
The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)
The PIES2012 Code for Calculating 3D Equilibria with Islands and Stochastic Regions
Monticello, Donald; Reiman, Allan; Raburn, Daniel
2013-10-01
We have made major modifications to the PIES 3D equilibrium code to produce a new version, PIES2012. The new version uses an adaptive radial grid for calculating equilibrium currents. A subset of the flux surfaces conform closely to island separatrices, providing an accurate treatment of the effects driving the neoclassical tearing mode. There is now a set of grid surfaces that conform to the flux surfaces in the interiors of the islands, allowing the proper treatment of the current profiles in the islands, which play an important role in tearing phenomena. We have verified that we can introduce appropriate current profiles in the islands to suppress their growth, allowing us to simulate situations where islands are allowed to grow at some rational surfaces but not others. Placement of grid surfaces between islands is guided by the locations of high order fixed points, allowing us to avoid spectral polution and providing a more robust, and smoother convergence of the code. The code now has an option for turning on a vertical magnetic field to fix the position of the magnetic axis, which models the horizontal feedback positioning of a tokamak plasma. The code has a new option for using a Jacobian-Free Newton Krylov scheme for convergence. The code now also contains a model that properly handles stochastic regions with nonzero pressure gradients. Work supported by DOE contract DE-AC02-09CH11466.
An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations
Energy Technology Data Exchange (ETDEWEB)
Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Nuclear Safety; Nikulshin, V. [Russian Research Center, Moscow (Russian Federation). Kurchatov Inst.
1996-09-01
This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases.
An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations
International Nuclear Information System (INIS)
Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E.; Nikulshin, V.
1996-09-01
This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases
Schweda, K
2002-01-01
The analysis of (e,e'n) experiments at the Darmstadt superconducting electron linear accelerator S-DALINAC required the calculation of neutron response functions for the NE213 liquid scintillation detectors used. In an open geometry, these response functions can be obtained using the Monte Carlo codes NRESP7 and NEFF7. However, for more complex geometries, an extended version of the Monte Carlo code MCNP exists. This extended version of the MCNP code was improved upon by adding individual light-output functions for charged particles. In addition, more than one volume can be defined as a scintillator, thus allowing the simultaneous calculation of the response for multiple detector setups. With the implementation of sup 1 sup 2 C(n,n'3 alpha) reactions, all relevant reactions for neutron energies E sub n <20 MeV are now taken into consideration. The results of these calculations were compared to experimental data using monoenergetic neutrons in an open geometry and a sup 2 sup 5 sup 2 Cf neutron source in th...
A Sample Calculation of Tritium Production and Distribution at VHTR by using TRITGO Code
Energy Technology Data Exchange (ETDEWEB)
Park, Ik Kyu; Kim, D. H.; Lee, W. J
2007-03-15
TRITGO code was developed for estimating the tritium production and distribution of high temperature gas cooled reactor(HTGR), especially GTMHR350 by General Atomics. In this study, the tritium production and distribution of NHDD was analyzed by using TRITGO Code. The TRITGO code was improved by a simple method to calculate the tritium amount in IS Loop. The improved TRITGO input for the sample calculation was prepared based on GTMHR600 because the NHDD has been designed referring GTMHR600. The GTMHR350 input with related to the tritium distribution was directly used. The calculated tritium activity among the hydrogen produced in IS-Loop is 0.56 Bq/g- H2. This is a very satisfying result considering that the limited tritium activity of Japanese Regulation Guide is 5.6 Bq/g-H2. The basic system to analyze the tritium production and the distribution by using TRITGO was successfully constructed. However, there exists some uncertainties in tritium distribution models, the suggested method for IS-Loop, and the current input was not for NHDD but for GTMHR600. The qualitative analysis for the distribution model and the IS-Loop model and the quantitative analysis for the input should be done in the future.
Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code
International Nuclear Information System (INIS)
Nunomiya, Tomoya; Iwase, Hiroshi; Nakamura, Takashi; Nakao, Noriaki
2002-03-01
A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were transported up to approximately 1 m before the region for benchmark calculation. Finally, the energy spectra of neutrons behind the very thick shield were calculated down to the thermal energy with good statistics, and typically agree well within a factor of two with the experimental data over a broad energy range. The 12 C(n,2n) 11 C reaction rates behind the bulk shield were also calculated, which agree with the experimental data typically within 60%. These results are quite impressive in calculation accuracy for deep-penetration problem. In this report, the calculation conditions, geometry and the variance reduction techniques used in the deep-penetration calculation with the MARS14 code are clarified, and several subroutines of MARS14 which were used in our calculation are also given in the appendix. The numerical data of the calculated neutron energy spectra, reaction rates, dose rates and their C/E (Calculation/Experiment) values are also summarized. The
Parameter calculation tool for the application of radiological dose projection codes
International Nuclear Information System (INIS)
Galindo G, I. F.; Vergara del C, J. A.; Galvan A, S. J.; Tijerina S, F.
2016-09-01
The use of specialized codes to estimate the radiation dose projection to an emergency postulated event at a nuclear power plant requires that certain plant data be available according to the event being simulated. The calculation of the possible radiological release is the critical activity to carry out the emergency actions. However, not all of the plant data required are obtained directly from the plant but need to be calculated. In this paper we present a computational tool that calculates the plant data required to use the radiological dose estimation codes. The tool provides the required information when there is a gas emergency venting event in the primary containment atmosphere, whether well or dry well and also calculates the time in which the spent fuel pool would be discovered in the event of a leak of water on some of the walls or floor of the pool. The tool developed has mathematical models for the processes involved such as: compressible flow in pipes considering area change and for constant area, taking into account the effects of friction and for the case of the spent fuel pool hydraulic models to calculate the time in which a container is emptied. The models implemented in the tool are validated with data from the literature for simulated cases. The results with the tool are very similar to those of reference. This tool will also be very supportive so that in postulated emergency cases can use the radiological dose estimation codes to adequately and efficiently determine the actions to be taken in a way that affects as little as possible. (Author)
Development of a relativistic Particle In Cell code PARTDYN for linear accelerator beam transport
Phadte, D.; Patidar, C. B.; Pal, M. K.
2017-04-01
A relativistic Particle In Cell (PIC) code PARTDYN is developed for the beam dynamics simulation of z-continuous and bunched beams. The code is implemented in MATLAB using its MEX functionality which allows both ease of development as well higher performance similar to a compiled language like C. The beam dynamics calculations carried out by the code are compared with analytical results and with other well developed codes like PARMELA and BEAMPATH. The effect of finite number of simulation particles on the emittance growth of intense beams has been studied. Corrections to the RF cavity field expressions were incorporated in the code so that the fields could be calculated correctly. The deviations of the beam dynamics results between PARTDYN and BEAMPATH for a cavity driven in zero-mode have been discussed. The beam dynamics studies of the Low Energy Beam Transport (LEBT) using PARTDYN have been presented.
Calculating the n-point correlation function with general and efficient python code
Genier, Fred; Bellis, Matthew
2018-01-01
There are multiple approaches to understanding the evolution of large-scale structure in our universe and with it the role of baryonic matter, dark matter, and dark energy at different points in history. One approach is to calculate the n-point correlation function estimator for galaxy distributions, sometimes choosing a particular type of galaxy, such as luminous red galaxies. The standard way to calculate these estimators is with pair counts (for the 2-point correlation function) and with triplet counts (for the 3-point correlation function). These are O(n2) and O(n3) problems, respectively and with the number of galaxies that will be characterized in future surveys, having efficient and general code will be of increasing importance. Here we show a proof-of-principle approach to the 2-point correlation function that relies on pre-calculating galaxy locations in coarse “voxels”, thereby reducing the total number of necessary calculations. The code is written in python, making it easily accessible and extensible and is open-sourced to the community. Basic results and performance tests using SDSS/BOSS data will be shown and we discuss the application of this approach to the 3-point correlation function.
Calculations to an IAHR-benchmark test using the CFD-code CFX-4
Energy Technology Data Exchange (ETDEWEB)
Krepper, E.
1998-10-01
The calculation concerns a test, which was defined as a benchmark for 3-D codes by the working group of advanced nuclear reactor types of IAHR (International Association of Hydraulic Research). The test is well documented and detailed measuring results are available. The test aims at the investigation of phenomena, which are important for heat removal at natural circulation conditions in a nuclear reactor. The task for the calculation was the modelling of the forced flow field of a single phase incompressible fluid with consideration of heat transfer and influence of gravity. These phenomena are typical also for other industrial processes. The importance of correct modelling of these phenomena also for other applications is a motivation for performing these calculations. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Tobita, Masahiro; Matsui, Yoshinori [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment
2003-03-01
Prediction of irradiation temperature is one of the important issues in the design of the capsule for irradiation test. Many kinds of capsules with complex structure have been designed for recent irradiation requests, and three-dimensional (3D) temperature calculation becomes inevitable for the evaluation of irradiation temperature. For such 3D calculation, however, many works are usually needed for input data preparation, and a lot of time and resources are necessary for parametric studies in the design. To improve such situation, JAERI introduced 3D-FEM (finite element method) code NISA (Numerically Integrated elements for System Analysis) and developed several subprograms, which enabled to support input preparation works in the capsule design. The 3D temperature calculation of the capsule are able to carried out in much easier way by the help of the subprograms, and specific features in the irradiation tests such as non-uniform gamma heating in the capsule, becomes to be considered. (author)
BEAVRS full core burnup calculation in hot full power condition by RMC code
International Nuclear Information System (INIS)
Liu, Shichang; Liang, Jingang; Wu, Qu; Guo, JuanJuan; Huang, Shanfang; Tang, Xiao; Li, Zeguang; Wang, Kan
2017-01-01
Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.
FOOD II: an interactive code for calculating concentrations of radionuclides in food products
International Nuclear Information System (INIS)
Zach, R.
1978-11-01
An interactive code, FOOD II, has been written in FORTRAN IV for the PDP 10 to allow calculation of concentrations of radionuclides in food products and internal doses to man under chronic release conditions. FOOD II uses models unchanged from a previous code, FOOD, developed at Battelle, Pacific Northwest Laboratories. The new code has different input and output features than FOOD and a number of options have been added to increase flexibility. Data files have also been updated. FOOD II takes into account contamination of vegetation by air and irrigation water containing radionuclides. Contamination can occur simultaneously by air and water. Both direct deposition of radionuclides on leaves, and their uptake from soil are possible. Also, animals may be contaminated by ingestion of vegetation and drinking water containing radionuclides. At present, FOOD II provides selection of 14 food types, 13 diets and numerous radionuclides. Provisions have been made to expand all of these categories. Six additional contaminated food products can also be entered directly into the dose model. Doses may be calculated for the total body and six internal organs. Summaries of concentrations in food products and internal doses to man can be displayed at a local terminal or at an auxiliary high-speed printer. (author)
Samsonov, Andrey; Gordeev, Evgeny; Sergeev, Victor
2017-04-01
As it was recently suggested (e.g., Gordeev et al., 2015), the global magnetospheric configuration can be characterized by a set of key parameters, such as the magnetopause distance at the subsolar point and on the terminator plane, the magnetic field in the magnetotail lobe and the plasma sheet thermal pressure, the cross polar cap electric potential drop and the total field-aligned current. For given solar wind conditions, the values of these parameters can be obtained from both empirical models and global MHD simulations. We validate the recently developed global MHD code SPSU-16 using the key magnetospheric parameters mentioned above. The code SPSU-16 can calculate both the isotropic and anisotropic MHD equations. In the anisotropic version, we use the modified double-adiabatic equations in which the T⊥/T∥ (the ratio of perpendicular to parallel thermal pressures) has been bounded from above by the mirror and ion-cyclotron thresholds and from below by the firehose threshold. The results of validation for the SPSU-16 code well agree with the previously published results of other global codes. Some key parameters coincide in the isotropic and anisotropic MHD simulations, but some are different.
CEQCSY: a new code for chemical equilibrium calculation in multiphased systems
International Nuclear Information System (INIS)
Lehmann, J.; Fabriol, R.
1989-01-01
As part of the CEC Chemval/mirage project, a method is presented for calculating the thermodynamic equilibrium state of a multiphase system, by minimizing its Gibbs free energy constrained by mass balances. Compared to the other algorithms available in the literature, the method has three main characteristics: - the sets of equations corresponding to the conditions of homogeneous and heterogeneous equilibria are simultaneously solved, - a mathematical criterion for bringing a new multicomponent phase in the system is rigorously demontrated. - It enables a detailed representation of the multisite solid solutions with constraints called site closure relation. The code CEQCSY (Chemical Equilibrium in Complex SYstem) uses this formalism, and works with the thermodynamic data base from the EQ3/6 code. This choice makes easier several compared tests with EQ6: quartz dissolution in water, water-atmospheric air equilibrium, theoretical re-equilibrium of seawater, hydrothermal alteration of granite including solid solutions. The test results demonstrate the high efficiency and velocity of the code CEQCSY, when working on equilibrium state of multiphase systems. This high velocity was the aim of this work, in order to couple with thermic, hydrodynamic or mechanic codes
SPARC-90: A code for calculating fission product capture in suppression pools
International Nuclear Information System (INIS)
Owczarski, P.C.; Burk, K.W.
1991-10-01
This report describes the technical bases and use of two updated versions of a computer code initially developed to serve as a tool for calculating aerosol particle retention in boiling water reactor (BWR) pressure suppression pools during severe accidents, SPARC-87 and SPARC-90. The most recent version is SPARC-90. The initial or prototype version (Owczarski, Postma, and Schreck 1985) was improved to include the following: rigorous treatment of local particle deposition velocities on the surface of oblate spherical bubbles, new correlations for hydrodynamic behavior of bubble swarms, models for aerosol particle growth, both mechanistic and empirical models for vent exit region scrubbing, specific models for hydrodynamics of bubble breakup at various vent types, and models for capture of vapor iodine species. A complete user's guide is provided for SPARC-90 (along with SPARC-87). A code description, code operating instructions, partial code listing, examples of the use of SPARC-90, and summaries of experimental data comparison studies also support the use of SPARC-90. 29 refs., 4 figs., 11 tabs
SPARC-90: A code for calculating fission product capture in suppression pools
Energy Technology Data Exchange (ETDEWEB)
Owczarski, P.C.; Burk, K.W. (Pacific Northwest Lab., Richland, WA (United States))
1991-10-01
This report describes the technical bases and use of two updated versions of a computer code initially developed to serve as a tool for calculating aerosol particle retention in boiling water reactor (BWR) pressure suppression pools during severe accidents, SPARC-87 and SPARC-90. The most recent version is SPARC-90. The initial or prototype version (Owczarski, Postma, and Schreck 1985) was improved to include the following: rigorous treatment of local particle deposition velocities on the surface of oblate spherical bubbles, new correlations for hydrodynamic behavior of bubble swarms, models for aerosol particle growth, both mechanistic and empirical models for vent exit region scrubbing, specific models for hydrodynamics of bubble breakup at various vent types, and models for capture of vapor iodine species. A complete user's guide is provided for SPARC-90 (along with SPARC-87). A code description, code operating instructions, partial code listing, examples of the use of SPARC-90, and summaries of experimental data comparison studies also support the use of SPARC-90. 29 refs., 4 figs., 11 tabs.
Progress on accelerated calculation of 3D MHD equilibrium with the PIES code
Raburn, Daniel; Reiman, Allan; Monticello, Donald
2016-10-01
Continuing progress has been made in accelerating the 3D MHD equilibrium code, PIES, using an external numerical wrapper. The PIES code (Princeton Iterative Equilibrium Solver) is capable of calculating 3D MHD equilibria with islands. The numerical wrapper has been demonstrated to greatly improve the rate of convergence in numerous cases corresponding to equilibria in the TFTR device where magnetic islands are present; the numerical wrapper makes use of a Jacobian-free Newton-Krylov solver along with adaptive preconditioning and a sophisticated subspace-restricted Levenberg backtracking algorithm. The wrapper has recently been improved by automation which combines the preexisting backtracking algorithm with insights gained from the stability of the Picard algorithm traditionally used with PIES. Improved progress logging and stopping criteria have also been incorporated in to the numerical wrapper.
Aerosol behaviour calculations with the code NAUA-Mod5M
International Nuclear Information System (INIS)
Bunz, H.; Koyro, M.
1995-03-01
This report presents the aerosol behaviour calculations within the framework of SEAFP task A8 'Radioactivity confinement analysis'. The retention capability for the aerosol-type activity of the containment has been evaluated for a number of different accident scenarios with the code NAUA-Mod5M. This code is designed to simulate the aerosol behaviour for an arbitrary multi-compartment containment originally for applications in LWR containments after severe accidents. Altogether six different scenarios have been evaluated, two for the He-cooled RPM and four for the watercooled APM. These scenarios differ mainly in the primary source taken into account, if e.g. the armour of the first wall consists of Be or W or if the divertor cooling loop or a primary cooling loop fails. The results show the positive influence of the system of step by step barriers already proved to be successful for other applications. (orig.) [de
A nonvariational code for calculating three-dimensional MHD [magnetohydrodynamic] equilibria
International Nuclear Information System (INIS)
Greenside, H.S.; Reiman, A.H.; Salas, A.
1987-09-01
Details are presented of the PIES code, which uses a nonvariational algorithm for calculating fully three-dimensional MHD equilibria. The MHD equilibrium equations are directly iterated in special coordinates to find self-consistent currents and magnetic fields for given pressure and current profiles and for a given outermost magnetic surface. Three important advantages of this approach over previous methods are the ease with which net current profiles can be imposed, the explicit treatment of resonances, and the ability to handle magnetic islands and stochastic field lines. The convergence properties of the code are studied for several axisymmetric and nonaxisymmetric finite-β equilibria that have magnetic surfaces. 36 refs., 14 figs., 3 tabs
ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®
Damian, F.; Brun, E.
2014-06-01
ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.
A code for the calculation of self-absorption fractions of photons
International Nuclear Information System (INIS)
Jaegers, P.; Landsberger, S.
1988-01-01
Neutron activation analysis (NAA) is now a well-established technique used by many researchers and commercial companies. It is often wrongly assumed that these NAA methods are matrix independent over a wide variety of samples. Accuracy at the level of a few percent is often difficult to achieve, since components such as timing, pulse pile-up, high dead-time corrections, sample positioning, and chemical separations may severely compromise the results. One area that has received little attention is the calculation of the effect of self-absorption of gamma-rays (including low-energy ones) in samples, particularly those with major components of high-Z values. The analysis of trace components in lead samples is an obvious example, but other high-Z matrices such as various permutations and combinations of zinc, tin, lead, copper, silver, antimony, etc.; ore concentrates; and meteorites are also affected. The authors have developed a simple but effective personal-computer-compatible user-friendly code, however, which can calculate the amount of energy signal that is lost due to the presence of any amount of one or more Z components. The program is based on Dixon's paper of 1951 for the calculation of self-absorption corrections for linear, cylindrical, and spherical sources. To determine the self-absorption fraction of a photon in a source, the FORTRAN computer code SELFABS was written
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)
Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code
Energy Technology Data Exchange (ETDEWEB)
Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)
2015-07-01
Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (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.
Calculation of nuclear data for incident energies to 200 MeV with the FKK-GNASH code system
International Nuclear Information System (INIS)
Chadwick, M.B.; Young, P.G.
1993-02-01
We describe how the FKK-GNASH code system has been extended to calculate nucleon-induced reactions up to 200 MeV, and used to predict (p,xn) and (p,xp) cross sections on 208 Pb at incident energies of 25, 45, 80 and 160 MeV, for an intermediate energy code intercomparison. Details of the reaction mechanisms calculated by FKK-GNASH are given, and the calculational procedure is described
Development of a nuclear spallation simulation code and calculations of primary spallation products
International Nuclear Information System (INIS)
Nishida, Takahiko; Nakahara, Yasuaki; Tsutsui, Tsuneo
1986-08-01
In order to make evaluations of computational models for the nuclear spallation reaction from a nuclear physics point of view, a simulation code NUCLEUS has been developed by modifying and combining the Monte Carlo codes NMTC/JAERI and NMTA/JAERI for calculating only the nuclear spallation reaction (intranuclear cascade + evaporation and/or fast fission) between a nucleus and a projectile without taking into consideration of internuclear transport. New several plotting routines have been provided for the rapid process of much more event data, obtained by using the ARGUS plotting system. The results obtained by our code can be directly compared with the experimental results using by thin foil experiments in which internuclear multiple collisions have little effects, and will serve to upgrade the calculational methods and the values of nuclear parameters currently used in the calculations. Some discussions are done about the preliminary computational results obtained by using NUCLEUS. The mass distribution and charge dispersion of reaction products are examined in some detail for the nuclear spallation reaction between incident protons and target nuclei, such as U, Pb and Ag, in the energy range from 0.5 GeV to 3.0 GeV. These results show that the distribution of reaction products ceases to change its form as the proton energy increases over about 2 GeV. The same tendency is seen in the energy dependence of the number of primary particles emitted from a nucleus. After spallation reactions, a variety of nuclei, especially many neutron deficient nuclides with nuclear charges nearly equal to ones of a target nucleus, are produced. Due to their short lifetime most of them will change to stable nuclides in due time. Finally, some important issues are discussed to improve the present simulation method. (author)
International Nuclear Information System (INIS)
Zachar, Matej; Necas, Vladimir; Daniska, Vladimir
2011-01-01
The activities performed during nuclear installation decommissioning process inevitably lead to the production of large amount of radioactive material to be managed. Significant part of materials has such low radioactivity level that allows them to be released to the environment without any restriction for further use. On the other hand, for materials with radioactivity slightly above the defined unconditional clearance level, there is a possibility to release them conditionally for a specific purpose in accordance with developed scenario assuring that radiation exposure limits for population not to be exceeded. The procedure of managing such decommissioning materials, mentioned above, could lead to recycling and reuse of more solid materials and to save the radioactive waste repository volume. In the paper an a implementation of the process of conditional release to the OMEGA Code is analyzed in details; the Code is used for calculation of decommissioning parameters. The analytical approach in the material parameters assessment, firstly, assumes a definition of radiological limit conditions, based on the evaluation of possible scenarios for conditionally released materials, and their application to appropriate sorter type in existing material and radioactivity flow system. Other calculation procedures with relevant technological or economical parameters, mathematically describing e.g. final radiation monitoring or transport outside the locality, are applied to the OMEGA Code in the next step. Together with limits, new procedures creating independent material stream allow evaluation of conditional material release process during decommissioning. Model calculations evaluating various scenarios with different input parameters and considering conditional release of materials to the environment are performed to verify the implemented methodology. Output parameters and results of the model assessment are presented, discussed and conduced in the final part of the paper
International Nuclear Information System (INIS)
Rojas C, E.L.; Varon T, C.F.; Pedraza N, R.
2007-01-01
The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)
International Nuclear Information System (INIS)
Cenerino, G.; Marbeuf, A.; Vahlas, C.
1992-01-01
Since 1974, Thermodata has been working on developing an Integrated Information System in Inorganic Chemistry. A major effort was carried on the thermochemical data assessment of both pure substances and multicomponent solution phases. The available data bases are connected to powerful calculation codes (GEMINI = Gibbs Energy Minimizer), which allow to determine the thermodynamical equilibrium state in multicomponent systems. The high interest of such an approach is illustrated by recent applications in as various fields as semi-conductors, chemical vapor deposition, hard alloys and nuclear safety. (author). 26 refs., 6 figs
Evaluation of Monte Carlo Codes Regarding the Calculated Detector Response Function in NDP Method
Energy Technology Data Exchange (ETDEWEB)
Tuan, Hoang Sy Minh; Sun, Gwang Min; Park, Byung Gun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
The basis of the NDP is the irradiation of a sample with a thermal or cold neutron beam and the subsequent release of charged particles due to neutron-induced exoergic charged particle reactions. Neutrons interact with the nuclei of elements and release mono-energetic charged particles, e.g. alpha particles or protons, and recoil atoms. Depth profile of the analyzed element can be obtained by making a linear transformation of the measured energy spectrum by using the stopping power of the sample material. A few micrometer of the material can be analyzed nondestructively, and on the order of 10nm depth resolution can be obtained depending on the material type with NDP method. In the NDP method, the one first steps of the analytical process is a channel-energy calibration. This calibration is normally made with the experimental measurement of NIST Standard Reference Material sample (SRM-93a). In this study, some Monte Carlo (MC) codes were tried to calculate the Si detector response function when this detector accounted the energy charges particles emitting from an analytical sample. In addition, these MC codes were also tried to calculate the depth distributions of some light elements ({sup 10}B, {sup 3}He, {sup 6}Li, etc.) in SRM-93a and SRM-2137 samples. These calculated profiles were compared with the experimental profiles and SIMS profiles. In this study, some popular MC neutron transport codes are tried and tested to calculate the detector response function in the NDP method. The simulations were modeled based on the real CN-NDP system which is a part of Cold Neutron Activation Station (CONAS) at HANARO (KAERI). The MC simulations are very successful at predicting the alpha peaks in the measured energy spectrum. The net area difference between the measured and predicted alpha peaks are less than 1%. A possible explanation might be bad cross section data set usage in the MC codes for the transport of low energetic lithium atoms inside the silicon substrate.
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)
Validation of VHTRC calculation benchmark of critical experiment using the MCB code
Directory of Open Access Journals (Sweden)
Stanisz Przemysław
2016-01-01
Full Text Available The calculation benchmark problem Very High Temperature Reactor Critical (VHTR a pin-in-block type core critical assembly has been investigated with the Monte Carlo Burnup (MCB code in order to validate the latest version of Nuclear Data Library based on ENDF format. Executed benchmark has been made on the basis of VHTR benchmark available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments. This benchmark is useful for verifying the discrepancies in keff values between various libraries and experimental values. This allows to improve accuracy of the neutron transport calculations that may help in designing the high performance commercial VHTRs. Almost all safety parameters depend on the accuracy of neutron transport calculation results that, in turn depend on the accuracy of nuclear data libraries. Thus, evaluation of the libraries applicability to VHTR modelling is one of the important subjects. We compared the numerical experiment results with experimental measurements using two versions of available nuclear data (ENDF-B-VII.1 and JEFF-3.2 prepared for required temperatures. Calculations have been performed with the MCB code which allows to obtain very precise representation of complex VHTR geometry, including the double heterogeneity of a fuel element. In this paper, together with impact of nuclear data, we discuss also the impact of different lattice modelling inside the fuel pins. The discrepancies of keff have been successfully observed and show good agreement with each other and with the experimental data within the 1 σ range of the experimental uncertainty. Because some propagated discrepancies observed, we proposed appropriate corrections in experimental constants which can improve the reactivity coefficient dependency. Obtained results confirm the accuracy of the new Nuclear Data Libraries.
International Nuclear Information System (INIS)
Ishiwatari, Nasumi
1978-11-01
The computer code ''FPRM-1'' has been developed for calculation of the quantities of fission products gases released from pellets into plenum in a fuel rod. On the assumption that the irradiation tests of plutonium fuel and others under development in an in-pile water loop were performed, FP generations and accumulations in the fuel rods were calculated by the code. The result of measurement of 131 I released from a fuel rod (UO 2 pellets, 235 U 1.5% Enriched) with an artificial hole through cladding in an in-pile water loop was compared with that of calculation by the code; both were in good agreement. (author)
STH-CFD Codes Coupled Calculations Applied to HLM Loop and Pool Systems
Directory of Open Access Journals (Sweden)
M. Angelucci
2017-01-01
Full Text Available This work describes the coupling methodology between a modified version of RELAP5/Mod3.3 and ANSYS Fluent CFD code developed at the University of Pisa. The described coupling procedure can be classified as “two-way,” nonoverlapping, “online” coupling. In this work, a semi-implicit numerical scheme has been implemented, giving greater stability to the simulations. A MATLAB script manages both the codes, oversees the reading and writing of the boundary conditions at the interfaces, and handles the exchange of data. A new tool was used to control the Fluent session, allowing a reduction of the time required for the exchange of data. The coupling tool was used to simulate a loop system (NACIE facility and a pool system (CIRCE facility, both working with Lead Bismuth Eutectic and located at ENEA Brasimone Research Centre. Some modifications in the coupling procedure turned out to be necessary to apply the methodology in the pool system. In this paper, the comparison between the obtained coupled numerical results and the experimental data is presented. The good agreement between experiments and calculations evinces the capability of the coupled calculation to model correctly the involved phenomena.
Decay heat experiment and validation of calculation code systems for fusion reactor
International Nuclear Information System (INIS)
Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki
1999-10-01
Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)
Calculation of the CAREM reactor with the HUEMUL-PUMA-THERMIT chain of codes
International Nuclear Information System (INIS)
Notari, Carla; Grant, Carlos R.
2000-01-01
The purpose of the work was the evaluation of the the CAREM 25 reactor core, using a chain of codes (HUEMUL-PUMA-THERMIT) different to the one used in the original design (CONDOR-CITVAP-THERMIT). First, we performed a partial validation of the our codes in lattices similar to CAREM and reproduced a benchmark for simulation of gadolinium burnup. The results were considered satisfactory for this stage of the project. Then, we calculated the core along the normal operating equilibrium cycle and in hot and cold shut-down conditions. The main outcome of our evaluation confirms the general behaviour of the reference calculations except in one important point referring to the cold shut down. In this condition, the failure of one single rod of bank number 13 of the shut down system, leaves the core in a supercritical state at the beginning of the cycle and this anomaly persists during almost a third of the overall cycle. A new design of the core is proposed with minor modifications of the reference one, without introducing new types of fuel elements and keeping the same fuel management scheme. This new core fulfills all the design requirements. (author)
Decay heat experiment and validation of calculation code systems for fusion reactor
Energy Technology Data Exchange (ETDEWEB)
Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki
1999-10-01
Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
Energy Technology Data Exchange (ETDEWEB)
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)
2014-08-15
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
Development and application of the PCRELAP5 - Data Calculation Program for RELAP 5 Code
Energy Technology Data Exchange (ETDEWEB)
Silvestre, Larissa J.B.; Sabundjian, Gaianê, E-mail: larissajbs@usp.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)
2017-07-01
Nuclear accidents in the world led to the establishment of rigorous criteria and requirements for nuclear power plant operations by the international regulatory bodies. By using specific computer programs, simulations of various accidents and transients likely to occur at any nuclear power plant are required for certifying and licensing a nuclear power plant. Some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most widely used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors in Brazil and worldwide. A major difficulty in the simulation by using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. Thus, for those calculations performance and preparation of RELAP5 input data, a friendly mathematical preprocessor was designed. The Visual Basic for Application (VBA) for Microsoft Excel demonstrated to be an effective tool to perform a number of tasks in the development of the program. In order to meet the needs of RELAP5 users, the RELAP5 Calculation Program (Programa de Cálculo do RELAP5 – PCRELAP5) was designed. The components of the code were codified; all entry cards including the optional cards of each one have been programmed. An English version for PCRELAP5 was provided. Furthermore, a friendly design was developed in order to minimize the time of preparation of input data and errors committed by users. The final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra-2. (author)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
International Nuclear Information System (INIS)
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H.
2014-08-01
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
International Nuclear Information System (INIS)
Bezrukov, Yu.A.; Shchekoldin, V.I.
2002-01-01
On the basis of the Gidropress OKB (Special Design Bureau) experimental data bank one verified the KORSAR code design models and correlations as to heat exchange crisis and overcrisis heat transfer as applied to the WWER reactor normal and emergency conditions. The VI.006.000 version of KORSAR code base calculations is shown to describe adequately the conducted experiments and to deviate insignificantly towards the conservative approach. So it may be considered as one of the codes ensuring more precise estimation [ru
Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C
International Nuclear Information System (INIS)
Suman, H.; Kharita, M. H.; Yousef, S.
2008-02-01
In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)
International Nuclear Information System (INIS)
Green, C.
1979-12-01
A set of FORTRAN subroutines is described for calculating water thermodynamic properties. These were written for use in a transient thermal-hydraulics program, where speed of execution is paramount. The choice of which subroutines to optimise depends on the primary variables in the thermal-hydraulics code. In this particular case the subroutine which has been optimised is the one which calculates pressure and specific enthalpy given the specific volume and the specific internal energy. Another two subroutines are described which complete a self-consistent set. These calculate the specific volume and the temperature given the pressure and the specific enthalpy, and the specific enthalpy and the specific volume given the pressure and the temperature (or the quality). The accuracy is high near the saturation lines, typically less than 1% relative error, and decreases as the fluid becomes more subcooled in the liquid region or more superheated in the steam region. This behaviour is inherent in the method which uses quantities defined on the saturation lines and assumes that certain derivatives are constant for excursions away from these saturation lines. The accuracy and speed of the subroutines are discussed in detail in this report. (author)
A code for calculating force and temperature of a bitter plate type toroidal field coil system
International Nuclear Information System (INIS)
Christensen, U.
1989-01-01
To assist the design effort of the TF coils for CIT, a set of programs was developed to calculate the transient spatial distribution of the current density, the temperature and the forces in the TF coil conductor region. The TF coils are of the Bitter (disk) type design and therefore have negligible variation of current density in the toroidal direction. During the TF pulse, voltages are induced which cause the field and current to diffuse in the minor radial direction. This penetration, combined with the increase of resistance due to the temperature rise determines the distribution of the current. After the current distribution has been determined, the in-plane (TF-TF) and the out-of-plane (TF-PF) forces in the conductor are computed. The predicted currents and temperatures have been independently corroborated using the SPARK code which has been modified for this type of problem. 6 figs
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
NXDC-neutron and x-ray diffraction code for crystal structures calculations
International Nuclear Information System (INIS)
Abbas, Y.; El-Sherif, A.
1982-01-01
A computer program NXDC for the calculations of neutron diffraction and x-ray diffraction intensities is reported. The program is very flexible and allows the intensity of a reflection with a given Miller indices to be calculated if the unit cell and its contents are specified together with the equipement used Neutrons or X-rays-and if necessary introducing temperature and absorption factors corrections. For the refinement of crystal structures provision is made for the comparison of the calculated intensities and the intergrated intensities observed from the diffraction diagrams using the least-squares analysis to obtain the reliability factor R. The program is written in FORTRAN Iv and is very suitable for minicomputers
BARS - a heterogeneous code for 3D pin-by-pin LWR steady-state and transient calculation
International Nuclear Information System (INIS)
Avvakumov, A.V.; Malofeev, V.M.
2000-01-01
A 3D pin-by-pin dynamic model for LWR detailed calculation was developed. The model is based on a coupling of the BARS neutronic code with the RELAP5/MOD3.2 thermal hydraulic code. This model is intended to calculate a fuel cycle, a xenon transient, and a wide range of reactivity initiated accidents in a WWER and a PWR. Galanin-Feinberg heterogeneous method was realized in the BARS code. Some results for a validation of the heterogeneous method are presented for reactivity coefficients, a pin-by-pin power distribution, and a fast pulse transient. (Authors)
Zhang, Bintuan; Dang, Bingrong; Wang, Zhuanzi; Wei, Wei; Li, Wenjian
2013-10-01
The skin tissue-equivalent slab reported in the International Commission on Radiological Protection (ICRP) Publication 116 to calculate the localised skin dose conversion coefficients (LSDCCs) was adopted into the Monte Carlo transport code Geant4. The Geant4 code was then utilised for computation of LSDCCs due to a circular parallel beam of monoenergetic electrons, protons and alpha particles electrons and alpha particles are found to be in good agreement with the results using the MCNPX code of ICRP 116 data. The present work thus validates the LSDCC values for both electrons and alpha particles using the Geant4 code.
International Nuclear Information System (INIS)
Utsumi, Takayuki; Sasaki, Akira
2000-02-01
The procedures of grasp92 code to calculate accurate (relative error nearly equal 10 -7 ) eigenvalue for the ground sate of helium atom of the Dirac-Coulomb Hamiltonian are presented. The grasp92 code, based on the multi-configuration Dirac-Fock method, is widely used to calculate the atomic properties. However, the main part of the accurate calculations, extended optimal level calculation (EOL), suffer frequently numerical instabilities due to the lack of the confident procedures. The purpose of this report is to illustrate the guideline for stable EOL calculations by calculating the most fundamental atomic system, i.e. the ground sate of helium atom ls 2 1 S 2 . This procedure could be extended for the high-precise eigenfunction calculation of more complex atomic systems, for example highly ionized atoms and high-Z atoms. (author)
International Nuclear Information System (INIS)
Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog
2005-03-01
Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too
Monte Carlo code Serpent calculation of the parameters of the stationary nuclear fission wave
Directory of Open Access Journals (Sweden)
V. M. Khotyayintsev
2017-12-01
Full Text Available n this work, propagation of the stationary nuclear fission wave was simulated for series of fixed power values using Monte Carlo code Serpent. The wave moved in the axial direction in 5 m long cylindrical core of fast reactor with pure 238U raw fuel. Stationary wave mode arises some period later after the wave ignition and lasts sufficiently long to determine kef with high enough accuracy. The velocity characteristic of the reactor was determined as the dependence of the wave velocity on the neutron multiplication factor. As we have recently shown within a one-group diffusion description, the velocity characteristic is two-valued due to the effect of concentration mechanisms, while thermal feedback affects it only quantitatively. The shape and parameters of the velocity characteristic critically affect feasibility of the reactor design since stationary wave solutions of the lower branch are unstable and do not correspond to any real waves in self-regulated reactor, like CANDLE. In this work calculations were performed without taking into account thermal feedback. They confirm that theoretical dependence correctly describes the shape of the velocity characteristic calculated using the results of the Serpent modeling.
Ion cyclotron emission calculations using a 2D full wave numerical code
International Nuclear Information System (INIS)
Batchelor, D.B.; Jaeger, E.F.; Colestock, P.L.
1987-01-01
Measurement of radiation in the HF band due to cyclotron emission by energetic ions produced by fusion reactions or neutral beam injection promises to be a useful diagnostic on large devices which are entering the reactor regime of operation. A number of complications make the modelling and interpretation of such measurements difficult using conventional geometrical optics methods. In particular the long wavelength and lack of high directivity of antennas in this frequency regime make observation of a single path across the plasma into a viewing dump impractical. Pickup antennas effectively see the whole plasma and wall reflection effects are important. We have modified our 2D full wave ICRH code 2 to calculate wave fields due to a distribution of energetic ions in tokamak geometry. The radiation is modeled as due to an ensemble of localized source currents distributed in space. The spatial structure of the coherent wave field is then calculated including cyclotron harmonic damping as compared to the usual procedure of incoherently summing powers of individual radiators. This method has the advantage that phase information from localized radiating currents is globally retained so the directivity of the pickup antennas is correctly represented. Also standing waves and wall reflections are automatically included
Energy Technology Data Exchange (ETDEWEB)
Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)
1997-12-31
Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.
International Nuclear Information System (INIS)
Marguet, S.D.
1997-01-01
Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF's neutronic code COCCINELLE uses the Rowland's formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission's products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on 'low' configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.)
Energy Technology Data Exchange (ETDEWEB)
Jones, H.D.
1976-06-01
The EXALPHA procedures provide a simplified method for running the MUSCATEL computer code, which in turn is used for calculating electronic properties of simple molecules and atomic clusters, based on the multiply scattered electron approximation for the wave equations. The use of the EXALPHA procedures to set up a run of MUSCATEL is described.
International Nuclear Information System (INIS)
Truchet, G.; Leconte, P.; Peneliau, Y.; Santamarina, A.
2013-01-01
The first goal of this paper is to present an exact method able to precisely evaluate very small reactivity effects with a Monte Carlo code (<10 pcm). it has been decided to implement the exact perturbation theory in TRIPOLI-4 and, consequently, to calculate a continuous-energy adjoint flux. The Iterated Fission Probability (IFP) method was chosen because it has shown great results in some other Monte Carlo codes. The IFP method uses a forward calculation to compute the adjoint flux, and consequently, it does not rely on complex code modifications but on the physical definition of the adjoint flux as a phase-space neutron importance. In the first part of this paper, the IFP method implemented in TRIPOLI-4 is described. To illustrate the efficiency of the method, several adjoint fluxes are calculated and compared with their equivalent obtained by the deterministic code APOLLO-2. The new implementation can calculate angular adjoint flux. In the second part, a procedure to carry out an exact perturbation calculation is described. A single cell benchmark has been used to test the accuracy of the method, compared with the 'direct' estimation of the perturbation. Once again the method based on the IFP shows good agreement for a calculation time far more inferior to the 'direct' method. The main advantage of the method is that the relative accuracy of the reactivity variation does not depend on the magnitude of the variation itself, which allows us to calculate very small reactivity perturbations with high precision. It offers the possibility to split reactivity contributions on both isotopes and reactions. Other applications of this perturbation method are presented and tested like the calculation of exact kinetic parameters (βeff, Λeff) or sensitivity parameters
Energy Technology Data Exchange (ETDEWEB)
Poletiko, C.; Hueber, C. [Inst. de Protection et de Surete Nucleaire, C.E. Cadarache, St. Paul-lez-Durance (France); Fabre, B. [CISI, C.E. Cadarache, St. Paul-lez-Durance (France)
1996-12-01
In case of severe nuclear accident, radioactive material may be released into the environment. Among the fission products involved, are the very volatile iodine isotopes. However, the chemical forms are not well known due to the presence of different species in the containment with which iodine may rapidly react to form aerosols, molecular iodine, hydroiodic acid and iodo-organics. Tentative explanations of different mechanisms were performed through benchscale tests. A series of tests has been performed at AEA Harwell (GB) to study parameters such as pH, dose rate, concentration, gas flow rate, temperature in relation to molecular iodine production, under dynamic conditions. Another set of tests has been performed in AECL Whiteshell (CA) to study the behaviour of painted coupons, standing in gas phase or liquid phase or both, with iodine compounds under radiation. The purpose of our paper is to synthesize the data and compare the results to the IODE code calculation. Some parameters of the code were studied to fit the experimental result the best. A law, concerning the reverse reaction of iodide radiolytic oxidation, has been proposed versus: pH, concentrations and gas flow-rate. This law does not apply for dose rate variations. For the study of painted coupons, it has been pointed out that molecular iodine tends to be adsorbed or chemically absorbed on the surface in gas phase, but the mechanism should be more sophisticated in the aqueous phase. The iodo-organics present in liquid phase tend to be partly or totally destroyed by oxidation under radiation (depending upon the dose delivered). These points are discussed. (author) 18 figs., 3 tabs., 15 refs.
International Nuclear Information System (INIS)
Faya, A.G.; Nakata, H.; Rodrigues, V.G.; Oosterkamp, W.J.
1974-01-01
The main variables for Obrigheim Reactor - KWO diffusion theory calculations, using the EREBUS code were defined. The variables under consideration were: mesh spacing for reactor description, time-step in burn-up calculation, and the temperature in both the moderator and the fuel. The best mesh spacing and time-step were defined considering the relative deviations and the computer time expended in each case. It has been verified that the error involved in the mean fuel temperature calculation (1317 0 K as given by SIEMENS and 1028 0 K as calculated by Dr. Penndorf) does not change substancially the calculation results
Comparison of Space Radiation Calculations from Deterministic and Monte Carlo Transport Codes
Adams, J. H.; Lin, Z. W.; Nasser, A. F.; Randeniya, S.; Tripathi, r. K.; Watts, J. W.; Yepes, P.
2010-01-01
The presentation outline includes motivation, radiation transport codes being considered, space radiation cases being considered, results for slab geometry, results from spherical geometry, and summary. ///////// main physics in radiation transport codes hzetrn uprop fluka geant4, slab geometry, spe, gcr,
European coding system for tissues and cells: a challenge unmet?
Reynolds, Melvin; Warwick, Ruth M; Poniatowski, Stefan; Trias, Esteve
2010-11-01
The Comité Européen de Normalisation (European Committee for Standardization, CEN) Workshop on Coding of Information and Traceability of Human Tissues and Cells was established by the Expert Working Group of the Directorate General for Health and Consumer Affairs of the European Commission (DG SANCO) to identify requirements concerning the coding of information and the traceability of human tissues and cells, and propose guidelines and recommendations to permit the implementation of the European Coding system required by the European Tissues and Cells Directive 2004/23/EC (ED). The Workshop included over 70 voluntary participants from tissue, blood and eye banks, national ministries for healthcare, transplant organisations, universities and coding organisations; mainly from Europe with a small number of representatives from professionals in Canada, Australia, USA and Japan. The Workshop commenced in April 2007 and held its final meeting in February 2008. The draft Workshop Agreement went through a public comment phase from 15 December 2007 until 15 January 2008 and the endorsement period ran from 9 April 2008 until 2 May 2008. The endorsed CEN Workshop Agreement (CWA) set out the issues regarding a common coding system, qualitatively assessed what the industry felt was required of a coding system, reviewed coding systems that were put forward as potential European coding systems and established a basic specification for a proposed European coding system for human tissues and cells, based on ISBT 128, and which is compatible with existing systems of donation identification, traceability and nomenclatures, indicating how implementation of that system could be approached. The CWA, and the associated Workshop proposals with recommendations, were finally submitted to the European Commission and to the Committee of Member States that assists its management process under article 29 of the Directive 2004/23/EC on May 25 2008. In 2009 the European Commission initiated an
Implementation of an implicit method into heat conduction calculation of TRAC-PF1/MOD2 code
International Nuclear Information System (INIS)
Akimoto, Hajime; Abe, Yutaka; Ohnuki, Akira; Murao, Yoshio
1990-08-01
A two-dimensional unsteady heat conduction equation is solved in the TRAC-PF/MOD2 code to calculate temperature transients in fuel rod. A large CPU time is often required to get stable solution of temperature transients in the TRAC calculation with a small axial node size (less than 1.0 mm), because the heat conduction equation is discretized explicitly. To eliminate the restriction of the maximum time step size by the heat conduction calculation, an implicit method for solving the heat condition equation was developed and implemented into the TRAC code. Several assessment calculations were performed with the original and modified TRAC codes. It is confirmed that the implicit method is reliable and is successfully implemented into the TRAC code through comparison with theoretical solutions and assessment calculation results. It is demonstrated that the implicit method makes the heat conduction calculation practical even for the analyses of temperature transients with the axial node size less than 0.1 mm. (author)
Calculation of Sodium Fire Test-I (Run-E6) using sodium combustion analysis code ASSCOPS version 2.0
Energy Technology Data Exchange (ETDEWEB)
Nakagiri, Toshio; Ohno, Shuji; Miyake, Osamu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1997-11-01
The calculation of Sodium Fire Test-I (Run-E6) was performed using the ASSCOPS (Analysis of Simultaneous Sodium Combustions in Pool and Spray) code version 2.0 in order to determine the parameters used in the code for the calculations of sodium combustion behavior of small or medium scale sodium leak, and to validate the applicability of the code. The parameters used in the code were determined and the validation of the code was confirmed because calculated temperatures, calculated oxygen concentration and other calculated values almost agreed with the test results. (author)
SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR
International Nuclear Information System (INIS)
Halsall, M.J.; Course, A.F.; Sidell, J.
1979-09-01
SPLOSH is a time dependent, one dimensional, finite difference (in time and space) coupled neutron kinetics and thermal hydraulics code for studying pressurised faults and control transients in water reactor systems. An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the reactor core. A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface. The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly. Circuit features modelled include pumps, spray cooling and the SGHWR steam drum. Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified. (UK)
Base data for looking-up tables of calculation errors in JACS code system
International Nuclear Information System (INIS)
Murazaki, Minoru; Okuno, Hiroshi
1999-03-01
The report intends to clarify the base data for the looking-up tables of calculation errors cited in 'Nuclear Criticality Safety Handbook'. The tables were obtained by classifying the benchmarks made by JACS code system, and there are two kinds: One kind is for fuel systems in general geometry with a reflected and another kind is for fuel systems specific to simple geometry with a reflector. Benchmark systems were further categorized into eight groups according to the fuel configuration: homogeneous or heterogeneous; and fuel kind: uranium, plutonium and their mixtures, etc. The base data for fuel systems in general geometry with a reflected are summarized in this report for the first time. The base data for fuel systems in simple geometry with a reflector were summarized in a technical report published in 1987. However, the data in a group named homogeneous low-enriched uranium were further selected out later by the working group for making the Nuclear Criticality Safety Handbook. This report includes the selection. As a project has been organized by OECD/NEA for evaluation of criticality safety benchmark experiments, the results are also described. (author)
International Nuclear Information System (INIS)
Scheglov, A.S.; Proselkov, V.N.; Sidorenko, V.D.; Passage, G.; Stefanova, S.; Haralampieva, Tz.; Peychinov, Tz.
2000-01-01
A short description of the TOPRA-s computer code is presented. The code is developed to calculate the thermophysical cross-section characteristics of the WWER fuel rods: fuel temperature distributions and fuel-to-cladding gap conductance. The TOPRA-s input does not require the fuel rod irradiation pre-history (time dependent distributions of linear power, fast neutron flux and coolant temperature along the rod). The required input consists of the considered cross-section data (coolant temperature, burnup, linear power) and the overall fuel rod data (burnup and linear power). TOPRA-s is included into the KASKAD code package. Some results of the TOPRA-s code validation using the SOFIT-1 and IFA-503.1 experimental data, are shown. A short description of the TRANSURANUS code for thermal and mechanical predictions of the LWR fuel rod behavior at various irradiation conditions and its version for WWER reactors, are presented. (Authors)
Energy Technology Data Exchange (ETDEWEB)
Sugino, Kazuteru [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-07-01
As a tool to perform a fast reactor core calculations with high accuracy, NSHEX the nodal transport calculation code for three-dimensional hexagonal-Z geometry is under development. To improve the practical applicability of NSHEX, for instance, in its application to safety analysis and commercial reactor core design studies, we investigated the basic theory used in it, improved the program performance, and evaluated its applicability to the analysis of commercial reactor cores. The current studies show the following: (1) An improvement in the treatment of radial leakage in the radial nodal coupling equation bettered calculational convergence for safety analysis calculation, so the applicability of NSHEX to safety analysis was improved. (2) As a result of comparison of results from NSHEX and the standard core calculation code, it was confirmed that there was consistency between them. (3) According to the evaluation of the effect due to the difference of calculational condition, it was found that the calculation under appropriate nodal expansion orders and Sn orders correspond to the one under most detailed condition. However further investigation is required to reduce the uncertainty in calculational results due to the treatment of high order flux moments. (4) A whole core version of NSHEX enabling calculation for any FBR core geometry has been developed, this improved general applicability of NSHEX. (5) An investigation of the applicability of the rebalance method to acceleration clarified that this improved calculational convergence and it was effective. (J.P.N.)
International Nuclear Information System (INIS)
Hordosy, G.
2006-01-01
KARATE is a code system developed in KFKI AERI. It is routinely used for core calculation. Its depletion module are now tested against the radiochemical measurements of spent fuel samples from the Novovoronezh Unit IV, performed in RIAR, Dimitrovgrad. Due to the insufficient knowledge of operational history of the unit, the irradiation history of the samples was taken from formerly published Russian calculations. The calculation of isotopic composition was performed by the MULTICEL module of program system. The agreement between the calculated and measured values of the concentration of the most important actinides and fission products is investigated (Authors)
Code for calculation of spreading of radioactivity in reactor containment systems
International Nuclear Information System (INIS)
Vertes, P.
1992-09-01
A detailed description of the new version of TIBSO code is given, with applications for accident analysis in a reactor containment system. The TIBSO code can follow the nuclear transition and the spatial migration of radioactive materials. The modelling of such processes is established in a very flexible way enabling the user to investigate a wide range of problems. The TIBSO code system is described in detail, taking into account the new developments since 1983. Most changes improve the capabilities of the code. The new version of TIBSO system is written in FORTRAN-77 and can be operated both under VAX VMS and PC DOS. (author) 5 refs.; 3 figs.; 21 tabs
International Nuclear Information System (INIS)
Cintra, Felipe Belonsi de
2010-01-01
This study made a comparison between some of the major transport codes that employ the Monte Carlo stochastic approach in dosimetric calculations in nuclear medicine. We analyzed in detail the various physical and numerical models used by MCNP5 code in relation with codes like EGS and Penelope. The identification of its potential and limitations for solving microdosimetry problems were highlighted. The condensed history methodology used by MCNP resulted in lower values for energy deposition calculation. This showed a known feature of the condensed stories: its underestimates both the number of collisions along the trajectory of the electron and the number of secondary particles created. The use of transport codes like MCNP and Penelope for micrometer scales received special attention in this work. Class I and class II codes were studied and their main resources were exploited in order to transport electrons, which have particular importance in dosimetry. It is expected that the evaluation of available methodologies mentioned here contribute to a better understanding of the behavior of these codes, especially for this class of problems, common in microdosimetry. (author)
International Nuclear Information System (INIS)
Simon-Cornu, Marie; Mourlon, Christophe; Bordy, J.M.; Daures, J.; Dusiac, D.; Moignau, F.; Gouriou, J.; Million, M.; Moreno, B.; Chabert, I.; Lazaro, D.; Barat, E.; Dautremer, T.; Montagu, T.; Agelou, M.; De Carlan, L.; Patin, D.; Le Loirec, C.; Dupuis, P.; Gassa, F.; Guerin, L.; Batalla, A.; Leni, Pierre-Emmanuel; Laurent, Remy; Gschwind, Regine; Makovicka, Libor; Henriet, Julien; Salomon, Michel; Vivier, Alain; Lopez, Gerald; Dossat, C.; Pourrouquet, P.; Thomas, J.C.; Sarie, I.; Peyrard, P.F.; Chatry, N.; Lavielle, D.; Loze, R.; Brun, E.; Damian, F.; Diop, C.; Dumonteil, E.; Hugot, F.X.; Jouanne, C.; Lee, Y.K.; Malvagi, F.; Mazzolo, A.; Petit, O.; Trama, J.C.; Visonneau, T.; Zoia, A.; Courageot, Estelle; Gaillard-Lecanu, Emmanuelle; Kutschera, Reinald; Le Meur, Gaelle; Uzio, Fabien; De Conto, Celine; Gschwind, Regine; Makovicka, Libor; Farah, Jad; Martinetti, Florent; Sayah, Rima; Donadille, Laurent; Herault, Joel; Delacroix, Sabine; Nauraye, Catherine; Lee, Choonsik; Bolch, Wesley; Clairand, Isabelle; Horodynski, Jean-Michel; Pauwels, Nicolas; Robert, Pierre; VOLLAIRE, Joachim; Nicoletti, C.; Kitsos, S.; Tardy, M.; Marchaud, G.; Stankovskiy, Alexey; Van Den Eynde, Gert; Fiorito, Luca; Malambu, Edouard; Dreuil, Serge; Mougeot, X.; Be, M.M.; Bisch, C.; Villagrasa, C.; Dos Santos, M.; Clairand, I.; Karamitros, M.; Incerti, S.; Petitguillaume, Alice; Franck, Didier; Desbree, Aurelie; Bernardini, Michela; Labriolle-Vaylet, Claire de; Gnesin, Silvano; Leadermann, Jean-Pascal; Paterne, Loic; Bochud, Francois O.; Verdun, Francis R.; Baechler, Sebastien; Prior, John O.; Thomassin, Alain; Arial, Emmanuelle; Laget, Michael; Masse, Veronique; Saldarriaga Vargas, Clarita; Struelens, Lara; Vanhavere, Filip; Perier, Aurelien; Courageot, Estelle; Gaillard-Lecanu, Emmanuelle; Le-Meur, Gaelle; Monier, Catherine; Thers, Dominique; Le-Guen, Bernard; Blond, Serge; Cordier, Gerard; Le Roy, Maiwenn; De Carlan, Loic; Bordy, Jean-Marc; Caccia, Barbara; Andenna, Claudio; Charimadurai, Arun; Selvam, T Palani; Czarnecki, Damian; Zink, Klemens; Gschwind, Regine; Martin, Eric; Huot, Nicolas; Zoubair, Mariam; El Bardouni, Tarek; Lazaro, Delphine; Barat, Eric; Dautremer, Thomas; Montagu, Thierry; Chabert, Isabelle; Guerin, Lucie; Batalla, Alain; Moignier, C.; Huet, C.; Bassinet, C.; Baumann, M.; Barraux, V.; Sebe-Mercier, K.; Loiseau, C.; Batalla, A.; Makovicka, L.; Desnoyers, Yvon; Juhel, Gabriel; Mattera, Christophe; Tempier, Maryline
2014-03-01
These scientific days were organised by the 'technical protection' Section of the French Society of Radiation Protection (SFRP) in cooperation with the French society of medical physicists (SFPM), the Swiss Romandie association of radioprotection (ARRAD) and the associated laboratories of radio-physics and dosimetry (LARD). The objective of these days was to review the existing calculation codes used in radiation transport, source estimation and dose management, and to identify some future prospects. This document brings together the available presentations (slides) together with their corresponding abstracts (in French) and dealing with: 1 - Presentation of the conference days (L. De Carlan); 2 - Simulating radionuclide transfers in the environment: what calculation codes and for what? (C. Mourlon); 3 - Contribution of Monte-Carlo calculation to the theoretical foundation analysis of calibration procedures and dosemeters design for radioprotection photon dosimetry (J.M. Bordy); 4 - Use of calculation codes in R and D for the development of a new passive dosemeter for photons and beta radiations (B. Moreno); 5 - Development of a new virtual sources model for the Monte-Carlo prediction of EPID (Electronic Portal Imaging Device) images and implementation in PENELOPE (I. Chabert); 6 - Prediction of high-resolution EPID images for in-vivo dosimetry (D. Patin); 7 - 4D thorax modeling by artificial neural networks (P.E. Leni); 8 - Presentation of the calculation utilities of the book 'Calculation of ionizing radiations generated doses' (Vivier, Lopez, EDP Sciences 2012) (A. Vivier); 9 - RayXpert C : a 3D modeling and Monte-Carlo dose rate calculation software (C. Dossat); 10 - TRIPOLI-4 R Version 9 S Monte-Carlo code for radioprotection (F. Damian); 11 - Realistic radioprotection training with the digital school workshop (E. Courageot); 12 - Use of BEAMNRC code for dental prostheses influence evaluation in ENT cancers treatment by external radiotherapy (C. De Conto); 13
Performance calculation of the Compact Disc error correcting code ona memoryless channel
Driessen, L.H.M.E.; Vries, L.B.
2015-01-01
Recently N.V. PHILIPS of The Netherlands and SONY CORP. of Japan made a joint pnoposal for standardization of their COMPACT DISC DigitalAudio system. This standard as agreed upon, includes the choice ofan error correcting code called CIRC (Cross Interleave Reed SolomonCode),according to which the
International Nuclear Information System (INIS)
Beckert, C.
2007-01-01
Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed [de
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
International Nuclear Information System (INIS)
Mota, F.; Ortiz, C. J.; Vila, R.
2012-01-01
Irradiation Experimental Area of TechnoFusion will emulate the extreme irradiation fusion conditions in materials by means of three ion accelerators: one used for self-implanting heavy ions (Fe, Si, C,...) to emulate the displacement damage induced by fusion neutrons and the other two for light ions (H and He) to emulate the transmutation induced by fusion neutrons. This Laboratory will play an essential role in the selection of functional materials for DEMO reactor since it will allow reproducing the effects of neutron radiation on fusion materials. Ion irradiation produces little or no residual radioactivity, allowing handling of samples without the need for special precautions. Currently, two different methods are used to calculate the primary displacement damage by neutron irradiation or by ion irradiation. On one hand, the displacement damage doses induced by neutrons are calculated considering the NRT model based on the electronic screening theory of Linhard. This methodology is commonly used since 1975. On the other hand, for experimental research community the SRIM code is commonly used to calculate the primary displacement damage dose induced by ion irradiation. Therefore, both methodologies of primary displacement damage calculation have nothing in common. However, if we want to design ion irradiation experiments capable to emulate the neutron fusion effect in materials, it is necessary to develop comparable methodologies of damage calculation for both kinds of radiation. It would allow us to define better the ion irradiation parameters (Ion, current, Ion energy, dose, etc) required to emulate a specific neutron irradiation environment. Therefore, our main objective was to find the way to calculate the primary displacement damage induced by neutron irradiation and by ion irradiation starting from the same point, that is, the PKA spectrum. In order to emulate the neutron irradiation that would prevail under fusion conditions, two approaches are contemplated: a) on
Computer code TOBUNRAD for PWR fuel bundle heat-up calculations
International Nuclear Information System (INIS)
Shimooke, Takanori; Yoshida, Kazuo
1979-05-01
The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)
International Nuclear Information System (INIS)
Lee, Kyung-Hoon; Kim, Kang-Seog; Cho, Jin-Young; Song, Jae-Seung; Noh, Jae-Man; Lee, Chung-Chan
2008-01-01
The IAEA's gas-cooled reactor program has coordinated international cooperation for an evaluation of a high temperature gas-cooled reactor's performance, which includes a validation of the physics analysis codes and the performance models for the proposed GT-MHR. This benchmark problem consists of the pin and block calculations and the reactor physics of the control rod worth for the GT-MHR with a weapon grade plutonium fuel. Benchmark analysis has been performed by using the HELIOS/MASTER deterministic code package and the MCNP Monte Carlo code. The deterministic code package adopts a conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation. In order to solve particular modeling issues in GT-MHR, recently developed technologies were utilized and new analysis procedure was devised. Double heterogeneity effect could be covered by using the reactivity-equivalent physical transformation (RPT) method. Strong core-reflector interaction could be resolved by applying an equivalence theory to the generation of the reflector cross sections. In order to accurately handle with very large control rods which are asymmetrically located in a fuel and a reflector block, the surface dependent discontinuity factors (SDFs) were considered in applying an equivalence theory. A new method has been devised to consider SDFs without any modification of the nodal solver in MASTER. All computational results of the HELIOS/MASTER code package were compared with those of MCNP. The multiplication factors of HELIOS for the pin cells are in very good agreement with those of MCNP to within a maximum error of 693 pcm Δρ. The maximum differences of the multiplication factors for the fuel blocks are about 457 pcm Δρ and the control rod worths of HELIOS are consistent with those of MCNP to within a maximum error of 3.09%. On considering a SDF in the core
International Nuclear Information System (INIS)
Khelifi, R.; Idiri, Z.; Bode, P.
2002-01-01
The CITATION code based on neutron diffusion theory was used for flux calculations inside voluminous samples in prompt gamma activation analysis with an isotopic neutron source (Am-Be). The code uses specific parameters related to the energy spectrum source and irradiation system materials (shielding, reflector). The flux distribution (thermal and fast) was calculated in the three-dimensional geometry for the system: air, polyethylene and water cuboidal sample (50x50x50 cm). Thermal flux was calculated in a series of points inside the sample. The results agreed reasonably well with observed values. The maximum thermal flux was observed at a distance of 3.2 cm while CITATION gave 3.7 cm. Beyond a depth of 7.2 cm, the thermal flux to fast flux ratio increases up to twice and allows us to optimise the detection system position in the scope of in-situ PGAA
Impact of thermoplastic mask on X-ray surface dose calculated with Monte Carlo code
International Nuclear Information System (INIS)
Zhao Yanqun; Li Jie; Wu Liping; Wang Pei; Lang Jinyi; Wu Dake; Xiao Mingyong
2010-01-01
Objective: To calculate the effects of thermoplastic mask on X-ray surface dose. Methods: The BEAMnrc Monte Carlo Code system, designed especially for computer simulation of radioactive sources, was performed to evaluate the effects of thermoplastic mask on X-ray surface dose.Thermoplastic mask came from our center with a material density of 1.12 g/cm 2 . The masks without holes, with holes size of 0.1 cm x 0.1 cm, and with holes size of 0. 1 cm x 0.2 cm, and masks with different depth (0.12 cm and 0.24 cm) were evaluated separately. For those with holes, the material width between adjacent holes was 0.1 cm. Virtual masks with a material density of 1.38 g/cm 3 without holes with two different depths were also evaluated. Results: Thermoplastic mask affected X-rays surface dose. When using a thermoplastic mask with the depth of 0.24 cm without holes, the surface dose was 74. 9% and 57.0% for those with the density of 1.38 g/cm 3 and 1.12 g/cm 3 respectively. When focusing on the masks with the density of 1.12 g/cm 3 , the surface dose was 41.2% for those with 0.12 cm depth without holes; 57.0% for those with 0. 24 cm depth without holes; 44.5% for those with 0.24 cm depth with holes size of 0.1 cm x 0.2 cm;and 54.1% for those with 0.24 cm depths with holes size of 0.1 cm x 0.1 cm.Conclusions: Using thermoplastic mask during the radiation increases patient surface dose. The severity is relative to the hole size and the depth of thermoplastic mask. The surface dose change should be considered in radiation planning to avoid severe skin reaction. (authors)
A thermo mechanical benchmark calculation of a hexagonal can in the BTI accident with INCA code
International Nuclear Information System (INIS)
Zucchini, A.
1988-01-01
The thermomechanical behaviour of an hexagonal can in a benchmark problem (simulating the conditions of a BTI accident in a fuel assembly) is examined by means of the INCA code and the results systematically compared with those of ADINA
Calculation of behaviour of the Juragua NPP containment with code TRACOV/MOD1
International Nuclear Information System (INIS)
Castillo Alvarez, J.; Valle Cepero, R.; Luis, J.; San Roman, J.C.; Pomier, L.
1996-01-01
The containment of Juragua NPP has some unique features, which differ from the rest of the PWR reactors design. Those features impose additional requirements for its numerical simulation. In this paper is analyzed the behaviour of the Juragua NPP containment during accident situation with double ended break of the primary pipelines with flow in both direction using the code TRACOV/MOD1. The results are compared with obtained by the designer. The main restrictions of the code are identified
Non-coding RNAs Functioning in Colorectal Cancer Stem Cells.
Fanale, Daniele; Barraco, Nadia; Listì, Angela; Bazan, Viviana; Russo, Antonio
In recent years, the hypothesis of the presence of tumor-initiating cancer stem cells (CSCs) has received a considerable support. This model suggested the existence of CSCs which, thanks to their self-renewal properties, are able to drive the expansion and the maintenance of malignant cell populations with invasive and metastatic potential in cancer. Increasing evidence showed the ability of such cells to acquire self-renewal, multipotency, angiogenic potential, immune evasion, symmetrical and asymmetrical divisions which, along with the presence of several DNA repair mechanisms, further enhance their oncogenic potential making them highly resistant to common anticancer treatments. The main signaling pathways involved in the homeostasis of colorectal (CRC) stem cells are the Wnt, Notch, Sonic Hedgehog, and Bone Morfogenic Protein (BMP) pathways, which are mostly responsible for all the features that have been widely referred to stem cells. The same pathways have been identified in colorectal cancer stem cells (CRCSCs), conferring a more aggressive phenotype compared to non-stem CRC cells. Recently, several evidences suggested that non-coding RNAs (ncRNAs) may play a crucial role in the regulation of different biological mechanisms in CRC, by modulating the expression of critical stem cell transcription factors that have been found active in CSCs. In this chapter, we will discuss the involvement of ncRNAs, especially microRNAs (miRNAs) and long non-coding RNAs (lncRNAs), in stemness acquisition and maintenance by CRCSCs, through the regulation of pathways modulating the CSC phenotype and growth, carcinogenesis, differentiation, and epithelial to mesenchymal transition (EMT).
The MARS15-based FermiCORD code system for calculation of the accelerator-induced residual dose
Grebe, A.; Leveling, A.; Lu, T.; Mokhov, N.; Pronskikh, V.
2018-01-01
The FermiCORD code system, a set of codes based on MARS15 that calculates the accelerator-induced residual doses at experimental facilities of arbitrary configurations, has been developed. FermiCORD is written in C++ as an add-on to Fortran-based MARS15. The FermiCORD algorithm consists of two stages: 1) simulation of residual doses on contact with the surfaces surrounding the studied location and of radionuclide inventories in the structures surrounding those locations using MARS15, and 2) simulation of the emission of the nuclear decay γ-quanta by the residuals in the activated structures and scoring the prompt doses of these γ-quanta at arbitrary distances from those structures. The FermiCORD code system has been benchmarked against similar algorithms based on other code systems and against experimental data from the CERF facility at CERN, and FermiCORD showed reasonable agreement with these. The code system has been applied for calculation of the residual dose of the target station for the Mu2e experiment and the results have been compared to approximate dosimetric approaches.
The MARS15-based FermiCORD code system for calculation of the accelerator-induced residual dose
Energy Technology Data Exchange (ETDEWEB)
Grebe, A.; Leveling, A.; Lu, T.; Mokhov, N.; Pronskikh, V.
2018-01-01
The FermiCORD code system, a set of codes based on MARS15 that calculates the accelerator-induced residual doses at experimental facilities of arbitrary configurations, has been developed. FermiCORD is written in C++ as an add-on to Fortran-based MARS15. The FermiCORD algorithm consists of two stages: 1) simulation of residual doses on contact with the surfaces surrounding the studied location and of radionuclide inventories in the structures surrounding those locations using MARS15, and 2) simulation of the emission of the nuclear decay gamma-quanta by the residuals in the activated structures and scoring the prompt doses of these gamma-quanta at arbitrary distances from those structures. The FermiCORD code system has been benchmarked against similar algorithms based on other code systems and showed a good agreement. The code system has been applied for calculation of the residual dose of the target station for the Mu2e experiment and the results have been compared to approximate dosimetric approaches.
International Nuclear Information System (INIS)
Androsenko, A.A.; Androsenko, P.A.; Kagalenko, I.Eh.; Mironovich, Yu.N.
1992-01-01
Consideration is given of a technique and algorithms of constructing neutron trajectories in the Monte-Carlo method taking into account the data on adjoint transport equation solution. When simulating the transport part of transfer kernel the use is made of piecewise-linear approximation of free path length density along the particle motion direction. The approach has been implemented in programs within the framework of the BRAND code system. The importance is calculated in the multigroup P 1 -approximation within the framework of the DD-30 code system. The efficiency of the developed computation technique is demonstrated by means of solution of two model problems. 4 refs.; 2 tabs
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
Load-balancing techniques for a parallel electromagnetic particle-in-cell code
International Nuclear Information System (INIS)
Plimpton, Steven J.; Seidel, David B.; Pasik, Michael F.; Coats, Rebecca S.
2000-01-01
QUICKSILVER is a 3-d electromagnetic particle-in-cell simulation code developed and used at Sandia to model relativistic charged particle transport. It models the time-response of electromagnetic fields and low-density-plasmas in a self-consistent manner: the fields push the plasma particles and the plasma current modifies the fields. Through an LDRD project a new parallel version of QUICKSILVER was created to enable large-scale plasma simulations to be run on massively-parallel distributed-memory supercomputers with thousands of processors, such as the Intel Tflops and DEC CPlant machines at Sandia. The new parallel code implements nearly all the features of the original serial QUICKSILVER and can be run on any platform which supports the message-passing interface (MPI) standard as well as on single-processor workstations. This report describes basic strategies useful for parallelizing and load-balancing particle-in-cell codes, outlines the parallel algorithms used in this implementation, and provides a summary of the modifications made to QUICKSILVER. It also highlights a series of benchmark simulations which have been run with the new code that illustrate its performance and parallel efficiency. These calculations have up to a billion grid cells and particles and were run on thousands of processors. This report also serves as a user manual for people wishing to run parallel QUICKSILVER
Load-balancing techniques for a parallel electromagnetic particle-in-cell code
Energy Technology Data Exchange (ETDEWEB)
PLIMPTON,STEVEN J.; SEIDEL,DAVID B.; PASIK,MICHAEL F.; COATS,REBECCA S.
2000-01-01
QUICKSILVER is a 3-d electromagnetic particle-in-cell simulation code developed and used at Sandia to model relativistic charged particle transport. It models the time-response of electromagnetic fields and low-density-plasmas in a self-consistent manner: the fields push the plasma particles and the plasma current modifies the fields. Through an LDRD project a new parallel version of QUICKSILVER was created to enable large-scale plasma simulations to be run on massively-parallel distributed-memory supercomputers with thousands of processors, such as the Intel Tflops and DEC CPlant machines at Sandia. The new parallel code implements nearly all the features of the original serial QUICKSILVER and can be run on any platform which supports the message-passing interface (MPI) standard as well as on single-processor workstations. This report describes basic strategies useful for parallelizing and load-balancing particle-in-cell codes, outlines the parallel algorithms used in this implementation, and provides a summary of the modifications made to QUICKSILVER. It also highlights a series of benchmark simulations which have been run with the new code that illustrate its performance and parallel efficiency. These calculations have up to a billion grid cells and particles and were run on thousands of processors. This report also serves as a user manual for people wishing to run parallel QUICKSILVER.
Calculs de doses générées par les rayonnements ionisants principes physiques et codes de calcul
Vivier, Alain
2016-01-01
Cet ouvrage et les codes associés s’adressent aux utilisateurs de sources de rayonnements ionisants : techniciens, ingénieurs de sécurité, personnes compétentes en radioprotection, mais aussi médecins, chercheurs, concepteurs, décideurs… Les contraintes croissantes liées à la radioprotection rendent indispensables l’utilisation de codes de calcul permettant d’évaluer les débits de doses générées par ces sources et la façon dont on peut s’en protéger au mieux. De nombreux codes existent, dont certains restent des références incontournables, mais ils sont relativement complexes à mettre en oeuvre et restent en général réservés aux bureaux d’études. En outre, ces codes sont souvent des « boîtes noires » qui ne permettent pas de comprendre la physique sous-jacente. L’objectif de cet ouvrage est double : - Exposer les principes physiques permettant de comprendre les phénomènes à l’oeuvre lorsque la matière est irradiée par des rayonnements ionisants. Il devient al...
Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors
International Nuclear Information System (INIS)
Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C.
2006-10-01
In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions
Energy Technology Data Exchange (ETDEWEB)
Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)
1994-04-01
Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions.
International Nuclear Information System (INIS)
Neymotin, L.
1994-04-01
Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions
International Nuclear Information System (INIS)
Reshak, Ali H.; Jamal, Morteza
2012-01-01
Highlights: ► A new package for calculating elastic constants of orthorhombic structure is released. ► The package called ortho-elastic. ► It is compatible with [FP-(L)APW+lo] method implemented in WIEN2k code. ► Several orthorhombic structure compounds were used to test the new package. ► Elastic constants calculated using this package show good agreement with experiment. - Abstract: A new package for calculating the elastic constants of orthorhombic structure is released. The package called ortho-elastic. The formalism of calculating the ortho-elastic constants is described in details. The package is compatible with the highly accurate all-electron full-potential (linearized) augmented plane-wave plus local orbital [FP-(L)APW+lo] method implemented in WIEN2k code. Several orthorhombic structure compounds were used to test the new package. We found that the calculated elastic constants using the new package show better agreement with the available experimental data than the previous theoretical results used different methods. In this package the second-order derivative E ″ (ε) of polynomial fit E=E(ε) of energy vs strains at zero strain (ε=0), used to calculate the orthorhombic elastic constants.
Application of the TWODANT code system to pressure vessel dosimetry calculations
International Nuclear Information System (INIS)
Parsons, D.K.; Alcouffe, R.E.; Marr, D.R.; Urban, W.T.
1993-01-01
The TWODANT code system has recently been enhanced to include TWODANT/GQ and THREEDANT. TWODANT/GQ solves the two-dimensional form of the discrete ordinates approximation to the transport equation on a generalized quadrilateral mesh. This geometric capability is very general and allows nearly exact representations of X-Y or R-Z geometries. THREEDANT solves the three-dimensional form of the discrete ordinates equations. In addition to the conventional coarse-mesh material zone input, THREEDANT can also be linked to a three-dimensional nested-region mesh generation code called FRAC-IN-THE-BOX. THREEDANT can thus model a much wider variety of geometric shapes than any other discrete ordinates code. These enhanced geometric modeling capabilities are applied here to the analysis of the VENUS PWR Mock-Up Facility
DRAGON 3.05D, Reactor Cell Calculation System with Burnup
International Nuclear Information System (INIS)
2007-01-01
1 - Description of program or function: The computer code DRAGON contains a collection of models that can simulate the neutron behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: the interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations that can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. 2 - Methods: The code DRAGON contains a multigroup flux solver conceived that can use a various algorithms to solve the neutron transport equation for the spatial and angular distribution of the flux. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are considered as sources. The current release of DRAGON contains five such algorithms. The JPM option that solves the integral transport equation using the J+- method, (interface current method applied to homogeneous blocks); the SYBIL option that solves the integral transport equation using the collision probability method for simple one dimensional (1-D) or two dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; the EXCELL/NXT option to solve the integral transport equation using the collision probability method for more general 2-D geometries and for three dimensional (3-D) assemblies; the MOCC option to solve the transport equation using the method of cyclic characteristics in 2-D Cartesian, and finally the MCU option to solve the transport equation using the method of characteristics (non cyclic) for 3-D Cartesian geometries. The execution of DRAGON is
Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.
Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K
2013-10-01
Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup. Copyright © 2013 Elsevier Ltd. All rights reserved.
On the calculation of the minimax-converse of the channel coding problem
Elkayam, Nir; Feder, Meir
2015-01-01
A minimax-converse has been suggested for the general channel coding problem by Polyanskiy etal. This converse comes in two flavors. The first flavor is generally used for the analysis of the coding problem with non-vanishing error probability and provides an upper bound on the rate given the error probability. The second flavor fixes the rate and provides a lower bound on the error probability. Both converses are given as a min-max optimization problem of an appropriate binary hypothesis tes...
Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET
Energy Technology Data Exchange (ETDEWEB)
Kliem, S.
1998-10-01
Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)
Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET
International Nuclear Information System (INIS)
Kliem, S.
1998-01-01
Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)
INTERTRAN 2 - A computer code for calculating the risk from transportation of radioactive materials
International Nuclear Information System (INIS)
Ericsson, A.M.; Jaernry, C.
1993-01-01
In this paper a description of IAEA Coordinated Research Program (CRP) dealing with the updating of the computer code INTERTRAN is given. The paper includes a summary of the work performed by several member states within the CRP as well as gives a description of the final product that will be presented to the IAEA. (J.P.N.)
The Wims-Traca code for the calculation of fuel elements. User's manual and input data
International Nuclear Information System (INIS)
Anhert, C.
1980-01-01
The set of modifications and new options developped for the Wims-D code is explained. The input data of the new version Wims-Traca are described. The printed output of results is also explained. The contents and the source of the nuclear data in the basic library is exposed. (author)
Calculation of the fine spectrum and integration of the resonance cross sections in the cells
International Nuclear Information System (INIS)
Paratte, J.M.
1986-10-01
The code BOXER is used for the neutronics calculations of two-dimensional LWR arrays. During the calculation of the group constants of the cells (pin, clad and moderator), the program SLOFIN, a BOXER module, allows taking into account the self-shielding of the resonances. The resonance range is devided into two parts: - above 907 eV the cross sections are condensed into groups by the library code ETOBOX. In SLOFIN, these values are interpolated over the equivalent cross section and the temperature. The interpolation formula chosen gives an accuracy better than 1% for values of the equivalent cross section larger than 5 barns. - between 4 and 907 eV, the cross sections are given in pointwise form as a function of the lethargy. At first a list of pointwise macroscopic cross section is established. Then the fine spectrum in the cell is calculated in 2 or 3 zones by means of the collision probability theory. In the central zone one resonant pseudo-nuclide is considered for the calculation of the scattering source, while the light nuclides are explicitly treated but under the assumption of energy independent cross sections. The fine spectrum is then used as a weihting function for the condensation of the pointwise cross sections of the resonant nuclides into energy groups. The procedure was checked on the basis of the TRX-1 to -4 and BAPL-UO 2 -1 to -3 experiments which are used as benchmarks for the tests of the ENDF/B libraries. The comparisons with other calculation results show that the deviations observed are typical for the basic cross sections. The method proposed shows a good accuracy in the application range foreseen for BOXER. It is also fast enough to be used as a standard method in a cell code. (author)
International Nuclear Information System (INIS)
Broeders, I.; Krieg, B.
1977-01-01
The code MIGROS-3 was developed from MIGROS-2. The main advantage of MIGROS-3 is its compatibility with the new conventions of the latest version of the Karlsruhe nuclear data library, KEDAK-3. Moreover, to some extent refined physical models were used and numerical methods were improved. MIGROS-3 allows the calculation of microscopic group cross sections of the ABBN type from isotopic neutron data given in KEDAK-format. All group constants, necessary for diffusion-, consistent P 1 - and Ssub(N)-calculations can be generated. Anisotropy of elastic scattering can be taken into account up to P 5 . A description of the code and the underlying theory is given. The input and output description, a sample problem and the program lists are provided. (orig.) [de
Particle-in-Cell Calculations of the Electron Cloud in the ILC Positron Damping Ring Wigglers
International Nuclear Information System (INIS)
Celata, C.M.; Furman, M.A.; Vay, J.-L.; Grote, D.P.
2007-01-01
The self-consistent code suite WARP-POSINST is being used to study electron cloud effects in the ILC positron damping ring wiggler. WARP is a parallelized, 3D particle-in-cell code which is fully self-consistent for all species. The POSINST models for the production of photoelectrons and secondary electrons are used to calculate electron creation. Mesh refinement and a moving reference frame for the calculation will be used to reduce the computer time needed by several orders of magnitude. We present preliminary results for cloud buildup showing 3D electron effects at the nulls of the vertical wiggler field. First results from a benchmark of WARP-POSINST vs. POSINST are also discussed
International Nuclear Information System (INIS)
Jameson, A.; Leicher, S.; Dawson, J.; Tel Aviv Univ., Israel)
1985-01-01
A multiblock modification of the FLO57 code for three-dimensional wing calculations is described and demonstrated. The theoretical basis of the multistage time-stepping algorithm is reviewed; the multiblock grid structure is explained; and results from a computation of vortical flow past a delta wing, using 2.5 x 10 to the 6th grid points and performed on a Cray X/MP computer with a 128-Mword solid-state storage device, are presented graphically. 6 references
Energy Technology Data Exchange (ETDEWEB)
Williams, M.L.; Yuecel, A.; Nadkarny, S.
1988-05-01
The HEATING6 heat conduction code is modified to (a) read the multigroup particle fluxes from a two-dimensional DOT-IV neutron- photon transport calculation, (b) interpolate the fluxes from the DOT-IV variable (optional) mesh to the HEATING6 control volume mesh, and (c) fold the interpolated fluxes with kerma factors to obtain a nuclear heating source for the heat conduction equation. The modified HEATING6 is placed as a module in the ORNL discrete ordinates system (DOS), and has been renamed DOS-HEATING6. DOS-HEATING6 provides the capability for determining temperature distributions due to nuclear heating in complex, multi-dimensional systems. All of the original capabilities of HEATING6 are retained for the nuclear heating calculation; e.g., generalized boundary conditions (convective, radiative, finned, fixed temperature or heat flux), temperature and space dependent thermal properties, steady-state or transient analysis, general geometry description, etc. The numerical techniques used in the code are reviewed and the user input instructions and JCL to perform DOS-HEATING6 calculations are presented. Finally a sample problem involving coupled DOT-IV and DOS-HEATING6 calculations of a complex space-reactor configurations described, and the input and output of the calculations are listed. 10 refs., 11 figs., 6 tabs.
Comparison calculations of WWER-1000 fuel assemblies by using the MCNP 4.2 a KASSETA codes
International Nuclear Information System (INIS)
Trgina, M.
1993-12-01
The power multiplication and distribution factors are compared for various geometries and material configurations of WWER-1000 fuel assemblies. The calculations were performed in 2 ways: (i) using nuclear data, employing older and current data collections, and (ii) using the author's own model based on the KASSETA code. The comparison code MCNP 4.2 is described, intended for computerized simulation of the transport of neutrons, photons and electrons. This code uses its own cross section library. The methodology is outlined and a specification of the Monte Carlo method employed is given. The use of the refined data library gave rise to appreciable deviations of the multiplication factors in all variants. The use of the older data library led to identical criticality results for the variant with water holes. For inserted absorbers the discrepancies in criticality and in power distribution data are appreciable. The marked disagreement between the results of application of the MCNP 4.2 and KASSETA codes for the variants with inserted control elements is indicative of inappropriateness of the approximation procedure in the latter code. (J.B.). 2 tabs., 11 figs., 11 refs
Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes
International Nuclear Information System (INIS)
Harrisson, G.; Marleau, G.
2012-01-01
The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)
Energy Technology Data Exchange (ETDEWEB)
Sharma, R.; Ratkowski, A.J.; Sundberg, R.L.; Duff, J.W.; Bernstein, L.S.
1989-02-01
The Strategic High-Altitude Radiance Code (SHARC ) is a new computer code that calculates atmospheric radiation for paths from 60 to 300 km altitude in the 2-40 micro spectral region. It models radiation due to NLTE (Non-Local Thermodynamic Equilibrium) molecular emissions which are the dominant sources at these altitudes. The initial version of SHARC, which is described in this paper, includes the five strongest IR radiators, CO/sub 2/, NO, O/sub 3/, H/sub 2/O and CO. Calculation of excited state populations is accomplished by interfacing a Monte Carlo model for radiative excitation and energy transfer with a highly flexible chemical kinetics module derived from the Sandia CHEMKIN Code. An equivalent-width, line-by-line approach for the radiation transport gives a spectral resolution of about 0.50/cm. The radiative-transport calculation includes the effects of combined Doppler-Lorentz (Voigt) line shapes. Particular emphasis was placed on modular construction and supporting data files so that models and model parameters can be modified or upgraded as additional data become available. The initial version of SHARC is now ready for distribution.
Energy Technology Data Exchange (ETDEWEB)
Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)
2016-06-15
Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
Energy Technology Data Exchange (ETDEWEB)
Elsawi, Mohamed A., E-mail: Mohamed.elsawi@kustar.ac.ae; Hraiz, Amal S. Bin, E-mail: Amal.Hraiz@kustar.ac.ae
2015-11-15
Highlights: • AP1000 core configuration is challenging due to its high degree of heterogeneity. • The proposed code was used to model neutronics/TH behavior of the AP1000 reactor. • Enhanced modeling features in WIMS9 facilitated neutronics modeling of the reactor. • PARCS/TRACE coupled code system was used to model the temperature feedback effects. • Final results showed reasonable agreement with publically available reactor data. - Abstract: The objective of this paper is to assess the accuracy of the WIMS9/PARCS/TRACE code system for power density calculations of the Westinghouse AP1000™ nuclear reactor, as a representative of modern pressurized water reactors (Gen III+). The cross section libraries were generated using the lattice physics code WIMS9 (the commercial version of the legacy lattice code WIMSD). Nine different fuel assembly types were analyzed in WIMS9 to generate the two-group cross sections required by the PARCS core simulator. The nine fuel assembly types were identified based on the distribution of Pyrex discrete burnable absorber (Borosilicate glass) and integral fuel burnable absorber (IFBA) rods in each fuel assembly. The generated cross sections were passed to the coupled core simulator PARCS/TRACE which performed 3-D, full-core diffusion calculations from within the US NRC Symbolic Nuclear Analysis Package (SNAP) interface. The results which included: assembly power distribution, effective multiplication factor (k{sub eff}), radial and axial power density, and whole core depletion were compared to reference Monte Carlo results and to a published reactor data available in the AP1000 Design Control Document (DCD). The results of the study show acceptable accuracy of the WIMS9/PARCS/TRACE code in predicting the power density of the AP1000 core and, hence, establish its adequacy in the evaluation of the neutronics parameters of modern PWRs of similar designs. The work reported here is new in that it uses, for the first time, the
The GRAPE code system for the calculation of precompound and compound nuclear reactions
International Nuclear Information System (INIS)
Gruppelaar, H.; Akkermans, J.M.
1985-02-01
The statistical exciton model following the master-equation approach has been improved and extended for application as an evaluation tool of double-differential reaction cross sections at incident nucleon energies of 5 to 50 MeV. For this purpose the code system GRAPE has been developed. An important characteristic of the proposed model is that consistency with equilibrium models has been demanded for the summed exciton-state densities as well as for the particle and γ-ray emission cross sections. Consistency with the adopted state densities has also been imposed upon the internal transition rates. A survey of the theory is given and the structure of the GRYPHON code is described. This report also contains a users' manual for GRYPHON
TBCI and URMEL - New computer codes for wake field and cavity mode calculations
International Nuclear Information System (INIS)
Weiland, T.
1983-01-01
Wake force computation is important for any study of instabilities in high current accelerators and storage rings. These forces are generated by intense bunches of charged particles passing cylindrically symmetric structures on or off axis. The adequate method for computing such forces is the time domain approach. The computer Code TBCI computes for relativistic as well as for nonrelativistic bunches of arbitrary shape longitudinal and transverse wake forces up to the octupole component. TBCI is not limited to cavity-like objects and thus applicable to bellows, beam pipes with varying cross sections and any other nonresonant structures. For the accelerating cavities one also needs to know the resonant modes and frequencies for the study of instabilities and mode couplers. The complementary code named URMEL computes these fields for any azimuthal dependence of the fields in ascending order. The mathematical procedure being used is very safe and does not miss modes. Both codes together represent a unique tool for accelerator design and are easy to use
Development of explicit solution scheme for the MATRA-LMR code and test calculation
Energy Technology Data Exchange (ETDEWEB)
Jeong, H. Y.; Ha, K. S.; Chang, W. P.; Kwon, Y. M.; Jeong, K. S
2003-01-01
The local blockage in a subassembly of a liquid metal reactor is of particular importance because local sodium boiling could occur at the downstream of the blockage and integrity of the fuel clad could be threatened. The explicit solution scheme of MATRA-LMR code is developed to analyze the flow blockage in a subassembly of a liquid metal cooled reactor. In the present study, the capability of the code is extended to the analysis of complete blockage of one or more subchannels. The results of the developed solution scheme shows very good agreement with the results obtained from the implicit scheme for the experiments of flow channel without any blockage. The applicability of the code is also evaluated for two typical experiments in a blocked channel. Through the sensitivity study, it is shown that the explicit scheme of MATRA-LMR predicts the flow and temperature profile after blockage reasonably if the effect of wire is suitably modeled. The simple assumption in wire-forcing function is effective for the un-blocked case or for the case of blockage with lower velocity. A different type of wire-forcing function describing the velocity reduction after blockage or an accurate distributed resistance model is required for more improved predictions.
BOA, Beam Optics Analyzer A Particle-In-Cell Code
International Nuclear Information System (INIS)
Bui, Thuc
2007-01-01
The program was tasked with implementing time dependent analysis of charges particles into an existing finite element code with adaptive meshing, called Beam Optics Analyzer (BOA). BOA was initially funded by a DOE Phase II program to use the finite element method with adaptive meshing to track particles in unstructured meshes. It uses modern programming techniques, state-of-the-art data structures, so that new methods, features and capabilities are easily added and maintained. This Phase II program was funded to implement plasma simulations in BOA and extend its capabilities to model thermal electrons, secondary emissions, self magnetic field and implement a more comprehensive post-processing and feature-rich GUI. The program was successful in implementing thermal electrons, secondary emissions, and self magnetic field calculations. The BOA GUI was also upgraded significantly, and CCR is receiving interest from the microwave tube and semiconductor equipment industry for the code. Implementation of PIC analysis was partially successful. Computational resource requirements for modeling more than 2000 particles begin to exceed the capability of most readily available computers. Modern plasma analysis typically requires modeling of approximately 2 million particles or more. The problem is that tracking many particles in an unstructured mesh that is adapting becomes inefficient. In particular memory requirements become excessive. This probably makes particle tracking in unstructured meshes currently unfeasible with commonly available computer resources. Consequently, Calabazas Creek Research, Inc. is exploring hybrid codes where the electromagnetic fields are solved on the unstructured, adaptive mesh while particles are tracked on a fixed mesh. Efficient interpolation routines should be able to transfer information between nodes of the two meshes. If successfully developed, this could provide high accuracy and reasonable computational efficiency.
A modified version of the Monte Carlo computer code for calculating neutron detection efficiencies
International Nuclear Information System (INIS)
Nakayama, K.; Pessoa, E.F.; Douglas, R.A.
1980-12-01
A calculation of neutron detection efficiencies has been performed for organic scintillators using the Monte Carlo Method. Effects which contribute to the detection efficiency have been incorporated in the calculations as thoroughly as possible. The reliability of the results is verified by comparison with the efficiency measurements available in the literature for neutrons in the energy range between 1 and 170 MeV with neutron detection thresholds between O.1 and 22.3 MeV. (Author) [pt
Computer code for the calculation of the performance of a pulsed chemical laser
International Nuclear Information System (INIS)
Waichman, K.; Rosenwaks, Z.; Chuchem, D.; Achiam, Ya.
1981-02-01
A model for the calculation of the performance of a pulsed HF chemical laser is presented. it serves to solve rate equations of the densities of gases, level populations and temperature equation, as well as to calculate the lasing energy. The assumptions are: gain equal loss condition; a Boltzmann distribution of the rotational level populations at the translational temperature; the cavity is of the Fabry-Perot type. (Author)
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.
Directory of Open Access Journals (Sweden)
V. V. Galchenko
2016-12-01
Full Text Available The description of calculation scheme of fuel assembly for preparation of few-group characteristics is considered with help of Serpent code. This code uses the Monte-Carlo method and energy continuous microscopic data libraries. Serpent code is devoted for calculation of fuel assembly characteristics, burnup calculations and preparation of few-group homogenized macroscopic cross-sections. The results of verification simulations in comparison with other codes (WIMS, HELIOS, NESSEL etc., which are used for neutron-physical analysis of VVER type fuel, are presented.
Energy Technology Data Exchange (ETDEWEB)
Wilson, W. B. (William B.); Perry, R. T. (Robert T.); Shores, E. F. (Erik F.); Charlton, W. S. (William S.); Parish, Theodore A.; Estes, G. P. (Guy P.); Brown, T. H. (Thomas H.); Arthur, Edward D. (Edward Dana),; Bozoian, Michael; England, T. R.; Madland, D. G.; Stewart, J. E. (James E.)
2002-01-01
SOURCES 4C is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to radionuclide decay. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., an intimate mixture of a-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 44 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 107 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code provides the magnitude and spectra, if desired, of the resultant neutron source in addition to an analysis of the'contributions by each nuclide in the problem. LASTCALL, a graphical user interface, is included in the code package.
State of the art of aerolastic codes for wind turbine calculations
Energy Technology Data Exchange (ETDEWEB)
Maribo Pedersen, B. [ed.
1996-09-01
The technological development of modern wind turbines has been dependent on the parallel development of the computational skills of the designers. The combination of the calculation of the flow field around the wind turbine rotor - both far field and near field - and the calculation of the response of the wind turbine structure to the resulting, non-stationary air loads, also known as aero-elastic calculations have now reached a reasonable degree of maturity. At this expert meeting two main points may be clarified. To what level of accuracy can we now determine the behaviour of the different elements of a wind turbine, i.e. how well are we able to compute deflections, fluctuating loads and power output. Which are the main outstanding areas upon which our next research efforts should be focused. (EG)
RMG An Open Source Electronic Structure Code for Multi-Petaflops Calculations
Briggs, Emil; Lu, Wenchang; Hodak, Miroslav; Bernholc, Jerzy
RMG (Real-space Multigrid) is an open source, density functional theory code for quantum simulations of materials. It solves the Kohn-Sham equations on real-space grids, which allows for natural parallelization via domain decomposition. Either subspace or Davidson diagonalization, coupled with multigrid methods, are used to accelerate convergence. RMG is a cross platform open source package which has been used in the study of a wide range of systems, including semiconductors, biomolecules, and nanoscale electronic devices. It can optionally use GPU accelerators to improve performance on systems where they are available. The recently released versions (>2.0) support multiple GPU's per compute node, have improved performance and scalability, enhanced accuracy and support for additional hardware platforms. New versions of the code are regularly released at http://www.rmgdft.org. The releases include binaries for Linux, Windows and MacIntosh systems, automated builds for clusters using cmake, as well as versions adapted to the major supercomputing installations and platforms. Several recent, large-scale applications of RMG will be discussed.
Beam Dynamics in an Electron Lens with the Warp Particle-in-cell Code
Stancari, Giulio; Redaelli, Stefano
2014-01-01
Electron lenses are a mature technique for beam manipulation in colliders and storage rings. In an electron lens, a pulsed, magnetically confined electron beam with a given current-density profile interacts with the circulating beam to obtain the desired effect. Electron lenses were used in the Fermilab Tevatron collider for beam-beam compensation, for abort-gap clearing, and for halo scraping. They will be used in RHIC at BNL for head-on beam-beam compensation, and their application to the Large Hadron Collider for halo control is under development. At Fermilab, electron lenses will be implemented as lattice elements for nonlinear integrable optics. The design of electron lenses requires tools to calculate the kicks and wakefields experienced by the circulating beam. We use the Warp particle-in-cell code to study generation, transport, and evolution of the electron beam. For the first time, a fully 3-dimensional code is used for this purpose.
International Nuclear Information System (INIS)
Dos Santos, Adimir; Siqueira, Paulo de Tarso D.; Andrade e Silva, Graciete Simões; Grant, Carlos; Tarazaga, Ariel E.; Barberis, Claudia
2013-01-01
In year 2008 the Atomic Energy National Commission (CNEA) of Argentina, and the Brazilian Institute of Energetic and Nuclear Research (IPEN), under the frame of Nuclear Energy Argentine Brazilian Agreement (COBEN), among many others, included the project “Validation and Verification of Calculation Methods used for Research and Experimental Reactors . At this time, it was established that the validation was to be performed with models implemented in the deterministic codes HUEMUL and PUMA (cell and reactor codes) developed by CNEA and those ones implemented in MCNP by CNEA and IPEN. The necessary data for these validations would correspond to theoretical-experimental reference cases in the research reactor IPEN/MB-01 located in São Paulo, Brazil. The staff of the group Reactor and Nuclear Power Studies (SERC) of CNEA, from the argentine side, performed calculations with deterministic models (HUEMUL-PUMA) and probabilistic methods (MCNP) modeling a great number of physical situations of de reactor, which previously have been studied and modeled by members of the Center of Nuclear Engineering of the IPEN, whose results were extensively provided to CNEA. In this paper results of comparison of calculated and experimental results for critical configurations, temperature coefficients, kinetic parameters and fission rates evaluated with probabilistic models spatial distributions are shown. (author)
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)
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
Decay heat measurement on fusion reactor materials and validation of calculation code system
Energy Technology Data Exchange (ETDEWEB)
Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)
Calculation of low-cycle fatigue in accordance with the national standard and strength codes
Kontorovich, T. S.; Radin, Yu. A.
2017-08-01
Over the most recent 15 years, the Russian power industry has largely relied on imported equipment manufactured in compliance with foreign standards and procedures. This inevitably necessitates their harmonization with the regulatory documents of the Russian Federation, which include calculations of strength, low cycle fatigue, and assessment of the equipment service life. An important regulatory document providing the engineering foundation for cyclic strength and life assessment for high-load components of the boiler and steamline of a water/steam circuit is RD 10-249-98:2000: Standard Method of Strength Estimation in Stationary Boilers and Steam and Water Piping. In January 2015, the National Standard of the Russian Federation 12952-3:2001 was introduced regulating the issues of design and calculation of the pressure parts of water-tube boilers and auxiliary installations. Thus, there appeared to be two documents simultaneously valid in the same energy field and using different methods for calculating the low-cycle fatigue strength, which leads to different results. In this connection, the current situation can lead to incorrect ideas about the cyclic strength and the service life of high-temperature boiler parts. The article shows that the results of calculations performed in accordance with GOST R 55682.3-2013/EN 12952-3: 2001 are less conservative than the results of the standard RD 10-249-98. Since the calculation of the expected service life of boiler parts should use GOST R 55682.3-2013/EN 12952-3: 2001, it becomes necessary to establish the applicability scope of each of the above documents.
Second order gyrokinetic theory for particle-in-cell codes
Tronko, Natalia; Bottino, Alberto; Sonnendrücker, Eric
2016-08-01
The main idea of the gyrokinetic dynamical reduction consists in a systematical removal of the fast scale motion (the gyromotion) from the dynamics of the plasma, resulting in a considerable simplification and a significant gain of computational time. The gyrokinetic Maxwell-Vlasov equations are nowadays implemented in for modeling (both laboratory and astrophysical) strongly magnetized plasmas. Different versions of the reduced set of equations exist, depending on the construction of the gyrokinetic reduction procedure and the approximations performed in the derivation. The purpose of this article is to explicitly show the connection between the general second order gyrokinetic Maxwell-Vlasov system issued from the modern gyrokinetic theory and the model currently implemented in the global electromagnetic Particle-in-Cell code ORB5. Necessary information about the modern gyrokinetic formalism is given together with the consistent derivation of the gyrokinetic Maxwell-Vlasov equations from first principles. The variational formulation of the dynamics is used to obtain the corresponding energy conservation law, which in turn is used for the verification of energy conservation diagnostics currently implemented in ORB5. This work fits within the context of the code verification project VeriGyro currently run at IPP Max-Planck Institut in collaboration with others European institutions.
A three-dimensional transient calculation for the reactor model RAMONA using the COMMIX-2(V) code
International Nuclear Information System (INIS)
Weinberg, D.; Frey, H.H.; Tschoeke, H.
1993-01-01
The safety graded decay heat removal system of the European Fast Reactor needs a high availability. This system operates entirely under natural convection. To guarantee a proper design, experiments are carried out to verify thermal-hydraulic computer codes able to predict precisely local temperature loadings of the components and the reactor tank in the transition region from nominal operation under forced convection to the decay heat removal operation. - With the COMMIX-2 (V) code three-dimensional transient calculations have been performed to simulate experiments in the 360 deg. reactor model RAMONA, scaled 1:20 to the reality with water as simulant fluid for sodium. The computed average and local temperatures as well as the velocity distributions show a good agreement with the experimental results. Further efforts are necessary to reduce the computation time. (orig.)
International Nuclear Information System (INIS)
Calando, J.P.
1991-07-01
The ORION code is designed to determine very quickly the immediate consequences (such as plume passage time, instantaneous maximum hazards irradiation, inhalation, deposit) due to an accident spreading out radioactive or chemical pollution into the atmosphere, from a source point, a stack release, (with heightening calculation) outspread sources (transport accident such as, for instance, road fire or car crash) or from a cylindrical cloud defined by different vertical sources (for instance pyrotechnical accident, missile firing...). The diffusion code DOURY type (french official methods) is written in FORTRAN. Data are entered in a conversational mode with auto-checking. Results are output to tables an isorisks curves drawn at map scales. At the Bruyeres-le-Chatel Radiation Protection Unit, a team is on permanent duty, can carry out results in a few minutes and transmit the evaluation by TELEFAX anywhere on the National territory [fr
Directory of Open Access Journals (Sweden)
Recep Ali Dedecan
2013-08-01
Full Text Available The goal of this paper is to review the validity of seismic assessment procedure given in the Turkish Earthquake Code by comparing the assessment results with real structures from Eastern Turkey, where the 2011 Van earthquake occurred. To test the analysis methods for a typically suitable structure, the cultural center building at Erciş with 3 stories, is selected. In order to compare the results of the three different analysis techniques, for an identical earthquake, the ground motion used in analysis was characterized by equivalent elastic earthquake spectra, which were developed from available time history at the nearest construction site. It was found that the damage predictions by using the by Turkish Earthquake Code procedures point out the different level of damages. But, it is concluded that nonlinear time history analysis calculated the best estimation of the damage observed in the site.
Prostate dose calculations for permanent implants using the MCNPX code and the Voxels phantom MAX
Energy Technology Data Exchange (ETDEWEB)
Reis Junior, Juraci Passos dos; Silva, Ademir Xavier da, E-mail: jjunior@con.ufrj.b, E-mail: Ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Facure, Alessandro N.S., E-mail: facure@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)
2010-07-01
This paper presents the modeling of 80, 88 and 100 of {sup 125}I seeds, punctual and volumetric inserted into the phantom spherical volume representing the prostate and prostate phantom voxels MAX. Starting values of minimum and maximum activity, 0.27 mCi and 0.38 mCi, respectively, were simulated in the Monte Carlo code MCNPX in order to determine whether the final dose, according to the integration of the equation of decay at time t = 0 to t = {infinity} corresponds to the default value set by the AAPM 64 which is 144 Gy. The results showed that consider sources results in doses exceeding the percentage discrepancy of the default value of 200%, while volumetric consider sources result in doses close to 144 Gy. (author)
EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes
Energy Technology Data Exchange (ETDEWEB)
Paolo Balestra; Carlo Parisi; Andrea Alfonsi
2016-02-01
The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.
Benchmarking of epithermal methods in the lattice-physics code EPRI-CELL
International Nuclear Information System (INIS)
Williams, M.L.; Wright, R.Q.; Barhen, J.; Rothenstein, W.; Toney, B.
1982-01-01
The epithermal cross section shielding methods used in the lattice physics code EPRI-CELL (E-C) have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology. These include: treatment of the external moderator source with intermediate resonance (IR) theory, development of a new Dancoff factor expression to account for clad interactions, development of a new method for treating resonance interference, and application of a generalized least squares method to compute best-estimate values for the Bell factor and group-dependent IR parameters. The modified E-C code with its new ENDF/B-V cross section library is tested for several numerical benchmark problems. Integral parameters computed by EC are compared with those obtained with point-cross section Monte Carlo calculations, and E-C fine group cross sections are benchmarked against point-cross section descrete ordinates calculations. It is found that the code modifications improve agreement between E-C and the more sophisticated methods. E-C shows excellent agreement on the integral parameters and usually agrees within a few percent on fine-group, shielded cross sections
International Nuclear Information System (INIS)
Tayal, M.
1987-01-01
Structures often operate at elevated temperatures. Temperature calculations are needed so that the design can accommodate thermally induced stresses and material changes. A finite element computer called FEAT has been developed to calculate temperatures in solids of arbitrary shapes. FEAT solves the classical equation for steady state conduction of heat. The solution is obtained for two-dimensional (plane or axisymmetric) or for three-dimensional problems. Gap elements are use to simulate interfaces between neighbouring surfaces. The code can model: conduction; internal generation of heat; prescribed convection to a heat sink; prescribed temperatures at boundaries; prescribed heat fluxes on some surfaces; and temperature-dependence of material properties like thermal conductivity. The user has a option of specifying the detailed variation of thermal conductivity with temperature. For convenience to the nuclear fuel industry, the user can also opt for pre-coded values of thermal conductivity, which are obtained from the MATPRO data base (sponsored by the U.S. Nuclear Regulatory Commission). The finite element method makes FEAT versatile, and enables it to accurately accommodate complex geometries. The optional link to MATPRO makes it convenient for the nuclear fuel industry to use FEAT, without loss of generality. Special numerical techniques make the code inexpensive to run, for the type of material non-linearities often encounter in the analysis of nuclear fuel. The code, however, is general, and can be used for other components of the reactor, or even for non-nuclear systems. The predictions of FEAT have been compared against several analytical solutions. The agreement is usually better than 5%. Thermocouple measurements show that the FEAT predictions are consistent with measured changes in temperatures in simulated pressure tubes. FEAT was also found to predict well, the axial variations in temperatures in the end-pellets(UO 2 ) of two fuel elements irradiated
Mairani, A; Valente, M; Battistoni, G; Botta, F; Pedroli, G; Ferrari, A; Cremonesi, M; Di Dia, A; Ferrari, M; Fasso, A
2011-01-01
Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy ((89)Sr, (90)Y, (131)I, (153)Sm, (177)Lu, (186)Re, and (188)Re). Point isotropic...
1983-07-08
tangent heights which define the looking geometry for each desired case. TANIN is the lowest tangent height, TANF the highest, and INTRVL is the spacing...wavenumber desired cm ~I Real, int 750 ’ (CARD IC) TANIN Initial tangent height km Integer 70 TANF Final tangent height km Integer 80 INTRVL...less than or equal to TANIN . The highest must be greater than or equal to TANF + 50, since the calculations are done using 50 layers. V A procedure for
Energy Technology Data Exchange (ETDEWEB)
Madland, D.G.; Arthur, E.D.; Estes, G.P.; Stewart, J.E.; Bozoian, M.; Perry, R.T.; Parish, T.A.; Brown, T.H.; England, T.R.; Wilson, W.B.; Charlton, W.S.
1999-09-01
SOURCES 4A is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., a mixture of {alpha}-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an analysis of the contributions to that source by each nuclide in the problem.
International Nuclear Information System (INIS)
Ohoka, Yasunori; Tatsumi, Masahiro; Sugimura, Naoki; Tabuchi, Masato
2011-01-01
In 2010, the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0) has been released by JAEA. JENDL-4.0 is major update from JENDL- 3.3, and confirmed to give good accuracy by integral test for fission reactor systems such as fast neutron system and thermal neutron system. In this study, we evaluated the reactivity impact due to difference between ENDF/B-VII.0 and JENDL-4.0 for PWR fuel assembly burnup calculation using AEGIS code which has been developed by Nuclear Engineering, Ltd. in cooperation with Nuclear Fuel Industries, Ltd. and Nagoya University
Menendez, Javier A; Alarcón, Tomás
2016-01-01
Generation of cancer stem cell (CSC)-like cells might occur through metabolic corruption of the epigenetic codes that govern cell identity. We recently identified how archetypal oncometabolites, without altering the baseline expression of endogenous stem cell maintenance genes but endowing cells with epigenetic states refractory to differentiation, considerably enhance the global kinetic efficiency of nuclear reprogramming processes that generate CSC-like states de novo . This study highlights that metabolo-epigenetic axes of communication can direct the development and maintenance of CSCs during the natural history of cancer diseases.
An interactive computer code for calculation of gas-phase chemical equilibrium (EQLBRM)
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.
Modelling of preheated regenerative chain in Cernavoda NPP using MMS calculation code
International Nuclear Information System (INIS)
Bigu, M.; Nita, I.; Prisecaru, I.; Dupleac, D.
2005-01-01
Full text: In this work it was studied operation of preheated regenerative chain from NPP Cernavoda. To obtain this analysis coupled analyses of condensate system, water supply system, and drain cooler system were effected. The analysis boundaries are: Upstream: - Steam condensers - Turbine Bleed Steam Downstream: - Steam Generators. The analysis was made in two steps: 1) Getting of hydraulic characteristic of pipe network from steam condensers to steam generators at nominal regime; this step was obtained with hydraulic package called PIPENET. 2) Real thermal hydraulic analyses were done based on hydraulic characteristic of pipe network and supplementary data required for heat transfer calculation in equipment of preheated regenerative chain. Thermal analyses were done using MMS package and refered to normal operating regimes, namely, nominal operating regime required for calibration of calculating model, shutdown regime, start-up regime from zero power hot to nominal power and to abnormal operating regimes, namely, turbine trip, reactor trip and loss of two condensate pumps. The results were compared with already existing analysis and showed the largest differences at interface areas (i.e. 5%). This led us to idea of extending analysis to all secondary circuits in order to reduce the number of boundary conditions which can generate uncertainty in analysis. In this analysis we obtained an advanced model of preheated regenerative chain of secondary circuit in Cernavoda NPP which could be extended up to cover the whole secondary circuit by including the analysis of steam generators, turbine, and steam condenser. (authors)
Modelling of preheated regenerative chain in Cernavoda NPP using MMS calculation code
International Nuclear Information System (INIS)
Bigu, M.; Nita, I.; Prisecaru, I.; Dupleac, D.
2005-01-01
In this work it was studied operation of preheated regenerative chain from NPP Cernavoda. To obtain this analysis coupled analyses of condensate system, water supply system, and drain cooler system were effected. The analysis boundaries are: Upstream: - Steam condensers - Turbine Bleed Steam Downstream: - Steam Generators. The analysis was made in two steps: 1) Getting of hydraulic characteristic of pipe network from steam condensers to steam generators at nominal regime; this step was obtained with hydraulic package called PIPENET. 2) Real thermal hydraulic analyses were done based on hydraulic characteristic of pipe network and supplementary data required for heat transfer calculation in equipment of preheated regenerative chain. Thermal analyses were done using MMS package and referred to normal operating regimes, namely, nominal operating regime required for calibration of calculating model, shutdown regime, start-up regime from zero power hot to nominal power and to abnormal operating regimes, namely, turbine trip, reactor trip and loss of two condensate pumps. The results were compared with already existing analysis and showed the largest differences at interface areas (i.e. 5%). This led US to idea of extending analysis to all secondary circuits in order to reduce the number of boundary conditions which can generate uncertainty in analysis. In this analysis we obtained an advanced model of preheated regenerative chain of secondary circuit in Cernavoda NPP which could be extended up to cover the whole secondary circuit by including the analysis of steam generators, turbine, and steam condenser. (authors)
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
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
International Nuclear Information System (INIS)
Campolina, Daniel de Almeida Magalhaes
2009-01-01
In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)
A calculation code for estimating the global inventory of tritium from the ICSI Pilot Installation
International Nuclear Information System (INIS)
Pavelescu, Alexandru Octavian; Prisecaru, Ilie; Stefanescu, Ioan
2007-01-01
The paper describes the TRITINV program which performs a global updating of the total tritium inventory within the whole ICSI heavy water detritiation installation. The calculation does not take into account the distribution or the type of the tritium inventory on the installation modules. The program stores all the values submitted by user or computed, into a database which is permanently upgraded. It can also create charts based on its data and print or export this data in other formats. The application has been successfully tested and responded well to the initial requirements. It can be modified or improved anytime using the Microsoft Access 2003 Program. An important advancement would consist in upgrading the TRITINV into a stand-alone application by editing and compiling it in Microsoft Visual Studio 2005. (authors)
PWR neutron ex-vessel detection calculations using three-dimensional codes
International Nuclear Information System (INIS)
Dekens, O.; Lefebvre, J.C.; Rohart, M.; Chiron, M.
1997-01-01
During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l'Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors)
International Nuclear Information System (INIS)
Mazurier, J.
1999-01-01
This thesis has been performed in the framework of national reference setting-up for absorbed dose in water and high energy photon beam provided with the SATURNE-43 medical accelerator of the BNM-LPRI (acronym for National Bureau of Metrology and Primary standard laboratory of ionising radiation). The aim of this work has been to develop and validate different user codes, based on PENELOPE Monte Carlo code system, to determine the photon beam characteristics and calculate the correction factors of reference dosimeters such as Fricke dosimeters and graphite calorimeter. In the first step, the developed user codes have permitted the influence study of different components constituting the irradiation head. Variance reduction techniques have been used to reduce the calculation time. The phase space has been calculated for 6, 12 and 25 MV at the output surface level of the accelerator head, then used for calculating energy spectra and dose distributions in the reference water phantom. Results obtained have been compared with experimental measurements. The second step has been devoted to develop an user code allowing calculation correction factors associated with both BNM-LPRI's graphite and Fricke dosimeters thanks to a correlated sampling method starting with energy spectra obtained in the first step. Then the calculated correction factors have been compared with experimental and calculated results obtained with the Monte Carlo EGS4 code system. The good agreement, between experimental and calculated results, leads to validate simulations performed with the PENELOPE code system. (author)
International Nuclear Information System (INIS)
Iwamoto, Osamu
2013-01-01
A nuclear reaction calculation code CCONE, which was developed for nuclear data evaluation for JENDL/AC-2008 and JENDL-4, has been upgraded to improve the prediction accuracy for calculated cross sections at nucleon incident energies higher than 20 MeV. Multiple particle emission, in which nucleons and complex particles up to α-particle are involved, from pre-equilibrium reaction process was implemented based on the sequential-decay calculations for all produced exciton states within the framework of the two-component exciton model. The effect of velocity-change of particle-emitting nuclei on the multiple emission in preequilibrium and compound processes, which was not included in the previous evaluations, was taken into account to obtain spectra in the laboratory system using an average velocity approximation for each composite/compound nucleus. Calculated nucleon emission spectra at nucleon incident energies from 20 to 200 MeV were compared with experimental and evaluated data for the proton- and neutron-induced reactions on 27 Al. The present results are in good agreement with experimental data. It was found that their predictions were better than those of JENDL/HE-2007 especially for low emission energies at high incident energies. (author)
International Nuclear Information System (INIS)
Barreras Caballero, A. A.; Hernandez Garcia, J.J.; Alfonso Laguardia, R.
2009-01-01
Were directly determined correction factors depending on the type camera beam quality, k, Q, and kQ, Qo, instead of the product (w, air p) Q, for three type cylindrical ionization chambers Pinpoint and divergent monoenergetic beams of photons in a wide range of energies (4-20 MV). The method of calculation used dispenses with the approaches taken in the classic procedure considered independent of braking power ratios and the factors disturbance of the camera. A detailed description of the geometry and materials chambers were supplied by the manufacturer and used as data input for the system 2006 of PENELOPE Monte Carlo calculation using a User code that includes correlated sampling, and forced interactions division of particles. We used a photon beam Co-60 as beam reference for calculating the correction factors for beam quality. No data exist for the cameras PTW 31014, 31015 and 31016 in the TRS-398 at they do not compare the results with data calculated or determined experimentally by other authors. (author)
Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes
Energy Technology Data Exchange (ETDEWEB)
Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)
2015-07-01
The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)
International Nuclear Information System (INIS)
Axmann, J.K.
1992-02-01
Research and development activities related to the light water high conversion reactor concept have been conducted at the Paul Scherrer Institut (PSI) in the framework of a joint Swiss/German co-operation, together with the Karlsruhe Nuclear Research Centre, Siemens/KWU and the Technical University of Braunschweig. The present report documents principally the validation of the DITUBS computer code system, developed at the Technical University of Braunschweig for LWHCR physics design calculations. Experimental results from six of the fourteen PROTEUS-LWHCR core configurations investigated in the Phase II programme serve as a bsis for the study. Thus, alternative methods and data-set options within the DITUBS system have been developed and applied for (a) obtaining an independent set of calculated correction factors for various individual effects in the experiments and (b) achieving improvements in C/E (calculation/experiment) values for the measured integral parameters, viz. k ∞ and reaction rate ratios. The solution of numerical benchmark problems - for validation of burnup calculations and fuel-element-geometry treatment - form part of the study, the DITUBS system being finally used to address questions related to technical and economic feasibility for range of LWHCR designs. (author) figs., tabs., 112 refs
Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes
International Nuclear Information System (INIS)
Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A.
2015-01-01
The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K eff at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)
International Nuclear Information System (INIS)
Petkov, Petko T.
2000-01-01
Most of the few-group three-dimensional nodal diffusion codes used for neutronics calculations of the WWER reactors use albedo type boundary conditions on the core-reflector boundary. The conventional albedo are group-to-group reflection probabilities, defined on each outer node face. The method of characteristics is used to calculate accurate albedo by the following procedure. A many-group two-dimensional heterogeneous core-reflector problem, including a sufficient part of the core and detailed description of the adjacent reflector, is solved first. From this solution the angular flux on the core-reflector boundary is calculated in all groups for all traced neutron directions. Accurate boundary conditions can be calculated for the radial, top and bottom reflectors as well as for the absorber part of the WWER-440 reactor control assemblies. The algorithm can be used to estimate also albedo, coupling outer node faces on the radial reflector in the axial direction. Numerical results for the WWER-440 reactor are presented. (Authors)
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)
International Nuclear Information System (INIS)
Dos Santos, Adimir; Siqueira, Paulo de Tarso D.; Andrade e Silva, Graciete Simões; Grant, Carlos; Tarazaga, Ariel E.; Barberis, Claudia
2013-01-01
In year 2008 the Atomic Energy National Commission (CNEA) of Argentina, and the Brazilian Institute of Energetic and Nuclear Research (IPEN), under the frame of Nuclear Energy Argentine Brazilian Agreement (COBEN), among many others, included the project “Validation and Verification of Calculation Methods used for Research and Experimental Reactors . At this time, it was established that the validation was to be performed with models implemented in the deterministic codes HUEMUL and PUMA (cell and reactor codes) developed by CNEA and those ones implemented in MCNP by CNEA and IPEN. The necessary data for these validations would correspond to theoretical-experimental reference cases in the research reactor IPEN/MB-01 located in São Paulo, Brazil. The staff of the group Reactor and Nuclear Power Studies (SERC) of CNEA, from the argentine side, performed calculations with deterministic models (HUEMUL-PUMA) and probabilistic methods (MCNP) modeling a great number of physical situations of de reactor, which previously have been studied and modeled by members of the Center of Nuclear Engineering of the IPEN, whose results were extensively provided to CNEA. In this paper results for critical configurations are shown. (author)
Iwamoto, Yosuke
2018-03-01
In this study, the Monte Carlo displacement damage calculation method in the Particle and Heavy-Ion Transport code System (PHITS) was improved to calculate displacements per atom (DPA) values due to irradiation by electrons (or positrons) and gamma rays. For the damage due to electrons and gamma rays, PHITS simulates electromagnetic cascades using the Electron Gamma Shower version 5 (EGS5) algorithm and calculates DPA values using the recoil energies and the McKinley-Feshbach cross section. A comparison of DPA values calculated by PHITS and the Monte Carlo assisted Classical Method (MCCM) reveals that they were in good agreement for gamma-ray irradiations of silicon and iron at energies that were less than 10 MeV. Above 10 MeV, PHITS can calculate DPA values not only for electrons but also for charged particles produced by photonuclear reactions. In DPA depth distributions under electron and gamma-ray irradiations, build-up effects can be observed near the target's surface. For irradiation of 90-cm-thick carbon by protons with energies of more than 30 GeV, the ratio of the secondary electron DPA values to the total DPA values is more than 10% and increases with an increase in incident energy. In summary, PHITS can calculate DPA values for all particles and materials over a wide energy range between 1 keV and 1 TeV for electrons, gamma rays, and charged particles and between 10-5 eV and 1 TeV for neutrons.
The programme PIP2 for lattice cell thermal calculations
International Nuclear Information System (INIS)
Clayton, A.J.
1964-08-01
The programme PIP2 solves the multigroup equations obtained by applying the method of collision probabilities to a fuel region (which may contain a cluster of fuel elements), and the SPECTROX flux assumption in a surrounding 'moderator'. The programme does not calculate collision probabilities for the fuel region and any geometry can be treated in the fuel region for which collision probabilities can be calculated. Lattice cell source problems may be treated and it is possible to include part of the physical moderator with the fuel region for treatment by the collision probability method. The programme is primarily intended for thermal fixed source problems, with the sources in the (physical moderator), but by including part of the moderator with the fuel it is possible to include fixed sources in the fuel for the study of fast effects. (author)
International Nuclear Information System (INIS)
Hively, L.M.; Miley, G.M.
1980-03-01
The code calculates flux-surfaced-averaged values of alpha density, current, and electron/ion heating profiles in realistic, non-circular tokamak plasmas. The code is written in FORTRAN and execute on the CRAY-1 machine at the Magnetic Fusion Energy Computer Center
Energy Technology Data Exchange (ETDEWEB)
Young, P.G.; Arthur, E.D.
1977-11-01
A new multistep Hauser--Feshbach code that includes corrections for preequilibrium effects is described. The code can calculate up to 60 decay reactions (cross sections and energy spectra) in one computation, and thereby provide considerable flexibility for handling processes with complicated reaction chains. Input parameter setup, problem output, and subroutine descriptions are given along with a sample problem calculation. A brief theoretical description is also included. 8 figures, 3 tables.
International Nuclear Information System (INIS)
Bubelis, E.; Pabarcius, R.; Demcenko, M.
2001-01-01
Calculations of the main neutron-physical characteristics of RBMK-1500 reactors of Ignalina NPP Unit 1 and Unit 2 were performed, taking real reactor core states as the basis for these calculations. Comparison of the calculation results, obtained using QUABOX/CUBBOX code, with experimental data and the calculation results, obtained using STEPAN code, showed that all the main neutron-physical characteristics of the reactors of Unit 1 and Unit 2 of Ignalina NPP are in the safe deviation range of die analyzed parameters, and that reactors of Ignalina NPP, during the process of the reactor core composition change, are operated in a safe and stable manner. (author)
Shielding calculations for industrial 5/7.5MeV electron accelerators using the MCNP Monte Carlo Code
Peri, Eyal; Orion, Itzhak
2017-09-01
High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.
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)
Botta, F; Mairani, A; Battistoni, G; Cremonesi, M; Di Dia, A; Fassò, A; Ferrari, A; Ferrari, M; Paganelli, G; Pedroli, G; Valente, M
2011-07-01
The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy (89Sr, 90Y, 131I 153Sm, 177Lu, 186Re, and 188Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. FLUKA outcomes have been compared to PENELOPE v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (ETRAN, GEANT4, MCNPX) has been done. Maximum percentage differences within 0.8.RCSDA and 0.9.RCSDA for monoenergetic electrons (RCSDA being the continuous slowing down approximation range) and within 0.8.X90 and 0.9.X90 for isotopes (X90 being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9.RCSDA and 0.9.X90 for electrons and isotopes, respectively. Concerning monoenergetic electrons, within 0.8.RCSDA (where 90%-97% of the particle energy is deposed), FLUKA and PENELOPE agree mostly within 7%, except for 10 and 20 keV electrons (12% in water, 8.3% in bone). The
Generalized perturbation theory in DRAGON: application to CANDU cell calculations
International Nuclear Information System (INIS)
Courau, T.; Marleau, G.
2001-01-01
Generalized perturbation theory (GPT) in neutron transport is a means to evaluate eigenvalue and reaction rate variations due to small changes in the reactor properties (macroscopic cross sections). These variations can be decomposed in two terms: a direct term corresponding to the changes in the cross section themselves and an indirect term that takes into account the perturbations in the neutron flux. As we will show, taking into account the indirect term using a GPT method is generally straight forward since this term is the scalar product of the unperturbed generalized adjoint with the product of the variation of the transport operator and the unperturbed flux. In the case where the collision probability (CP) method is used to solve the transport equation, evaluating the perturbed transport operator involves calculating the variations in the CP matrix for each change in the reactor properties. Because most of the computational effort is dedicated to the CP matrix calculation the gains expected form the GPT method would therefore be annihilated. Here we will present a technique to approximate the variations in the CP matrices thereby replacing the variations in the transport operator with source term variations. We will show that this approximation yields errors fully compatible with the standard generalized perturbation theory errors. Results for 2D CANDU cell calculations will be presented. (author)
Comparison between HELIOS calculations and a PWR cell benchmark for actinides transmutation
Energy Technology Data Exchange (ETDEWEB)
Guzman, Rafael [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Francois, Juan-Luis [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico)]. E-mail: jlfl@fi-b.unam.mx
2007-01-15
This paper shows a comparison between the results obtained with the HELIOS code and other similar codes used in the international community, with respect to the transmutation of actinides. To do this, the international benchmark: 'Calculations of Different Transmutation Concepts' of the Nuclear Energy Agency is analyzed. In this benchmark, two types of cells are analyzed: a small cell corresponding to a standard pressurized water reactor (PWR), and a wide cell corresponding to a highly moderated PWR. Two types of discharge burnup are considered: 33 GWd/tHM and 50 GWd/tHM. The following results are analyzed: the neutron multiplication factor as a function of burnup, the atomic density of the principal actinide isotopes, the radioactivity of selected actinides at reactor shutdown and cooling times from 7 until 50,000 years, the void reactivity and the Doppler reactivity. The results are compared with the following codes: KAPROS/KARBUS (FZK, Germany), SRAC95 (JAERI, Japan), TRIFON (ITTEP, Russian Federation) and WIMS (IPPE, Russian Federation). For the neutron multiplication factor, the results obtained with HELIOS show a difference of around 1% {delta}k/k. For the isotopic concentrations: {sup 241}Pu, {sup 242}Pu, and {sup 242m}Am, the results of all the institutions present a difference that increases at higher burnup; for the case of {sup 237}Np, the results of FZK diverges from the other results as the burnup increases. Regarding the activity, the difference of the results is acceptable, except for the case of {sup 241}Pu. For the Doppler coefficient, the results are acceptable, except for the cells with high moderation. In the case of the void coefficient, the difference of the results increases at higher void fractions, being the highest at 95%. In summary, for the PWR benchmark, the results obtained with HELIOS agree reasonably well within the limits of the multiple plutonium recycling established by the NEA working party on plutonium fuels and
Energy Technology Data Exchange (ETDEWEB)
Romero, L.; Travesi, A.
1983-07-01
A codes, BETAL, was developed, written in FORTRAN IV, to automatize calculations and presentations of the result of the total alpha-beta activities measurements in environmental samples. This code performs the necessary calculations for transformation the activities measured in total counts, to pCi/1., bearing in mind the efficiency of the detector used and the other necessary parameters. Further more, it appraise the standard deviation of the result, and calculus the Lower limit of detection for each measurement. This code is written in iterative way by screen-operator dialogue, and asking the necessary data to perform the calculation of the activity in each case by a screen label. The code could be executed through any screen and keyboard terminal, (whose computer accepts Fortran IV) with a printer connected to the said computer. (Author) 5 refs.
International Nuclear Information System (INIS)
Srinivasan, M.G.; Kot, C.A.; Hsieh, B.J.
1991-09-01
The objective of this effort was to evaluate the piping analysis part of the SMACS code for estimating the response of realistic piping systems subjected to seismic excitation, given as multiple, independent support acceleration histories. The experimental data from the seismic testing of an in-plant piping system at the HDR Test Facility in Germany were used for this purpose. Of the six different support systems tested, two were selected for the evaluation: one a ''stiff'' configuration containing both struts and snubbers, and the other, a more flexible configuration with no snubbers. Described are the analytical modeling, calculations, and results of the posttest simulation of two sets each for both support configurations, with excitations at 100% and 200/300% of safe-shutdown-earthquake loading. Almost all the calculated peak response quantities were smaller (by different amounts) than the corresponding test measurements. However, pipe displacements and bending stresses were better estimated than the pipe accelerations and support forces. The discrepancies are mainly attributable to the inability of the linear analysis to model the nonlinear behavior of the VKL piping system, characterized by gaps in support connections and the friction at pipe clamps. A similar trend of underestimating test responses was observed in the linear analyses performed by other investigators. 13 refs., 14 figs., 8 tabs
International Nuclear Information System (INIS)
Son, Young-Seok; Shin, Jee-Young; Lim, Ho-Gon; Park, Jin-Hee; Jang, Seung-Cheol
2005-01-01
The methods developed for full-power probabilistic safety assessment, including thermal-hydraulic methods, have been widely applied to low power and shutdown conditions. Experience from current low power and shutdown probabilistic safety assessments, however, indicates that the thermal-hydraulic methods developed for full-power probabilistic safety assessments are not always reliable when applied to low power and shutdown conditions and consequently may yield misleading and inaccurate risk insights. To increase the usefulness of the low power and shutdown risk insights, the current methods and tools used for thermal-hydraulic calculations should be examined to ascertain whether they function effectively for low power and shutdown conditions. In this study, a platform for relatively detailed thermal-hydraulic calculations applied to low power and shutdown conditions in a pressurized water reactor was developed based on the best estimate thermal-hydraulic analysis code, MARS2.1. To confirm the applicability of the MARS platform to low power and shutdown conditions, many thermal-hydraulic analyses were performed for the selected topic, i.e. the loss of shutdown cooling events for various plant operating states at the Korean standard nuclear power plant. The platform developed in this study can deal effectively with low power and shutdown conditions, as well as assist the accident sequence analysis in low power and shutdown probabilistic safety assessments by providing fundamental data. Consequently, the resulting analyses may yield more realistic and accurate low power and shutdown risk insights
International Nuclear Information System (INIS)
Ding, Aiping; Liu, Tianyu; Liang, Chao; Ji, Wei; Shephard, Mark S.; Xu, X George; Brown, Forrest B.
2011-01-01
Monte Carlo simulation is ideally suited for solving Boltzmann neutron transport equation in inhomogeneous media. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop system. The interest in adopting GPUs for Monte Carlo acceleration is rapidly mounting, fueled partially by the parallelism afforded by the latest GPU technologies and the challenge to perform full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem and an eigenvalue/criticality problem were developed for CPU and GPU environments, respectively, to evaluate issues associated with computational speedup afforded by the use of GPUs. The results suggest that a speedup factor of 30 in Monte Carlo radiation transport of neutrons is within reach using the state-of-the-art GPU technologies. However, for the eigenvalue/criticality problem, the speedup was 8.5. In comparison, for a task of voxelizing unstructured mesh geometry that is more parallel in nature, the speedup of 45 was obtained. It was observed that, to date, most attempts to adopt GPUs for Monte Carlo acceleration were based on naïve implementations and have not yielded the level of anticipated gains. Successful implementation of Monte Carlo schemes for GPUs will likely require the development of an entirely new code. Given the prediction that future-generation GPU products will likely bring exponentially improved computing power and performances, innovative hardware and software solutions may make it possible to achieve full-core Monte Carlo calculation within one hour using a desktop computer system in a few years. (author)
Sabouri, P.; Bidaud, A.; Dabiran, S.; Lecarpentier, D.; Ferragut, F.
2014-04-01
The development of tools for nuclear data uncertainty propagation in lattice calculations are presented. The Total Monte Carlo method and the Generalized Perturbation Theory method are used with the code DRAGON to allow propagation of nuclear data uncertainties in transport calculations. Both methods begin the propagation of uncertainties at the most elementary level of the transport calculation - the Evaluated Nuclear Data File. The developed tools are applied to provide estimates for response uncertainties of a PWR cell as a function of burnup.
Energy Technology Data Exchange (ETDEWEB)
Mazurier, J
1999-05-28
This thesis has been performed in the framework of national reference setting-up for absorbed dose in water and high energy photon beam provided with the SATURNE-43 medical accelerator of the BNM-LPRI (acronym for National Bureau of Metrology and Primary standard laboratory of ionising radiation). The aim of this work has been to develop and validate different user codes, based on PENELOPE Monte Carlo code system, to determine the photon beam characteristics and calculate the correction factors of reference dosimeters such as Fricke dosimeters and graphite calorimeter. In the first step, the developed user codes have permitted the influence study of different components constituting the irradiation head. Variance reduction techniques have been used to reduce the calculation time. The phase space has been calculated for 6, 12 and 25 MV at the output surface level of the accelerator head, then used for calculating energy spectra and dose distributions in the reference water phantom. Results obtained have been compared with experimental measurements. The second step has been devoted to develop an user code allowing calculation correction factors associated with both BNM-LPRI's graphite and Fricke dosimeters thanks to a correlated sampling method starting with energy spectra obtained in the first step. Then the calculated correction factors have been compared with experimental and calculated results obtained with the Monte Carlo EGS4 code system. The good agreement, between experimental and calculated results, leads to validate simulations performed with the PENELOPE code system. (author)
International Nuclear Information System (INIS)
Gomes, Renato G.; Rebello, Wilson F.; Vellozo, Sergio O.; Moreira Junior, Luis; Vital, Helio C.; Rusin, Tiago; Silva, Ademir X.
2013-01-01
In order to evaluate new lines of research in the area of irradiation of materials external to the research irradiating facility Army Technology Center (CTEx), it is necessary to study security parameters and magnitude of the dose rates from their channels of escape. The objective was to calculate, with the code MCNPX, dose rates (Gy / min) on the interior and exterior of the four-channel leakage gamma irradiator. The channels were designed to leak radiation on materials properly disposed in the area outside the irradiator larger than the expected volume of irradiation chambers (50 liters). This study aims to assess the magnitude of dose rates within the channels, as well as calculate the angle of beam output range outside the channel for analysis as to its spread, and evaluation of safe conditions of their operators (protection radiological). The computer simulation was performed by distributing virtual dosimeter ferrous sulfate (Fricke) in the longitudinal axis of the vertical drain channels (anterior and posterior) and horizontal (top and bottom). The results showed a collimating the beams irradiated on each of the channels to the outside, with values of the order of tenths of Gy / min as compared to the maximum amount of operation of the irradiator chamber (33 Gy / min). The external beam irradiation in two vertical channels showed a distribution shaped 'trunk pyramid', not collimated, so scattered, opening angle 83 ° in the longitudinal direction and 88 in the transverse direction. Thus, the cases allowed the evaluation of materials for irradiation outside the radiator in terms of the magnitude of the dose rates and positioning of materials, and still be able to take the necessary care in mounting shield for radiation protection by operators, avoiding exposure to ionizing radiation. (author)
International Nuclear Information System (INIS)
Segarceanu, A. M.; Valeca, S. C.; Constantin, M.
2016-01-01
The void coefficient of the reactivity in a typical CANDU cell is positive and it represents a sensitive issue for the regulatory discussions. The paper investigates the influence of some parameters (as geometry, composition, temperatures) on the void reactivity and power distributions. The methodology is based on integral transport calculation by the first flight collision probability approach (WIMS and CP 2 D codes). The results offer a general view regarding the importance of different parameters and physical phenomena on the magnitude of the void reactivity, useful for educational & training purpose, and for the future development of new fuel bundles. The influence of each parameter is separately treated in order to understand the individual importance in the global effects. (authors)
International Nuclear Information System (INIS)
Capote, R.; Lopez, R.; Herrera, E.; Piris, M.; Osorio, V.
1991-01-01
A new PC code PCROSS for neutron induced reaction calculations up to 25 MeV incident was developed, where the latest theoretical development in the model was employed. A combination of exciton model and multistep direct reaction model parametrization was used in order to describe the high energy part of the spectra. In the PCROSS code several models for level density calculations are available. The code includes a subroutine to generate the input data. In the present paper some calculation results for (n,n') and (n,p) emission spectra in the range of 5 to 25 MeV and for (n,p) and (n,2n) excitation functions up to 20 MeV are shown. A good description of the experimental data was achieved. (author). 32 refs, 3 figs
Development of 2D particle-in-cell code to simulate high current, low ...
Indian Academy of Sciences (India)
Abstract. A code for 2D space-charge dominated beam dynamics study in beam trans- port lines is developed. The code is used for particle-in-cell (PIC) simulation of z-uniform beam in a channel containing solenoids and drift space. It can also simulate a transport line where quadrupoles are used for focusing the beam.
Pacilio, M; Lanconelli, N; Lo, Meo S; Betti, M; Montani, L; Torres, Aroche L A; Coca, Pérez M A
2009-05-01
Several updated Monte Carlo (MC) codes are available to perform calculations of voxel S values for radionuclide targeted therapy. The aim of this work is to analyze the differences in the calculations obtained by different MC codes and their impact on absorbed dose evaluations performed by voxel dosimetry. Voxel S values for monoenergetic sources (electrons and photons) and different radionuclides (90Y, 131I, and 188Re) were calculated. Simulations were performed in soft tissue. Three general-purpose MC codes were employed for simulating radiation transport: MCNP4C, EGSnrc, and GEANT4. The data published by the MIRD Committee in Pamphlet No. 17, obtained with the EGS4 MC code, were also included in the comparisons. The impact of the differences (in terms of voxel S values) among the MC codes was also studied by convolution calculations of the absorbed dose in a volume of interest. For uniform activity distribution of a given radionuclide, dose calculations were performed on spherical and elliptical volumes, varying the mass from 1 to 500 g. For simulations with monochromatic sources, differences for self-irradiation voxel S values were mostly confined within 10% for both photons and electrons, but with electron energy less than 500 keV, the voxel S values referred to the first neighbor voxels showed large differences (up to 130%, with respect to EGSnrc) among the updated MC codes. For radionuclide simulations, noticeable differences arose in voxel S values, especially in the bremsstrahlung tails, or when a high contribution from electrons with energy of less than 500 keV is involved. In particular, for 90Y the updated codes showed a remarkable divergence in the bremsstrahlung region (up to about 90% in terms of voxel S values) with respect to the EGS4 code. Further, variations were observed up to about 30%, for small source-target voxel distances, when low-energy electrons cover an important part of the emission spectrum of the radionuclide (in our case, for 131I
International Nuclear Information System (INIS)
Kawasaki, Nobuchika; Asayama, Tai
2001-09-01
Both reliability and safety have to be further improved for the successful commercialization of FBRs. At the same time, construction and operation costs need to be reduced to a same level of future LWRs. To realize compatibility among reliability, safety and, cost, the Structural Mechanics Research Group in JNC started the development of System Based Code for Integrity of FBR. This code extends the present structural design standard to include the areas of fabrication, installation, plant system design, safety design, operation and maintenance, and so on. A quantitative index is necessary to connect different partial standards in this code. Failure probability is considered as a candidate index. Therefore we decided to make a model calculation using failure probability and judge its applicability. We first investigated other probabilistic standards like ASME Code Case N-578. A probabilistic approach in the structural integrity evaluation was created based on these results, and also an evaluation flow was proposed. According to this flow, a model calculation of creep-fatigue damage was performed. This trial calculation was for a vessel in a sodium-cooled FBR. As the result of this model calculation, a crack initiation probability and a crack penetration probability were found to be effective indices. Last we discussed merits of this System Based Code, which are presented in this report. Furthermore, this report presents future development tasks. (author)
Energy Technology Data Exchange (ETDEWEB)
Artmann, Andreas; Meyering, Henrich
2016-11-15
The GRS research project was aimed to the test the appropriateness of the software package ''residual radioactivity'' (RESRAD) for the calculation of clearance values according to German and European regulations. Comparative evaluations were performed with RESRAD-OFFSITE and the code SiWa-PRO DSS used by GRS and the GRS program code ARTM. It is recommended to use RESRAD-OFFSITE for comparative calculations. The dose relevant air-path dispersion of radionuclides should not be modeled using RESRAD-OFFSITE, the use of ARTM is recommended. The sensitivity analysis integrated into RESRAD-OFFSITE allows a fast identification of crucial parameters.
International Nuclear Information System (INIS)
Sugimoto, Osamu; Sawaguchi, Yusuke; Kaneko, Masahito.
1979-01-01
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 γ-exposure at an observation point can be calculated within a few minutes with comparable precision to other existing codes. (author)
The CORSYS neutronics code system
International Nuclear Information System (INIS)
Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.
1994-01-01
The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs
Extracellular vesicle associated long non-coding RNAs functionally enhance cell viability
Directory of Open Access Journals (Sweden)
Chris Hewson
2016-10-01
Full Text Available Cells communicate with one another to create microenvironments and share resources. One avenue by which cells communicate is through the action of exosomes. Exosomes are extracellular vesicles that are released by one cell and taken up by neighbouring cells. But how exosomes instigate communication between cells has remained largely unknown. We present evidence here that particular long non-coding RNA molecules are preferentially packaged into exosomes. We also find that a specific class of these exosome associated non-coding RNAs functionally modulate cell viability by direct interactions with l-lactate dehydrogenase B (LDHB, high-mobility group protein 17 (HMG-17, and CSF2RB, proteins involved in metabolism, nucleosomal architecture and cell signalling respectively. Knowledge of this endogenous cell to cell pathway, those proteins interacting with exosome associated non-coding transcripts and their interacting domains, could lead to a better understanding of not only cell to cell interactions but also the development of exosome targeted approaches in patient specific cell-based therapies. Keywords: Non-coding RNA, Extracellular RNA, Exosomes, Retroelement, Pseudogene
International Nuclear Information System (INIS)
Toda, Shin-ichi; Yoshikawa, Shinji; Oketani, Kazuhiro
2003-05-01
The improved version of the MSG code (Multi-dimensional Thermal-hydraulic Analysis Code for Steam Generators) has been released. It has been carried out to improve based on the original version in order to calculate reverse flow on water/steam side, and to animate the post-processing data. To calculate reverse flow locally, modification to set pressure at each divided node point of water/steam region in the helical-coil heat transfer tubes has been carried out. And the matrix solver has been also improved to treat a problem within practical calculation time against increasing the pressure points. In this case pressure and enthalpy have to be calculated simultaneously, however, it was found out that using the block-Jacobean method make a diagonal-dominant matrix, and solve the matrix efficiently with a relaxation method. As the result of calculations of a steady-state condition and a transient of SG blow down with manual trip operation, the improvement on calculation function of the MSG code was confirmed. And an animation function of temperature contour in the sodium shell side as a post processing has been added. Since the animation is very effective to understand thermal-hydraulic behavior on the sodium shell side of the SG, especially in case of transient condition, the analysis and evaluation of the calculation results will be enabled to be more quickly and effectively. (author)
Energy Technology Data Exchange (ETDEWEB)
Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik
2014-06-15
In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.
International Nuclear Information System (INIS)
Khattab, K.; Bush, M; Kassery, H.
2009-03-01
A 3-D model for the irradiation plant which belongs to the Atomic Energy Commission, Department of Radiation Technology in the Deir Al-Hajar area near Damascus, is presented in this work using the MCNP-4C code. This model is used to calculate the spatial gamma ray dose in the (x, y, z) coordinate. Good agreements are noticed between the measured and the calculated results. (author)
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 Place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie - Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO, SIS/GAM, 6, Avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
Two French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. Development of analytical methods has been made for the last 10 years through a collaboration between CEA, EDF and AREVA-NP, and through R and D actions involving CEA and IRSN. These activities have led to unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of papers is composed of five parts: the first presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Part V presents validation of the methods, with details on their accuracy. This paper presents the stress intensity factor and J calculation for cracked elbows. General data applicable for all defect geometries are first presented, and then, compendia for K{sub I} and {sigma}{sub ref} calculations are provided for the available defect geometries.
International Nuclear Information System (INIS)
Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.
2007-01-01
Two French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. Development of analytical methods has been made for the last 10 years through a collaboration between CEA, EDF and AREVA-NP, and through R and D actions involving CEA and IRSN. These activities have led to unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of papers is composed of five parts: the first presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Part V presents validation of the methods, with details on their accuracy. This paper presents the stress intensity factor and J calculation for cracked elbows. General data applicable for all defect geometries are first presented, and then, compendia for K I and σ ref calculations are provided for the available defect geometries
KIM, Jong Woon; LEE, Young-Ouk
2017-09-01
As computing power gets better and better, computer codes that use a deterministic method seem to be less useful than those using the Monte Carlo method. In addition, users do not like to think about space, angles, and energy discretization for deterministic codes. However, a deterministic method is still powerful in that we can obtain a solution of the flux throughout the problem, particularly as when particles can barely penetrate, such as in a deep penetration problem with small detection volumes. Recently, a new state-of-the-art discrete-ordinates code, ATTILA, was developed and has been widely used in several applications. ATTILA provides the capabilities to solve geometrically complex 3-D transport problems by using an unstructured tetrahedral mesh. Since 2009, we have been developing our own code by benchmarking ATTILA. AETIUS is a discrete ordinates code that uses an unstructured tetrahedral mesh such as ATTILA. For pre- and post- processing, Gmsh is used to generate an unstructured tetrahedral mesh by importing a CAD file (*.step) and visualizing the calculation results of AETIUS. Using a CAD tool, the geometry can be modeled very easily. In this paper, we describe a brief overview of AETIUS and provide numerical results from both AETIUS and a Monte Carlo code, MCNP5, in a deep penetration problem with small detection volumes. The results demonstrate the effectiveness and efficiency of AETIUS for such calculations.
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki
2005-06-01
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)
A thermo-mechanical benchmark calculation of an hexagonal can in the BTI accident with ABAQUS code
International Nuclear Information System (INIS)
Zucchini, A.
1988-07-01
The thermo-mechanical behaviour of an hexagonal can in a benchmark problem (simulating the conditions of a BTI accident in a fuel assembly) is examined by means of the ABAQUS code: the effects of the geometric nonlinearity are shown and the results are compared with those of a previous analysis performed with the INCA code. (author)
Energy Technology Data Exchange (ETDEWEB)
Willmann, P.A.; Hooper, E.B. Jr.
1977-02-01
A computer program was written to calculate the stored energy in a transformer. This result easily yields the inductance and leakage reactance of the transformer and is estimated to be accurate to better than 5 percent. The program was used to calculate the leakage reactance of the main transformer for the LLL neutral beam High Voltage Test Stand.
International Nuclear Information System (INIS)
Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.
2010-10-01
Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)
Energy Technology Data Exchange (ETDEWEB)
Jan, S. [CEA Direction des Sciences du Vivant, Institut d ' Imagerie Bio-Medicale, Service Hospitalier Frederic Joliot, 4 pl. du Gn. Leclerc 91401 Orsay Cedex (France)
2010-07-01
The author presents the GATE code, a simulation software based on the Geant4 development environment developed by the CERN (the European organization for nuclear research) which enables Monte-Carlo type simulation to be developed for tomography imagery using ionizing radiation, and radiotherapy examinations (conventional and hadron therapy) to be simulated. The authors concentrate on the use of medical imagery in carcinology. They comment some results obtained in nuclear imagery and in radiotherapy
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.
International Nuclear Information System (INIS)
Alvarez R, J.T.
1991-01-01
In this work the implementation of a modification of the VARSKIN code for calculation of absorbed dose for contamination in skin imparted by external radiation fields generated by Beta emitting is presented. The modification consists on the inclusion of 47 isotopes of interest even Nuclear Plants for the dose evaluation in skin generated by 'hot particles'. The approach for to add these isotopes is the correlation parameter F and the average energy of the Beta particle, with relationship to those 75 isotopes of the original code. The methodology of the dose calculation of the VARSKIN code is based on the interpolation, (and integration of the interest geometries: punctual or plane sources), of the distribution functions scaled doses in water for beta and electrons punctual sources, tabulated by Berger. Finally a brief discussion of the results for their interpretation and use with purposes of radiological protection (dose insurance in relation to the considered biological effects) is presented
Lopez-Piñeiro, A.; Sanchez, M. L.; Moreno, B.
1992-06-01
The computer program MORSMATEL has been developed to calculate vibrational-rotational matrix elements of several r-dependent operators of two Morse oscillators. This code is based on a set of recurrence relations which are valid for any value of the power and of the quantum numbers v and J of each oscillator.
Energy Technology Data Exchange (ETDEWEB)
Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)
2016-09-15
The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.
International Nuclear Information System (INIS)
Antunes, Alberi
2008-01-01
This work presents the Physics of Source Driven Systems (ADS). It shows some statics and K i netics parameters of the reactor Physics and when it is sub critical, that are important in evaluation and definition of these systems. The objective is to demonstrate that there are differences in parameters when the reactor is critical. Moreover, the work shows the differences observed in the parameters for different calculation models. Two calculation methodologies are shown In this dissertation: Gandini and Salvatores and Dulla, and some parameters are calculated. The ANISN deterministic transport code is used in calculation in order to compare these parameters. In a subcritical configuration of IPEN-MB-01 Reactor driven by an external source some parameters are calculated. The conclusions about calculation realized are presented in end of work. (author)
International Nuclear Information System (INIS)
Strickert, R.; Friedman, A.M.; Fried, S.
1979-04-01
A computer program (ARDISC, the Argonne Dispersion Code) is described which simulates the migration of nuclides in porous media and includes first order kinetic effects on the retention constants. The code allows for different absorption and desorption rates and solves the coupled migration equations by arithmetic reiterations. Input data needed are the absorption and desorption rates, equilibrium surface absorption coefficients, flow rates and volumes, and media porosities
Development of 2D particle-in-cell code to simulate high current, low ...
Indian Academy of Sciences (India)
Home; Journals; Pramana – Journal of Physics; Volume 69; Issue 4. Development of 2D particle-in-cell code to simulate high current, low energy beam in a beam transport system. S C L Srivastava S V L S Rao ... Keywords. Space-charge; particle-in-cell; beam dynamics; Poisson's equation; solenoids; quadrupole magnets.
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa
2017-03-01
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)
International Nuclear Information System (INIS)
Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta
2003-01-01
All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important
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.
International Nuclear Information System (INIS)
Khan, M.J.H.; Sarker, M.M.; Islam, S.M.A.
2013-01-01
Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations
Energy Technology Data Exchange (ETDEWEB)
Gonzalez, Alejandro; Milian, Daniel [Instituto Superior de Ciencias y Tecnologias Nucleares (ISCTN), La Habana (Cuba). E-mail: agg@ctn.isctn.edu.cu
2000-07-01
The task of the physical calculation of the reactor demand of the management of a great volume of information and inclose the stages for processing of data, calculations and analysis of their results. These stages are highly sensible to human mistakes, that's why is imprescindible that them undergo automatization, doing tracked all the process against mistake or unexpected result. The user-interface SYSMOD was developed over the platform IDE Delphi 3.0, visual language driven to events. It to consist in of the principal menu, which inclose between its options the preparation of the input data (File and Edit) to the pre-processors for the calculation codes of reactors. The output information may be showed in graphic and/or alphanumeric format (Data-Process). SYSMOD endures two applications for the management of the data base for the data during the preparation of the input for the pre-processors of the spectral calculation, so as for the organization, conservation and presentation for the obtained results. The carried out of the lattices and global codes, takes place from this application, over the platform MS-DOS (Run). SYSMOD regards the possibility for the debugging of the codes (Debugging), so as the benchmarks qualified to so effect (Benchmark). SYSMOD has been applied for the analysis of te WWER-440 of the first unity of Juragua Nuclear Power Plant. (author)
International Nuclear Information System (INIS)
Kedziur, F.
1982-07-01
The computer code CALIPSO was developed for the calculation of a hypothetical accident in an LMFBR (Liquid Metal Fast Breeder Reactor), where the failure of fuel pins is assumed. It calculates two-dimensionally the thermodynamics, fluiddynamics and changes in geometry of a single fuel pin and its coolant channel in a time period between failure of the pin and a state, at which the geometry is nearly destroyed. The determination of temperature profiles in the fuel pin cladding and the channel wall make it possible to take melting and freezing processes into account. Further features of CALIPSO are the variable channel cross section in order to model disturbances of the channel geometry as well as the calculation of two velocity fields including the consideration of virtual mass effects. The documented version of CALIPSO is especially suited for the calculation of the SIMBATH experiments carried out at the Kernforschungszentrum Karlsruhe, which simulate the above-mentioned accident. The report contains the complete documentation of the CALIPSO code: the modeling of the geometry, the equations used, the structure of the code and the solution procedure as well as the instructions for use with an application example. (orig.) [de
Uncovering the roles of long non-coding RNAs in cancer stem cells
Directory of Open Access Journals (Sweden)
Xiaoxing Huang
2017-02-01
Full Text Available Abstract Cancer has been a major public health problem that has threatened human life worldwide throughout history. The main causes that contribute to the poor prognosis of cancer are metastasis and recurrence. Cancer stem cells are a group of tumor cells that possess self-renewal and differentiation ability, which is a vital cause of cancer metastasis and recurrence. Long non-coding RNAs refer to a class of RNAs that are longer than 200 nt and have no potential to code proteins, some of which can be specifically expressed in different tissues and different tumors. Long non-coding RNAs have great biological significance in the occurrence and progression of cancers. However, how long non-coding RNAs interact with cancer stem cells and then affect cancer metastasis and recurrence is not yet clear. Therefore, this review aims to summarize recent studies that focus on how long non-coding RNAs impact tumor occurrence and progression by affecting cancer stem cell self-renewal and differentiation in liver cancer, prostate cancer, breast cancer, and glioma.
The curvature calculation mechanism based on simple cell model.
Yu, Haiyang; Fan, Xingyu; Song, Aiqi
2017-07-20
A conclusion has not yet been reached on how exactly the human visual system detects curvature. This paper demonstrates how orientation-selective simple cells can be used to construct curvature-detecting neural units. Through fixed arrangements, multiple plurality cells were constructed to simulate curvature cells with a proportional output to their curvature. In addition, this paper offers a solution to the problem of narrow detection range under fixed resolution by selecting an output value under multiple resolution. Curvature cells can be treated as concrete models of an end-stopped mechanism, and they can be used to further understand "curvature-selective" characteristics and to explain basic psychophysical findings and perceptual phenomena in current studies.
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan); Mochizuki, Hiroki [The Japan Research Institute Ltd., Tokyo (Japan)
2001-11-01
For providing conservative PWR spent fuel compositions from the view point of nuclear criticality safety, correction factors applicable for result of burnup calculation by ORIGEN2 were evaluated. Its conservativeness was verified by criticality calculations using MVP. To calculate these correction factors, analyses of spent fuel isotopic composition data were performed by ORIGEN2. Maximum or minimum value of the ratio of calculation result to experimental data was chosen as correction factor. These factors are given to each set of fuel assembly and ORIGEN2 library. They could be considered as the re-definition of recommended isotopic composition given in Nuclear Criticality Safety Handbook. (author)
International Nuclear Information System (INIS)
Bencik, M.; Hadek, J.
2011-01-01
The paper gives a brief survey of the seventh three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at Nuclear Research Institute Rez. This benchmark was defined at the twentieth AER Symposium in Hanassari (Finland). It is focused on investigation of transient behaviour in a WWER-440 nuclear power plant. Its initiating event is opening of the main isolation valve and re-connection of the loop with its main circulation pump in operation. The WWER-440 plant is at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations were performed with the code DYN3D. Transient calculation was made with the system code RELAP5-3D. The two-group homogenized cross sections library HELGD05 created by HELIOS code was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP Dukovany was adopted for the seventh AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was coupled with 49 core thermal-hydraulic channels and 8 reflector channels connected with the three-dimensional model of the reactor vessel. The detailed nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5-3D system code and a short description of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters. (Authors)
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie-Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO SIS/GAM, 6, avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for the development and evaluation of the accuracy of the proposed analytical methods. This second article in the series presents all details for the stress intensity factor and J calculations for cracked plates. General data applicable for all defect geometries are first presented, and then, available defect geometries where compendia for K{sub I} and {sigma}{sub ref} calculation are provided are given.
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie-Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO SIS/GAM, 6, avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, Avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high-temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress-intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of the RSE-M and in the 2007 edition of the RCC-MR. This series of articles is composed of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Part V presents validation, with details on the accuracy of the proposed analytical method. This third part in the series presents details of the stress intensity factor and J calculations for cracked pipes. General data applicable for all defect geometries are first presented, and then, compendia for K{sub I} and {sigma}{sub ref} calculations are provided for specific cases.
International Nuclear Information System (INIS)
Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.
2007-01-01
French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for the development and evaluation of the accuracy of the proposed analytical methods. This second article in the series presents all details for the stress intensity factor and J calculations for cracked plates. General data applicable for all defect geometries are first presented, and then, available defect geometries where compendia for K I and σ ref calculation are provided are given
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie-Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO SIS/GAM, 6, avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
Two French nuclear codes include flaw assessment procedures: the RSE-M code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of five parts: this first one presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components: plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for their development and on evaluation of the accuracy of the proposed analytical methods. This first article of the series presents an overview of the calculation of K{sub I} and J in these two codes and describes briefly the defect assessment analyses. Specific details in the Appendix A16 of RCC-MR (LBB procedure and creep analyses) are also introduced in this article.
International Nuclear Information System (INIS)
Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.
2007-01-01
Two French nuclear codes include flaw assessment procedures: the RSE-M code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of five parts: this first one presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components: plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for their development and on evaluation of the accuracy of the proposed analytical methods. This first article of the series presents an overview of the calculation of K I and J in these two codes and describes briefly the defect assessment analyses. Specific details in the Appendix A16 of RCC-MR (LBB procedure and creep analyses) are also introduced in this article
Energy Technology Data Exchange (ETDEWEB)
Dellin, T.A.; Fish, M.J.; Yang, C.L.
1981-08-01
DELSOL2 is a revised and substantially extended version of the DELSOL computer program for calculating collector field performance and layout, and optimal system design for solar thermal central receiver plants. The code consists of a detailed model of the optical performance, a simpler model of the non-optical performance, an algorithm for field layout, and a searching algorithm to find the best system design. The latter two features are coupled to a cost model of central receiver components and an economic model for calculating energy costs. The code can handle flat, focused and/or canted heliostats, and external cylindrical, multi-aperture cavity, and flat plate receivers. The program optimizes the tower height, receiver size, field layout, heliostat spacings, and tower position at user specified power levels subject to flux limits on the receiver and land constraints for field layout. The advantages of speed and accuracy characteristic of Version I are maintained in DELSOL2.
Energy Technology Data Exchange (ETDEWEB)
Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology
1997-12-31
At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.
Methods for Using Small Non-Coding RNAs to Improve Recombinant Protein Expression in Mammalian Cells
Directory of Open Access Journals (Sweden)
Sarah Inwood
2018-01-01
Full Text Available The ability to produce recombinant proteins by utilizing different “cell factories” revolutionized the biotherapeutic and pharmaceutical industry. Chinese hamster ovary (CHO cells are the dominant industrial producer, especially for antibodies. Human embryonic kidney cells (HEK, while not being as widely used as CHO cells, are used where CHO cells are unable to meet the needs for expression, such as growth factors. Therefore, improving recombinant protein expression from mammalian cells is a priority, and continuing effort is being devoted to this topic. Non-coding RNAs are RNA segments that are not translated into a protein and often have a regulatory role. Since their discovery, major progress has been made towards understanding their functions. Non-coding RNA has been investigated extensively in relation to disease, especially cancer, and recently they have also been used as a method for engineering cells to improve their protein expression capability. In this review, we provide information about methods used to identify non-coding RNAs with the potential of improving recombinant protein expression in mammalian cell lines.
International Nuclear Information System (INIS)
Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.
1993-08-01
The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie-Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO, SIS/GAM, 6, avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction Rules for Mechanical Components of FBR Nuclear Islands and High Temperature Applications'. Development of analytical methods has been made for the last 10 years in the framework of a collaboration between CEA, EDF and AREVA-NP, and by R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, in particular the stress intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of articles consists of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide the compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). This part presents validation of the methods, with details on the process followed for their development and of the evaluation accuracy of the proposed analytical methods.
International Nuclear Information System (INIS)
Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.
2007-01-01
French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction Rules for Mechanical Components of FBR Nuclear Islands and High Temperature Applications'. Development of analytical methods has been made for the last 10 years in the framework of a collaboration between CEA, EDF and AREVA-NP, and by R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of articles consists of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide the compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). This part presents validation of the methods, with details on the process followed for their development and of the evaluation accuracy of the proposed analytical methods
International Nuclear Information System (INIS)
Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.
1993-08-01
The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de
Particle-in-Cell Codes for plasma-based particle acceleration
Pukhov, Alexander
2016-01-01
Basic principles of particle-in-cell (PIC ) codes with the main application for plasma-based acceleration are discussed. The ab initio full electromagnetic relativistic PIC codes provide the most reliable description of plasmas. Their properties are considered in detail. Representing the most fundamental model, the full PIC codes are computationally expensive. The plasma-based acceler- ation is a multi-scale problem with very disparate scales. The smallest scale is the laser or plasma wavelength (from one to hundred microns) and the largest scale is the acceleration distance (from a few centimeters to meters or even kilometers). The Lorentz-boost technique allows to reduce the scale disparity at the costs of complicating the simulations and causing unphysical numerical instabilities in the code. Another possibility is to use the quasi-static approxi- mation where the disparate scales are separated analytically.
Directory of Open Access Journals (Sweden)
Jingang Liang
2016-06-01
Full Text Available Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC codes in accomplishing pin-wise three-dimensional (3D full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
International Nuclear Information System (INIS)
Iijima, Toshinori; Shiraishi, Tadao
1979-10-01
For environmental doses from routine releases of LWRs effluents to meet the Criterion 'As Low As is Practicable (ALAP)', Japan Atomic Energy Commission (JAEC) established a series of guides, the first for 'Dose Objectives' (May 1975), the second for models and parameters for calculating the environmental doses to compare with the 'Dose Objectives' (September 1976), and the third providing onsite meteorological programs, statistics of the data obtained and atmospheric dispersion models (June 1977). JAERI has developed a computer code, designated as ANDOSE, for calculating annual releases of radioactive gaseous and liquid effluents and, then, total body doses and thyroid doses to individuals around sites on the basis of these guides. The total body doses are from radioactive noble gases as well as from radioactive materials taken with marine food. For the calculation of thyroid doses are taken into account exposure pathways via inhalation and ingestion of leafy vegetables, cow's milk and marine food. The age-specific thyroid doses are evaluated. The doses are summed up when multisource or multisite conditions need to be evaluated (Nuclear Safety Bureau's requirement). In the present report, are described source-term models, environmental transport models and dose models used in the code, of which most are provided in the guides but some are complemented by the authors, the functions of ANDOSE and the manual for users of the code. The program lists and the latter two guides mentioned above are included in the appendices. (author)
photon-plasma: A modern high-order particle-in-cell code
International Nuclear Information System (INIS)
Haugbølle, Troels; Frederiksen, Jacob Trier; Nordlund, Åke
2013-01-01
We present the photon-plasma code, a modern high order charge conserving particle-in-cell code for simulating relativistic plasmas. The code is using a high order implicit field solver and a novel high order charge conserving interpolation scheme for particle-to-cell interpolation and charge deposition. It includes powerful diagnostics tools with on-the-fly particle tracking, synthetic spectra integration, 2D volume slicing, and a new method to correctly account for radiative cooling in the simulations. A robust technique for imposing (time-dependent) particle and field fluxes on the boundaries is also presented. Using a hybrid OpenMP and MPI approach, the code scales efficiently from 8 to more than 250.000 cores with almost linear weak scaling on a range of architectures. The code is tested with the classical benchmarks particle heating, cold beam instability, and two-stream instability. We also present particle-in-cell simulations of the Kelvin-Helmholtz instability, and new results on radiative collisionless shocks
Energy Technology Data Exchange (ETDEWEB)
Beckert, C.
2007-12-19
Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed. [German] Standardmaessig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte fuer
Energy Technology Data Exchange (ETDEWEB)
Kodeli, I.; Sartori, E. [Organization for Econimic Co-Operation and Development (OECD NEA DB), 91 - Issy les Moulineaux (France)
2003-07-01
The Nuclear energy agency is a specialised agency of OECD (organization economic co-operation and development). These missions are to help its members to keep and improve by international cooperation, the scientific, technological and legal bases necessary to a peaceful use of nuclear energy. Nea includes twenty eight countries. Nea works in collaboration with IAEA. The field of activities concerns the acquisition, validation and distribution of nuclear data, calculation codes and experiments. To help users, it organises conferences and training about the calculation codes that it shares out. (N.C.)
International Nuclear Information System (INIS)
Niklaus, F.; Korteniemi, V.
1996-01-01
At the Department of Energy Technology at Lappeenranta University of Technology a CATHARE model of one unit of the St. Petersburg (RBMK) nuclear power plant has been generated. The investigations have been done in order to understand better the thermal-hydraulic behaviour of RBMK type reactors and in order to see how far the French thermal-hydraulic safety code CATHARE can predict the physical phenomena during various RBMK transients. (12 refs.)
International Nuclear Information System (INIS)
Blink, J.A.
1985-03-01
In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs
2009-10-01
Becker-Kistiakowsky-Wilson (BKW) 2. Jacobs-Cowperthwaite-Zwisler ( JCZ ) 3. Hayes 4. Davis 5. Williamsburg 6. JWL 7. HOM the JWL and HOM EOS...have often been used in hydrocode/CFD simulations. On the other hand, the BKW and JCZ equations remain the EOS of choice in chemical equilibrium code...development for condensed explosives. Various databases have been constructed for the BKW and JCZ equations of state. These include: BKWC
Energy Technology Data Exchange (ETDEWEB)
Galindo G, I. F.; Vergara del C, J. A.; Galvan A, S. J. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Tijerina S, F., E-mail: francisco.tijerina@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)
2016-09-15
The use of specialized codes to estimate the radiation dose projection to an emergency postulated event at a nuclear power plant requires that certain plant data be available according to the event being simulated. The calculation of the possible radiological release is the critical activity to carry out the emergency actions. However, not all of the plant data required are obtained directly from the plant but need to be calculated. In this paper we present a computational tool that calculates the plant data required to use the radiological dose estimation codes. The tool provides the required information when there is a gas emergency venting event in the primary containment atmosphere, whether well or dry well and also calculates the time in which the spent fuel pool would be discovered in the event of a leak of water on some of the walls or floor of the pool. The tool developed has mathematical models for the processes involved such as: compressible flow in pipes considering area change and for constant area, taking into account the effects of friction and for the case of the spent fuel pool hydraulic models to calculate the time in which a container is emptied. The models implemented in the tool are validated with data from the literature for simulated cases. The results with the tool are very similar to those of reference. This tool will also be very supportive so that in postulated emergency cases can use the radiological dose estimation codes to adequately and efficiently determine the actions to be taken in a way that affects as little as possible. (Author)
Energy Technology Data Exchange (ETDEWEB)
Blazy-Aubignac, L
2007-09-15
The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)
Relay-aided multi-cell broadcasting with random network coding
DEFF Research Database (Denmark)
Lu, Lu; Sun, Fan; Xiao, Ming
2010-01-01
We investigate a relay-aided multi-cell broadcasting system using random network codes, where the focus is on devising efficient scheduling algorithms between relay and base stations. Two scheduling algorithms are proposed based on different feedback strategies; namely, a one-step scheduling...
Energy Technology Data Exchange (ETDEWEB)
Boehlke, S.; Niegoth, H. [STEAG Energy Services GmbH, Essen (Germany). Nuclear Technologies; Stalder, I. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland)
2012-11-01
In the nuclear power plant Leibstadt (KKL) during the next year large components will be dismantled and stored for final disposal within the interim storage facility ZENT at the NPP site. Before construction of ZENT appropriate estimations of the local dose rate inside and outside the building and the collective dose for the normal operation have to be performed. The shielding calculations are based on the properties of the stored components and radiation sources and on the concepts for working place requirements. The installation of control and monitoring areas will depend on these calculations. For the determination of the shielding potential of concrete walls and steel doors with the defined boundary conditions point-kernel codes like MICROSHIELd {sup registered} are used. Complex problems cannot be modeled with this code. Therefore the point-kernel code VISIPLAN {sup registered} was developed for the determination of the local dose distribution functions in 3D models. The possibility of motion sequence inputs allows an optimization of collective dose estimations for the operational phases of a nuclear facility.
Energy Technology Data Exchange (ETDEWEB)
Kai, Tetsuya; Maekawa, Fujio; Kasugai, Yoshimi; Takada, Hiroshi; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kosako, Kazuaki [Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan)
2002-03-01
Reliability assessment for the high energy particle induced radioactivity calculation code DCHAIN-SP 2001 was carried out through analysis of integral activation experiments with 14-MeV neutrons aiming at validating the cross section and decay data revised from previous version. The following three kinds of experiments conducted at the D-T neutron source facility, FNS, in JAERI were employed: (1) the decay gamma-ray measurement experiment for fusion reactor materials, (2) the decay heat measurement experiment for 32 fusion reactor materials, and (3) the integral activation experiment on mercury. It was found that the calculations with DCHAIN-SP 2001 predicted the experimental data for (1) - (3) within several tens of percent. It was concluded that the cross section data below 20 MeV and the associated decay data as well as the calculation algorithm for solving the Beteman equation that was the master equation of DCHAIN-SP were adequate. (author)
International Nuclear Information System (INIS)
Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.
2006-01-01
The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)
International Nuclear Information System (INIS)
Damiani, Daniela D.; Cruz, Carlos M.; Pinnera, Ibrahin; Abreu, Yamiel; Leyva, Antonio
2015-01-01
New developments and simulations on regard to the interactions of incident gamma radiation over solids materials using the MCSAD (Monte Carlo Simulation of Atom Displacement) code are presented. In this code Monte Carlo algorithms are applied in order to sample all electrons and gamma interaction processes occurring during their transport through a solid target, especially those connected to the output of atom displacements events. Particularly, it is calculated the limit angle to elastic scattering for the electrons on a new approach, which allows correctly the splitting of the electron single processes at higher scattering angles. On this way, the probability of single electron scattering processes transferring high recoil atomic energy leading to atom displacement effects is calculated and consequently sampled in the MCSAD code. In addition, it is considered some other new theoretical aspects in order to improve previous versions, like the one concerning the selection of threshold energy for displacements at a given atom site in dependence of the atom recoil direction. (Author)
International Nuclear Information System (INIS)
Boehlke, S.; Niegoth, H.
2012-01-01
In the nuclear power plant Leibstadt (KKL) during the next year large components will be dismantled and stored for final disposal within the interim storage facility ZENT at the NPP site. Before construction of ZENT appropriate estimations of the local dose rate inside and outside the building and the collective dose for the normal operation have to be performed. The shielding calculations are based on the properties of the stored components and radiation sources and on the concepts for working place requirements. The installation of control and monitoring areas will depend on these calculations. For the determination of the shielding potential of concrete walls and steel doors with the defined boundary conditions point-kernel codes like MICROSHIELd registered are used. Complex problems cannot be modeled with this code. Therefore the point-kernel code VISIPLAN registered was developed for the determination of the local dose distribution functions in 3D models. The possibility of motion sequence inputs allows an optimization of collective dose estimations for the operational phases of a nuclear facility.
International Nuclear Information System (INIS)
Paulus, J.; Brockmeier, U.
1992-01-01
This report deals with a CORA-9 post-test-calculation with the computer-code RELAP5/SCDAP Mod.2.5 Ver.3f. It is the second technical report concerning the BMFT founded project ''Comparative Assessment of Different Computer Codes for Severe Accident Analysis. Contribution to the ATHLET/SA-Code Development'', which is performed at the ''Department of Nuclear and New Energy Systems (NES)'' at the Ruhr-University of Bochum. Three main phases during the temperature transient in CORA-9, namely the heatup phase, the escalation phase and the cooldown phase are covered by the calculations. The computations confirm that the governing phenomena, heat up, oxidation with the onset of the temperature excursion due to Zircaloy/steam reaction, and the cooldown phase are modeled satisfactorily by RELAP5/SCDAP. Investigations indicate that code underpredictions of the cladding surface temperature during the heatup phase are probably due to condensation effects occuring in the thermal hydraulic volumes. (orig.) [de
Magro, G.; Dahle, T. J.; Molinelli, S.; Ciocca, M.; Fossati, P.; Ferrari, A.; Inaniwa, T.; Matsufuji, N.; Ytre-Hauge, K. S.; Mairani, A.
2017-05-01
Particle therapy facilities often require Monte Carlo (MC) simulations to overcome intrinsic limitations of analytical treatment planning systems (TPS) related to the description of the mixed radiation field and beam interaction with tissue inhomogeneities. Some of these uncertainties may affect the computation of effective dose distributions; therefore, particle therapy dedicated MC codes should provide both absorbed and biological doses. Two biophysical models are currently applied clinically in particle therapy: the local effect model (LEM) and the microdosimetric kinetic model (MKM). In this paper, we describe the coupling of the NIRS (National Institute for Radiological Sciences, Japan) clinical dose to the FLUKA MC code. We moved from the implementation of the model itself to its application in clinical cases, according to the NIRS approach, where a scaling factor is introduced to rescale the (carbon-equivalent) biological dose to a clinical dose level. A high level of agreement was found with published data by exploring a range of values for the MKM input parameters, while some differences were registered in forward recalculations of NIRS patient plans, mainly attributable to differences with the analytical TPS dose engine (taken as reference) in describing the mixed radiation field (lateral spread and fragmentation). We presented a tool which is being used at the Italian National Center for Oncological Hadrontherapy to support the comparison study between the NIRS clinical dose level and the LEM dose specification.
International Nuclear Information System (INIS)
Auder, Benjamin
2011-01-01
This research thesis has been made within the frame of a project on nuclear reactor vessel life. It deals with the use of numerical codes aimed at estimating probability densities for every input parameter in order to calculate probability margins at the output level. More precisely, it deals with codes with one-dimensional functional responses. The author studies the numerical simulation of a pressurized thermal shock on a nuclear reactor vessel, i.e. one of the possible accident types. The study of the vessel integrity relies on a thermal-hydraulic analysis and on a mechanical analysis. Algorithms are developed and proposed for each of them. Input-output data are classified using a clustering technique and a graph-based representation. A method for output dimension reduction is proposed, and a regression is applied between inputs and reduced representations. Applications are discussed in the case of modelling and sensitivity analysis for the CATHARE code (a code used at the CEA for the thermal-hydraulic analysis)
International Nuclear Information System (INIS)
Dos Santos, Adimir; Siqueira, Paulo de Tarso D.; Andrade e Silva, Graciete Simões; Grant, Carlos; Tarazaga, Ariel E.; Barberis, Claudia
2013-01-01
In year 2008 the Atomic Energy National Commission (CNEA) of Argentina, and the Brazilian Institute of Energetic and Nuclear Research (IPEN), under the frame of Nuclear Energy Argentine Brazilian Agreement (COBEN), among many others, included the project “Validation and Verification of Calculation Methods used for Research and Experimental Reactors . At this time, it was established that the validation was to be performed with models implemented in the deterministic codes HUEMUL and PUMA (cell and reactor codes) developed by CNEA and those ones implemented in MCNP by CNEA and IPEN. The necessary data for these validations would correspond to theoretical-experimental reference cases in the research reactor IPEN/MB-01 located in São Paulo, Brazil. The staff of the group Reactor and Nuclear Power Studies (SERC) of CNEA, from the argentine side, performed calculations with deterministic models (HUEMUL-PUMA) and probabilistic methods (MCNP) modeling a great number of physical situations of de reactor, which previously have been studied and modeled by members of the Center of Nuclear Engineering of the IPEN, whose results were extensively provided to CNEA. In this paper results of comparison of calculated and experimental results for temperature coefficients, kinetic parameters and fission rates spatial distributions are shown. (author)
International Nuclear Information System (INIS)
Fryer, L.S.
1978-12-01
TIRION 4 is the most recent program in a series designed to calculate the consequences of releasing radioactive material to the atmosphere. A brief description of the models used in the program and full details of the various control cards necessary to run TIRION 4 are given. (author)
International Nuclear Information System (INIS)
Madeira, A.A.
1985-01-01
It's studied the improvement for the calculation of bubble rise velocity, by adding two different ways to estimate this velocity, one of which more adequate to pressures normally found in the Reactor Cooling System. Additionally, a limitation in bubble rise velocity growth was imposed, to account for the actual behavior of bubble rise in two-phase mixtures. (Author) [pt
International Nuclear Information System (INIS)
Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.
2007-01-01
In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required
Color-coded Live Imaging of Heterokaryon Formation and Nuclear Fusion of Hybridizing Cancer Cells.
Suetsugu, Atsushi; Matsumoto, Takuro; Hasegawa, Kosuke; Nakamura, Miki; Kunisada, Takahiro; Shimizu, Masahito; Saji, Shigetoyo; Moriwaki, Hisataka; Bouvet, Michael; Hoffman, Robert M
2016-08-01
Fusion of cancer cells has been studied for over half a century. However, the steps involved after initial fusion between cells, such as heterokaryon formation and nuclear fusion, have been difficult to observe in real time. In order to be able to visualize these steps, we have established cancer-cell sublines from the human HT-1080 fibrosarcoma, one expressing green fluorescent protein (GFP) linked to histone H2B in the nucleus and a red fluorescent protein (RFP) in the cytoplasm and the other subline expressing RFP in the nucleus (mCherry) linked to histone H2B and GFP in the cytoplasm. The two reciprocal color-coded sublines of HT-1080 cells were fused using the Sendai virus. The fused cells were cultured on plastic and observed using an Olympus FV1000 confocal microscope. Multi-nucleate (heterokaryotic) cancer cells, in addition to hybrid cancer cells with single-or multiple-fused nuclei, including fused mitotic nuclei, were observed among the fused cells. Heterokaryons with red, green, orange and yellow nuclei were observed by confocal imaging, even in single hybrid cells. The orange and yellow nuclei indicate nuclear fusion. Red and green nuclei remained unfused. Cell fusion with heterokaryon formation and subsequent nuclear fusion resulting in hybridization may be an important natural phenomenon between cancer cells that may make them more malignant. The ability to image the complex processes following cell fusion using reciprocal color-coded cancer cells will allow greater understanding of the genetic basis of malignancy. Copyright© 2016 International Institute of Anticancer Research (Dr. John G. Delinassios), All rights reserved.
Effect of cosine current approximation in lattice cell calculations in cylindrical geometry
International Nuclear Information System (INIS)
Mohanakrishnan, P.
1978-01-01
It is found that one-dimensional cylindrical geometry reactor lattice cell calculations using cosine angular current approximation at spatial mesh interfaces give results surprisingly close to the results of accurate neutron transport calculations as well as experimental measurements. This is especially true for tight light water moderated lattices. Reasons for this close agreement are investigated here. By re-examining the effects of reflective and white cell boundary conditions in these calculations it is concluded that one major reason is the use of white boundary condition necessitated by the approximation of the two-dimensional reactor lattice cell by a one-dimensional one. (orig.) [de
The use of electromagnetic particle-in-cell codes in accelerator applications
International Nuclear Information System (INIS)
Eppley, K.
1988-12-01
The techniques developed for the numerical simulation of plasmas have numerous applications relevant to accelerators. The operation of many accelerator components involves transients, interactions between beams and rf fields, and internal plasma oscillations. These effects produce non-linear behavior which can be represented accurately by particle in cell (PIC) simulations. We will give a very brief overview of the algorithms used in PIC Codes. We will examine the range of parameters over which they are useful. We will discuss the factors which determine whether a two or three dimensional simulation is most appropriate. PIC codes have been applied to a wide variety of diverse problems, spanning many of the systems in a linear accelerator. We will present a number of practical examples of the application of these codes to areas such as guns, bunchers, rf sources, beam transport, emittance growth and final focus. 8 refs., 8 figs., 2 tabs
Final report for a Ministry of Industry and Trade Project: ISP 42 - calculations by the CATHARE code
International Nuclear Information System (INIS)
Malacka, M.
2001-12-01
The following phases of the ISP-42 experiment were analyzed: A - start of the passive cooling system (steam injection into vessels with air); B - emptying of the gravity-controlled cooling system (flooding of the reactor core); long-term removal of residual heat (heat exchangers); and D - overload of the passive cooling system. The results of the pre-test and post-test calculation are compared with respect to the observed values. (P.A.)
Energy Technology Data Exchange (ETDEWEB)
New, Joshua Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Levinson, Ronnen [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Huang, Yu [White Box Technologies, Salt Lake City, UT (United States); Sanyal, Jibonananda [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mellot, Joe [The Garland Company, Cleveland, OH (United States); Childs, Kenneth W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kriner, Scott [Green Metal Consulting, Inc., Macungie, PA (United States)
2014-06-01
The Roof Savings Calculator (RSC) was developed through collaborations among Oak Ridge National Laboratory (ORNL), White Box Technologies, Lawrence Berkeley National Laboratory (LBNL), and the Environmental Protection Agency in the context of a California Energy Commission Public Interest Energy Research project to make cool-color roofing materials a market reality. The RSC website and a simulation engine validated against demonstration homes were developed to replace the liberal DOE Cool Roof Calculator and the conservative EPA Energy Star Roofing Calculator, which reported different roof savings estimates. A preliminary analysis arrived at a tentative explanation for why RSC results differed from previous LBNL studies and provided guidance for future analysis in the comparison of four simulation programs (doe2attic, DOE-2.1E, EnergyPlus, and MicroPas), including heat exchange between the attic surfaces (principally the roof and ceiling) and the resulting heat flows through the ceiling to the building below. The results were consolidated in an ORNL technical report, ORNL/TM-2013/501. This report is an in-depth inter-comparison of four programs with detailed measured data from an experimental facility operated by ORNL in South Carolina in which different segments of the attic had different roof and attic systems.
Energy Technology Data Exchange (ETDEWEB)
Silvestre, Larissa Jacome Barros
2016-07-01
Nuclear accidents in the world led to the establishment of rigorous criteria and requirements for nuclear power plant operations by the international regulatory bodies. By using specific computer programs, simulations of various accidents and transients likely to occur at any nuclear power plant are required for certifying and licensing a nuclear power plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most widely used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors in Brazil and worldwide. A major difficulty in the simulation by using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data requires a great number of mathematical operations to calculate the geometry of the components. Thus, for those calculations performance and preparation of RELAP5 input data, a friendly mathematical preprocessor was designed. The Visual Basic for Application (VBA) for Microsoft Excel demonstrated to be an effective tool to perform a number of tasks in the development of the program. In order to meet the needs of RELAP5 users, the RELAP5 Calculation Program (Programa de Calculo do RELAP5 - PCRELAP5) was designed. The components of the code were codified; all entry cards including the optional cards of each one have been programmed. In addition, an English version for PCRELAP5 was provided. Furthermore, a friendly design was developed in order to minimize the time of preparation of input data and errors committed by users. In this work, the final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra 2. (author)
A general concurrent algorithm for plasma particle-in-cell simulation codes
International Nuclear Information System (INIS)
Liewer, P.C.; Decyk, V.K.
1989-01-01
We have developed a new algorithm for implementing plasma particle-in-cell (PIC) simulation codes on concurrent processors with distributed memory. This algorithm, named the general concurrent PIC algorithm (GCPIC), has been used to implement an electrostatic PIC code on the 33-node JPL Mark III Hypercube parallel computer. To decompose at PIC code using the GCPIC algorithm, the physical domain of the particle simulation is divided into sub-domains, equal in number to the number of processors, such that all sub-domains have roughly equal numbers of particles. For problems with non-uniform particle densities, these sub-domains will be of unequal physical size. Each processor is assigned a sub-domain and is responsible for updating the particles in its sub-domain. This algorithm has led to a a very efficient parallel implementation of a well-benchmarked 1-dimensional PIC code. The dominant portion of the code, updating the particle positions and velocities, is nearly 100% efficient when the number of particles is increased linearly with the number of hypercube processors used so that the number of particles per processor is constant. For example, the increase in time spent updating particles in going from a problem with 11,264 particles run on 1 processor to 360,448 particles on 32 processors was only 3% (parallel efficiency of 97%). Although implemented on a hypercube concurrent computer, this algorithm should also be efficient for PIC codes on other parallel architectures and for large PIC codes on sequential computers where part of the data must reside on external disks. copyright 1989 Academic Press, Inc
International Nuclear Information System (INIS)
De Carlan, L.; Bordy, J.M.; Gouriou, J.
2010-01-01
Within the frame of the CONRAD European project (Coordination Network for Radiation Dosimetry), and more precisely within a work group paying attention to uncertainty assessment in computational dosimetry and aiming at comparing different approaches, the authors report the simulation of an irradiator containing a caesium 137 source to calculate the kerma in air as well as its uncertainty due to different parameters. They present the problem geometry, recall the studied issues (kerma uncertainty, influence of capsule source, influence of the collimator, influence of the air volume surrounding the source). They indicate the codes which have been used (MNCP, Fluka, Penelope, etc.) and discuss the obtained results for the first issue
Energy Technology Data Exchange (ETDEWEB)
Pietrzak, Robert [Department of Nuclear Physics and Its Applications, Institute of Physics, University of Silesia, Katowice (Poland); Konefał, Adam, E-mail: adam.konefal@us.edu.pl [Department of Nuclear Physics and Its Applications, Institute of Physics, University of Silesia, Katowice (Poland); Sokół, Maria; Orlef, Andrzej [Department of Medical Physics, Maria Sklodowska-Curie Memorial Cancer Center, Institute of Oncology, Gliwice (Poland)
2016-08-01
The success of proton therapy depends strongly on the precision of treatment planning. Dose distribution in biological tissue may be obtained from Monte Carlo simulations using various scientific codes making it possible to perform very accurate calculations. However, there are many factors affecting the accuracy of modeling. One of them is a structure of objects called bins registering a dose. In this work the influence of bin structure on the dose distributions was examined. The MCNPX code calculations of Bragg curve for the 60 MeV proton beam were done in two ways: using simple logical detectors being the volumes determined in water, and using a precise model of ionization chamber used in clinical dosimetry. The results of the simulations were verified experimentally in the water phantom with Marcus ionization chamber. The average local dose difference between the measured relative doses in the water phantom and those calculated by means of the logical detectors was 1.4% at first 25 mm, whereas in the full depth range this difference was 1.6% for the maximum uncertainty in the calculations less than 2.4% and for the maximum measuring error of 1%. In case of the relative doses calculated with the use of the ionization chamber model this average difference was somewhat greater, being 2.3% at depths up to 25 mm and 2.4% in the full range of depths for the maximum uncertainty in the calculations of 3%. In the dose calculations the ionization chamber model does not offer any additional advantages over the logical detectors. The results provided by both models are similar and in good agreement with the measurements, however, the logical detector approach is a more time-effective method. - Highlights: • Influence of the bin structure on the proton dose distributions was examined for the MC simulations. • The considered relative proton dose distributions in water correspond to the clinical application. • MC simulations performed with the logical detectors and the
Long non-coding RNA LINC00261 sensitizes human colon cancer cells to cisplatin therapy.
Wang, Z K; Yang, L; Wu, L L; Mao, H; Zhou, Y H; Zhang, P F; Dai, G H
2017-12-11
Colon cancer is one of the most common digestive tumors. The present study aimed to explore the functional role, as well as the underlying mechanism of long non-coding RNA LINC00261 in colon cancer. Expression of LINC00261 was analyzed in colon cancer cell lines and human normal cell lines. Acquired resistance cell lines were then built and the acquired resistance efficiency was detected by evaluating cell viability. Thereafter, the effects of LINC00261 overexpression on cisplatin-resistant colon cancer cells were measured, as well as cell apoptosis, viability, migration, and invasion. Subsequently, we investigated the interaction of LINC00261 and β-catenin. The results showed that the LINC00261 gene was down-regulated in colon cancer cell lines and tissues, and in cisplatin-resistant cells. LINC00261 overexpression might relieve cisplatin resistance of colon cancer cells via promoting cell apoptosis, and inhibiting cell viability, migration, and invasion. Moreover, LINC00261 might down-regulate nuclear β-catenin through restraining β-catenin from cytoplasm into nuclei or it could also promote β-catenin degradation and inhibit activation of Wnt pathway. Finally, LINC00261 reduced cisplatin resistance of colon cancer in vivo and enhanced the anti-colon cancer effect of cisplatin through reducing tumor volume and weight.
Khattab, K; Sulieman, I
2009-04-01
The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.
First experience with particle-in-cell plasma physics code on ARM-based HPC systems
Sáez, Xavier; Soba, Alejandro; Sánchez, Edilberto; Mantsinen, Mervi; Mateo, Sergi; Cela, José M.; Castejón, Francisco
2015-09-01
In this work, we will explore the feasibility of porting a Particle-in-cell code (EUTERPE) to an ARM multi-core platform from the Mont-Blanc project. The used prototype is based on a system-on-chip Samsung Exynos 5 with an integrated GPU. It is the first prototype that could be used for High-Performance Computing (HPC), since it supports double precision and parallel programming languages.
Hoffman, Robert M
2016-03-01
Fluorescent proteins are very bright and available in spectrally-distinct colors, enable the imaging of color-coded cancer cells growing in vivo and therefore the distinction of cancer cells with different genetic properties. Non-invasive and intravital imaging of cancer cells with fluorescent proteins allows the visualization of distinct genetic variants of cancer cells down to the cellular level in vivo. Cancer cells with increased or decreased ability to metastasize can be distinguished in vivo. Gene exchange in vivo which enables low metastatic cancer cells to convert to high metastatic can be color-coded imaged in vivo. Cancer stem-like and non-stem cells can be distinguished in vivo by color-coded imaging. These properties also demonstrate the vast superiority of imaging cancer cells in vivo with fluorescent proteins over photon counting of luciferase-labeled cancer cells.
Directory of Open Access Journals (Sweden)
Xuefei Yu
2018-01-01
Full Text Available The mean amplitude of glycemic excursions (MAGE is an essential index for glycemic variability assessment, which is treated as a key reference for blood glucose controlling at clinic. However, the traditional “ruler and pencil” manual method for the calculation of MAGE is time-consuming and prone to error due to the huge data size, making the development of robust computer-aided program an urgent requirement. Although several software products are available instead of manual calculation, poor agreement among them is reported. Therefore, more studies are required in this field. In this paper, we developed a mathematical algorithm based on integer nonlinear programming. Following the proposed mathematical method, an open-code computer program named MAGECAA v1.0 was developed and validated. The results of the statistical analysis indicated that the developed program was robust compared to the manual method. The agreement among the developed program and currently available popular software is satisfied, indicating that the worry about the disagreement among different software products is not necessary. The open-code programmable algorithm is an extra resource for those peers who are interested in the related study on methodology in the future.
International Nuclear Information System (INIS)
Sato, Tatsuhiko; Endo, Akira; Zankl, Maria; Petoussi-Henss, Nina; Niita, Koji
2009-01-01
The fluence to organ-dose and effective-dose conversion coefficients for neutrons and protons with energies up to 100 GeV was calculated using the PHITS code coupled to male and female adult reference computational phantoms, which are to be released as a common ICRP/ICRU publication. For the calculation, the radiation and tissue weighting factors, w R and w T , respectively, as revised in ICRP Publication 103 were employed. The conversion coefficients for effective dose equivalents derived using the radiation quality factors of both Q(L) and Q(y) relationships were also estimated, utilizing the functions for calculating the probability densities of the absorbed dose in terms of LET (L) and lineal energy (y), respectively, implemented in PHITS. By comparing these data with the corresponding data for the effective dose, we found that the numerical compatibilities of the revised w R with the Q(L) and Q(y) relationships are fairly established. The calculated data of these dose conversion coefficients are indispensable for constructing the radiation protection systems based on the new recommendations given in ICRP103 for aircrews and astronauts, as well as for workers in accelerators and nuclear facilities.
Energy Technology Data Exchange (ETDEWEB)
Montégiani, Jean-François; Gaudin, Émilie; Després, Philippe [Physics, Engineering Physics and Optics, Université Laval, Quebec City, QC (Canada); Jackson, Price A. [Molecular Imaging and Targeted Therapeutics, Peter MacCallum Cancer Centre, Melbourne, VIC (Australia); Beauregard, Jean-Mathieu [Radiology, Université Laval, Quebec City, QC (Canada)
2014-08-15
In peptide receptor radionuclide therapy (PRRT), huge inter-patient variability in absorbed radiation doses per administered activity mandates the utilization of individualized dosimetry to evaluate therapeutic efficacy and toxicity. We created a reliable GPU-calculated dosimetry code (irtGPUMCD) and assessed {sup 177}Lu-octreotate renal dosimetry in eight patients (4 cycles of approximately 7.4 GBq). irtGPUMCD was derived from a brachytherapy dosimetry code (bGPUMCD), which was adapted to {sup 177}Lu PRRT dosimetry. Serial quantitative single-photon emission computed tomography (SPECT) images were obtained from three SPECT/CT acquisitions performed at 4, 24 and 72 hours after {sup 177}Lu-octreotate administration, and registered with non-rigid deformation of CT volumes, to obtain {sup 177}Lu-octreotate 4D quantitative biodistribution. Local energy deposition from the β disintegrations was assumed. Using Monte Carlo gamma photon transportation, irtGPUMCD computed dose rate at each time point. Average kidney absorbed dose was obtained from 1-cm{sup 3} VOI dose rate samples on each cortex, subjected to a biexponential curve fit. Integration of the latter time-dose rate curve yielded the renal absorbed dose. The mean renal dose per administered activity was 0.48 ± 0.13 Gy/GBq (range: 0.30–0.71 Gy/GBq). Comparison to another PRRT dosimetry code (VRAK: Voxelized Registration and Kinetics) showed fair accordance with irtGPUMCD (11.4 ± 6.8 %, range: 3.3–26.2%). These results suggest the possibility to use the irtGPUMCD code in order to personalize administered activity in PRRT. This could allow improving clinical outcomes by maximizing per-cycle tumor doses, without exceeding the tolerable renal dose.
Directory of Open Access Journals (Sweden)
Peter J. Arthur-Farraj
2017-09-01
Full Text Available Repair Schwann cells play a critical role in orchestrating nerve repair after injury, but the cellular and molecular processes that generate them are poorly understood. Here, we perform a combined whole-genome, coding and non-coding RNA and CpG methylation study following nerve injury. We show that genes involved in the epithelial-mesenchymal transition are enriched in repair cells, and we identify several long non-coding RNAs in Schwann cells. We demonstrate that the AP-1 transcription factor C-JUN regulates the expression of certain micro RNAs in repair Schwann cells, in particular miR-21 and miR-34. Surprisingly, unlike during development, changes in CpG methylation are limited in injury, restricted to specific locations, such as enhancer regions of Schwann cell-specific genes (e.g., Nedd4l, and close to local enrichment of AP-1 motifs. These genetic and epigenomic changes broaden our mechanistic understanding of the formation of repair Schwann cell during peripheral nervous system tissue repair.
Directory of Open Access Journals (Sweden)
Anthony A Philippakis
2006-05-01
Full Text Available While combinatorial models of transcriptional regulation can be inferred for metazoan systems from a priori biological knowledge, validation requires extensive and time-consuming experimental work. Thus, there is a need for computational methods that can evaluate hypothesized cis regulatory codes before the difficult task of experimental verification is undertaken. We have developed a novel computational framework (termed "CodeFinder" that integrates transcription factor binding site and gene expression information to evaluate whether a hypothesized transcriptional regulatory model (TRM; i.e., a set of co-regulating transcription factors is likely to target a given set of co-expressed genes. Our basic approach is to simultaneously predict cis regulatory modules (CRMs associated with a given gene set and quantify the enrichment for combinatorial subsets of transcription factor binding site motifs comprising the hypothesized TRM within these predicted CRMs. As a model system, we have examined a TRM experimentally demonstrated to drive the expression of two genes in a sub-population of cells in the developing Drosophila mesoderm, the somatic muscle founder cells. This TRM was previously hypothesized to be a general mode of regulation for genes expressed in this cell population. In contrast, the present analyses suggest that a modified form of this cis regulatory code applies to only a subset of founder cell genes, those whose gene expression responds to specific genetic perturbations in a similar manner to the gene on which the original model was based. We have confirmed this hypothesis by experimentally discovering six (out of 12 tested new CRMs driving expression in the embryonic mesoderm, four of which drive expression in founder cells.
International Nuclear Information System (INIS)
Baranenko, V.I.; Oleynik, S.G.; Istomin, R.S.; Kumov, A.V.; Dolgikh, A.P.; Kornienko, K.A.
2006-09-01
Full text of publication follows: The flow - accelerated corrosion (FAC) is a widespread source of damage in NPP piping. Last years many tasks concerning service life control of pipe line system which are exposed to FAC were solved. The tasks included development of computer codes (CC) for calculations of FAC rate, thinning degree of pipe walls and assessment of tolerance of wall thickness for sections with common and local thinning of the wall. The Russian NPPs began to organize the life control service for the pipe systems exposed to FAC, later than in other countries. Usage of foreign and domestic experience allowed reducing of time consumed for CC development for FAC calculation, for wall thickness tolerance and other information. In this report the experience of development, certification of control authorities and CC usage are analyzed on NPPs with WWER reactor systems. In accordance with rules and regulations for NPP in force in Russia the utilization of developed CC have to be certified by control authorities. Other worked out documentation demands approval of control organs, as well. The list of CCs in Russia (EKI-01, EKI-02, EKI-03 and EKI-04) is the same as the well known foreign CCs, namely CHECWORKS in USA and COMSY in Germany. In contrast to Russia, the foreign CCs are not calibrated by authorised organisations. As the Russian experience shows, the CC calibration accompanies the increasing of requirements for vindication of the calculation accuracy and a more precise definition of a factor range included in CC. Type of reactor and pipe lines system list have to be put in the verification reports. The report of CC development has to include results of FAC calculation of reactor pipe lines indicated in the verification documents. Verification necessity increases time required by the CC development. The certification of Russian control organisations is in conformity with a new developed (2000-2002) management directive. The computer codes EKI-01 and EKI
Directory of Open Access Journals (Sweden)
Burkhard eSchütz
2015-03-01
Full Text Available The mouse gastro-intestinal and biliary tract mucosal epithelia harbor choline acetyltransferase (ChAT-positive brush cells with taste cell-like traits. With the aid of two transgenic mouse lines that express green fluorescent protein (EGFP under the control of the ChAT promoter (EGFPChAT and by using in situ hybridization and immunohistochemistry we found that EGFPChAT cells were clustered in the epithelium lining the gastric groove. EGFPChAT cells were numerous in the gall bladder and bile duct, and found scattered as solitary cells along the small and large intestine. While all EGFPChAT cells were also ChAT-positive, expression of the high-affinity choline transporter (ChT1 was never detected. Except for the proximal colon, EGFPChAT cells also lacked detectable expression of the vesicular acetylcholine transporter (VAChT. EGFPChAT cells were found to be separate from enteroendocrine cells, however they were all immunoreactive for cytokeratin 18 (CK18, transient receptor potential melastatin-like subtype 5 channel (TRPM5, and for cyclooxygenases 1 (COX1 and 2 (COX2. The ex vivo stimulation of colonic EGFPChAT cells with the bitter substance denatonium resulted in a strong increase in intracellular calcium, while in other epithelial cells such an increase was significantly weaker and also timely delayed. Subsequent stimulation with cycloheximide was ineffective in both cell populations. Given their chemical coding and chemosensory properties, EGFPChAT brush cells thus may have integrative functions and participate in induction of protective reflexes and inflammatory events by utilizing ACh and prostaglandins for paracrine signaling.
Shielding Calculations for Industrial 5/7.5MeV Electron Accelerators Using the MCNP Monte Carlo Code
International Nuclear Information System (INIS)
Peri, E.; Orion, I.
2014-01-01
High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, in order to extend the shelf life of products. High energy photons can cause food activation due to (D 3 ,n) reactions. Until 2004, to eliminate the possibility of food activation, the electron energy was limited to 5 MeV X-rays for food irradiation. In 2004, the FDA approved the usage of up to 7.5 MeV, but only with tantalum and gold targets (1). Higher X-ray energy results an increased flux of X-rays in the forward direction, increased penetration, and higher photon dose rate due to better electron-to-photon conversion. These improvements could decrease the irradiation time and allow irradiation of larger packages, thereby providing higher production rates with lower treatment cost. Medical accelerators usually work with 6-18 MV electron energy with tungsten target to convert the electron beam to X-rays. In order to protect the patients, the accelerator head is protected with a heavy lead shielding; therefore, the bremsstrahlung is emitted only in the forward direction. There are many publications and standards that guide how to design optimal shielding for medical accelerator rooms. The shielding data for medical accelerators is not applicable for industrial accelerators, since the data is for different conversion targets, different X-Ray energies, and only for the forward direction. Collimators are not always in use in industrial accelerators, and therefore bremsstrahlung photons can be emitted in all directions. The bremsstrahlung spectrum and dose rate change as a function of the emission angle. The dose rate decreases from maximum in the forward direction (0°) to minimum at 180° by 1-2 orders of magnitude. In order to design and calculate optimal shielding for food accelerator rooms, there is a need to have the bremsstrahlung spectrum data, dose rates and concrete attenuation data in all emission directions
Long non-coding RNA TUG1 promotes cell proliferation and metastasis in human breast cancer.
Li, Teng; Liu, Yun; Xiao, Haifeng; Xu, Guanghui
2017-07-01
Long non-coding RNAs (LncRNAs) utilize a wide variety of mechanisms to regulate RNAs or proteins on the transcriptional or post-transcriptional levels. Accumulating studies have identified numerous LncRNAs to exert critical effects on different physiological processes, genetic disorders, and human diseases. Both clinical tissues from breast cancer patients and cultured cells were used for the qRT-PCR analysis. Specific siRNAs were included to assess the roles of TUG1 with cell viability assay, transwell assay, and cell apoptosis assay, respectively. The expression of TUG1 was enhanced in breast cancerous tissues and in highly invasive breast cancer cell lines and was associated with clinical variables, including tumor size, distant metastasis and TNM staging. Knockdown of TUG1 significantly slowed down cell proliferation, cell migration, and invasion in breast cancer cell lines MDA-MB-231 and MDA-MB-436. In addition, cell apoptotic rate was shown to increase upon siTUG1 treatment as evidenced by increases of the activities of caspase-3 and caspase-9. The identification of TUG1 as a critical mediator of breast cancer progression implied that it might serve as a biomarker for the diagnosis and treatment of breast cancer in clinic.
International Nuclear Information System (INIS)
White, Morgan C.
2000-01-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to
Energy Technology Data Exchange (ETDEWEB)
White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second
Directory of Open Access Journals (Sweden)
Mohammad Mirzaie
2012-09-01
Full Text Available Background: The 131I radioisotope is used for diagnosis and treatment of hyperthyroidism and thyroid cancer. In optimized Iodine therapy, a specific dose must be reached to the thyroid gland with minimum radiation to the cervical spine, cervical vertebrae, neck tissue, subcutaneous fat and skin. Dose measurement inside the alive organ is difficult therefore the aim of this research was dose calculation in the organs by MCNPX code. Materials and Methods: First of all, the input file for MCNPX code has been prepared to calculate F6 and F8 tallies for ellipsoidal thyroid lobes with long axes is tow times of short axes which the 131I is distributed uniformly inside the lobes. Then the code has been run for F6 and F8 tallies for variation of lobe volume from 1 to 25 milliliters. From the output file of tally F6, the gamma absorbed dose in ellipsoidal thyroid, spinal neck, neck bone, neck tissue, subcutaneous fat layer and skin for the volume lobe variation from 1 ml to 25 ml have been derived and the graphs are drew. As well as, form the output of F8 tally the absorbed energy of beta in thyroid and soft tissue of neck is obtained and listed in the table and then absorbed dose of bate has been calculated. Results: The results of this research show that for constant activity in thyroid, the absorbed dose of gamma decreases about 88.3% in thyroid, 6.9% at soft tissue, 19.3% in adipose layer and 17.4% in skin, but it increases 32.1% in spinal of neck and 32.3% in neck bone when the lobe volume varied from 1 to 25 milliliters. For the same situation, the beta absorbed dose decreases 95.9% in thyroid and 64.2% in soft tissue. Conclusion: For the constant activity in thyroid by increasing the thyroid volume, absorbed dose of gamma in thyroid and soft tissue of neck, adipose layer under the skin and skin of neck decreased, but it increased at spinal of neck and neck bone. Also, by increasing of the lobe volume in constant activity, the beta absorbed dose
International Nuclear Information System (INIS)
Tsuchihashi, Keichiro; Ishiguro, Yukio; Kaneko, Kunio; Ido, Masaru.
1986-09-01
Since the publication of JAERI-1285 in 1983 for the preliminary version of the SRAC code system, a number of additions and modifications to the functions have been made to establish an overall neutronics code system. Major points are (1) addition of JENDL-2 version of data library, (2) a direct treatment of doubly heterogeneous effect on resonance absorption, (3) a generalized Dancoff factor, (4) a cell calculation based on the fixed boundary source problem, (5) the corresponding edit required for experimental analysis and reactor design, (6) a perturbation theory calculation for reactivity change, (7) an auxiliary code for core burnup and fuel management, etc. This report is a revision of the users manual which consists of the general description, input data requirements and their explanation, detailed information on usage, mathematics, contents of libraries and sample I/O. (author)
Garcia, F.; Mesa, J.; Arruda-Neto, J. D. T.; Helene, O.; Vanin, V.; Milian, F.; Deppman, A.; Rodrigues, T. E.; Rodriguez, O.
2007-03-01
The code STATFLUX, implementing a new and simple statistical procedure for the calculation of transfer coefficients in radionuclide transport to animals and plants, is proposed. The method is based on the general multiple-compartment model, which uses a system of linear equations involving geometrical volume considerations. Flow parameters were estimated by employing two different least-squares procedures: Derivative and Gauss-Marquardt methods, with the available experimental data of radionuclide concentrations as the input functions of time. The solution of the inverse problem, which relates a given set of flow parameter with the time evolution of concentration functions, is achieved via a Monte Carlo simulation procedure. Program summaryTitle of program:STATFLUX Catalogue identifier:ADYS_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADYS_v1_0 Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Licensing provisions: none Computer for which the program is designed and others on which it has been tested:Micro-computer with Intel Pentium III, 3.0 GHz Installation:Laboratory of Linear Accelerator, Department of Experimental Physics, University of São Paulo, Brazil Operating system:Windows 2000 and Windows XP Programming language used:Fortran-77 as implemented in Microsoft Fortran 4.0. NOTE: Microsoft Fortran includes non-standard features which are used in this program. Standard Fortran compilers such as, g77, f77, ifort and NAG95, are not able to compile the code and therefore it has not been possible for the CPC Program Library to test the program. Memory required to execute with typical data:8 Mbytes of RAM memory and 100 MB of Hard disk memory No. of bits in a word:16 No. of lines in distributed program, including test data, etc.:6912 No. of bytes in distributed program, including test data, etc.:229 541 Distribution format:tar.gz Nature of the physical problem:The investigation of transport mechanisms for
International Nuclear Information System (INIS)
Smith, M.A.; Tsoulfanidis, N.; Lewis, E.E.; Palmiotti, G.
2001-01-01
Increasing computer power is allowing higher-order angular approximations to replace diffusion theory methods in whole core reactor physics computations. Spherical harmonic (P n ), simplified spherical harmonic (SP n ), and discrete ordinates (S n ) methods are capable of performing such calculations in three dimensions. Most advantages of such transport methods are gained by eliminating fuel assembly homogenization, thus allowing pin powers to be calculated directly. A further step, currently under investigation, is the elimination of spatial homogenization at the pin cell level as well. The fuel-moderator interfaces may be treated explicitly in P n , S n , or SP n calculations by applying triangular finite elements (FEM) to the spatial variables. Early results using a modified form of the VARIANT code, however, indicate that without pin cell homogenization, high-order angular approximations may be required to represent the lattice effects accurately within the whole-core calculations. To examine these lattice effects further, a modified form of VARIANT was created to use the spatial triangular finite element scheme. The program was set up to treat a single heterogeneous pin cell coupled with P n , SP n , or S n angular approximations. Additional modifications replaced the nodal interface approximations with exact reflected boundary conditions to increase the accuracy of the results. Several pressurized water reactor pin cells, taken from a previous benchmark specification, were examined. However, the results shown here focus only on the most severe case, i.e., a pin cell containing 8.7% mixed-oxide enriched fuel. The DRAGON collision probability code was used to collapse a 69-group cross-section library to a more manageable 7-group library that contained cross sections for the fuel-cladding mixture and for the water. Eigenvalue results are shown in Figs. 1 and 2 using the modified VARIANT code with P n , SP n , and S n angular approximations. A 7-group MCNP Monte
Directory of Open Access Journals (Sweden)
Yong Sun Lee
2012-10-01
Full Text Available nc886 (= pre-miR-886 or vtRNA2-1 is a non-coding RNA that has been recently identified as a natural repressor for the activity of PKR (Protein Kinase R. The suppression of nc886 activates PKR and thereby provokes a cell death pathway. When combined with the fact that nc886 is suppressed in a wide range of cancer cells, the nc886-PKR relationship suggests a tumor surveillance model. When neoplastic cells develop and nc886 decreases therein, PKR is released from nc886 and becomes the active phosphorylated form, which initiates an apoptotic cascade to eliminate those cells. The nc886-PKR pathway is distinct from conventional mechanisms, such as the immune surveillance hypothesis or intrinsic mechanisms that check/proofread the genomic integrity, and thus represents a novel example of tumor surveillance.
Fluorogenic RNA Mango aptamers for imaging small non-coding RNAs in mammalian cells.
Autour, Alexis; C Y Jeng, Sunny; D Cawte, Adam; Abdolahzadeh, Amir; Galli, Angela; Panchapakesan, Shanker S S; Rueda, David; Ryckelynck, Michael; Unrau, Peter J
2018-02-13
Despite having many key roles in cellular biology, directly imaging biologically important RNAs has been hindered by a lack of fluorescent tools equivalent to the fluorescent proteins available to study cellular proteins. Ideal RNA labelling systems must preserve biological function, have photophysical properties similar to existing fluorescent proteins, and be compatible with established live and fixed cell protein labelling strategies. Here, we report a microfluidics-based selection of three new high-affinity RNA Mango fluorogenic aptamers. Two of these are as bright or brighter than enhanced GFP when bound to TO1-Biotin. Furthermore, we show that the new Mangos can accurately image the subcellular localization of three small non-coding RNAs (5S, U6, and a box C/D scaRNA) in fixed and live mammalian cells. These new aptamers have many potential applications to study RNA function and dynamics both in vitro and in mammalian cells.
Expression of a novel non-coding mitochondrial RNA in human proliferating cells.
Villegas, Jaime; Burzio, Veronica; Villota, Claudio; Landerer, Eduardo; Martinez, Ronny; Santander, Marcela; Martinez, Rodrigo; Pinto, Rodrigo; Vera, María I; Boccardo, Enrique; Villa, Luisa L; Burzio, Luis O
2007-01-01
Previously, we reported the presence in mouse cells of a mitochondrial RNA which contains an inverted repeat (IR) of 121 nucleotides (nt) covalently linked to the 5' end of the mitochondrial 16S RNA (16S mtrRNA). Here, we report the structure of an equivalent transcript of 2374 nt which is over-expressed in human proliferating cells but not in resting cells. The transcript contains a hairpin structure comprising an IR of 815 nt linked to the 5' end of the 16S mtrRNA and forming a long double-stranded structure or stem and a loop of 40 nt. The stem is resistant to RNase A and can be detected and isolated after digestion with the enzyme. This novel transcript is a non-coding RNA (ncRNA) and several evidences suggest that the transcript is synthesized in mitochondria. The expression of this transcript can be induced in resting lymphocytes stimulated with phytohaemagglutinin (PHA). Moreover, aphidicolin treatment of DU145 cells reversibly blocks proliferation and expression of the transcript. If the drug is removed, the cells re-assume proliferation and over-express the ncmtRNA. These results suggest that the expression of the ncmtRNA correlates with the replicative state of the cell and it may play a role in cell proliferation.
Energy Technology Data Exchange (ETDEWEB)
Kumar, A.; Abdou, M.A.; Eggleston, J. [Univ. of California, Los Angeles, CA (United States). School of Engineering and Applied Science; Kugel, H.W.; Ascione, G.; Elwood, S. [Princeton Plasma Physics Lab., NJ (United States)
1995-12-31
TFTR test cell has a major penetration that exists on the inner side of the north-west corner of the test cell wall. There is no known direct line of sight from the plasma source to the door. The biological doses were measured at various locations inside both the legs of the labyrinth, in addition to, scattered locations on outer segments of the labyrinth facing the TFTR D-T plasma. In addition, six sets of activation foil detectors were placed covering the south-to-north leg, labyrinth entrance, and west-to-east leg. The doses have been calculated using 3-D monte carlo code MCNP with ENDF/B-VI. The calculations reproduce the broad features of the measured data. However, the authors find significant differences in the calculations for a number of locations. The possible reasons for the differences are discussed.
Calculation of absorbed dose of anchorage-dependent cells from internal beta-rays irradiation
International Nuclear Information System (INIS)
Chen Jianwei; Huang Gang; Li Shijun
2001-01-01
Objective: To elicit the formula of internal dosimetry in anchorage-dependent cells by beta-emitting radionuclides from uniformly distributed volume sources. Methods: By means of the definition of absorbed dose and the MIRD (Medical International Radiation Dose) scheme the formula of internal dosimetry was reasonably deduced. Firstly, studying the systems of suspension culture cells. Then, taking account of the speciality of the systems of the anchorage-dependent cells and the directions of irradiation, the absorbed dose of anchorage -dependent cells was calculated by the accumulated radioactivity, beta-ray energy, and the volume of the cultured systems. Results: The formula of internal dosimetry of suspension culture cells and anchorage-dependent cells were achieved. At the same time, the formula of internal dosimetry of suspension culture cells was compared with that of MIRD and was confirmed accurate. Conclusion: The formula of internal dosimetry is concise, reliable and accurate
Energy Technology Data Exchange (ETDEWEB)
Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A. [and others
1995-12-31
In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.
Liu, Qian; Sun, Shanquan; Yu, Wei; Jiang, Jin; Zhuo, Fei; Qiu, Guoping; Xu, Shiye; Jiang, Xuli
2015-04-01
Long non-coding RNAs (lncRNAs), a recently discovered class of non-coding genes, are transcribed throughout the genome. Emerging evidence suggests that lncRNAs may be involved in modulating various aspects of tumor biology, including regulating gene activity in response to external stimuli or DNA damage. No data are available regarding the expression of lncRNAs during genotoxic stress-induced apoptosis and/or necrosis in human glioma cells. In this study, we detected a change in the expression of specific candidate lncRNAs (neat1, GAS5, TUG1, BC200, Malat1, MEG3, MIR155HG, PAR5, and ST7OT1) during DNA damage-induced apoptosis in human glioma cell lines (U251 and U87) using doxorubicin (DOX) and resveratrol (RES). We also detected the expression pattern of these lncRNAs in human glioma cell lines under necrosis induced using an increased dose of DOX. Our results reveal that the lncRNA expression patterns are distinct between genotoxic stress-induced apoptosis and necrosis in human glioma cells. The sets of lncRNA expressed during genotoxic stress-induced apoptosis were DNA-damaging agent-specific. Generally, MEG3 and ST7OT1 are up-regulated in both cell lines under apoptosis induced using both agents. The induction of GAS5 is only clearly detected during DOX-induced apoptosis, whereas the up-regulation of neat1 and MIR155HG is only found during RES-induced apoptosis in both cell lines. However, TUG1, BC200 and MIR155HG are down regulated when necrosis is induced using a high dose of DOX in both cell lines. In conclusion, our findings suggest that the distinct regulation of lncRNAs may possibly involve in the process of cellular defense against genotoxic agents.
Particle-in-Cell Code BEAMPATH for Beam Dynamics Simulations in Linear Accelerators and Beamlines
Energy Technology Data Exchange (ETDEWEB)
Batygin, Y.
2004-10-28
A code library BEAMPATH for 2 - dimensional and 3 - dimensional space charge dominated beam dynamics study in linear particle accelerators and beam transport lines is developed. The program is used for particle-in-cell simulation of axial-symmetric, quadrupole-symmetric and z-uniform beams in a channel containing RF gaps, radio-frequency quadrupoles, multipole lenses, solenoids and bending magnets. The programming method includes hierarchical program design using program-independent modules and a flexible combination of modules to provide the most effective version of the structure for every specific case of simulation. Numerical techniques as well as the results of beam dynamics studies are presented.
Long Non-Coding RNA MEG3 Inhibits Cell Proliferation and Induces Apoptosis in Prostate Cancer
Directory of Open Access Journals (Sweden)
Gang Luo
2015-11-01
Full Text Available Background/Aims: Long non-coding RNAs (lncRNAs play important roles in diverse biological processes, such as cell growth, apoptosis and migration. Although downregulation of lncRNA maternally expressed gene 3 (MEG3 has been identified in several cancers, little is known about its role in prostate cancer progression. The aim of this study was to detect MEG3 expression in clinical prostate cancer tissues, investigate its biological functions in the development of prostate cancer and the underlying mechanism. Methods: MEG3 expression levels were detected by qRT-PCR in both tumor tissues and adjacent non-tumor tissues from 21 prostate cancer patients. The effects of MEG3 on PC3 and DU145 cells were assessed by MTT assay, colony formation assay, western blot and flow cytometry. Transfected PC3 cells were transplanted into nude mice, and the tumor growth curves were determined. Results: MEG3 decreased significantly in prostate cancer tissues relative to adjacent normal tissues. MEG3 inhibited intrinsic cell survival pathway in vitro and in vivo by reducing the protein expression of Bcl-2, enhancing Bax and activating caspase 3. We further demonstrated that MEG3 inhibited the expression of cell cycle regulatory protein Cyclin D1 and induced cell cycle arrest in G0/G1 phase. Conclusions: Our study presents an important role of MEG3 in the molecular etiology of prostate cancer and implicates the potential application of MEG3 in prostate cancer therapy.
Non-coding RNA regulation in pathogenic bacteria located inside eukaryotic cells.
Ortega, Alvaro D; Quereda, Juan J; Pucciarelli, M Graciela; García-del Portillo, Francisco
2014-01-01
Intracellular bacterial pathogens have evolved distinct lifestyles inside eukaryotic cells. Some pathogens coexist with the infected cell in an obligate intracellular state, whereas others transit between the extracellular and intracellular environment. Adaptation to these intracellular lifestyles is regulated in both space and time. Non-coding small RNAs (sRNAs) are post-transcriptional regulatory molecules that fine-tune important processes in bacterial physiology including cell envelope architecture, intermediate metabolism, bacterial communication, biofilm formation, and virulence. Recent studies have shown production of defined sRNA species by intracellular bacteria located inside eukaryotic cells. The molecules targeted by these sRNAs and their expression dynamics along the intracellular infection cycle remain, however, poorly characterized. Technical difficulties linked to the isolation of "intact" intracellular bacteria from infected host cells might explain why sRNA regulation in these specialized pathogens is still a largely unexplored field. Transition from the extracellular to the intracellular lifestyle provides an ideal scenario in which regulatory sRNAs are intended to participate; so much work must be done in this direction. This review focuses on sRNAs expressed by intracellular bacterial pathogens during the infection of eukaryotic cells, strategies used with these pathogens to identify sRNAs required for virulence, and the experimental technical challenges associated to this type of studies. We also discuss varied techniques for their potential application to study RNA regulation in intracellular bacterial infections.
International Nuclear Information System (INIS)
Pantazi, D.; Mateescu, S.; Stanciu, M.; Mete, M.
2001-01-01
The modulated code system SCALE is used to perform a standardized shielding analysis for any facility containing spent fuel: handling devices, transport cask, intermediate and final storage facility. The neutron and gamma sources as well as the dose rates can be obtained using either discrete-ordinates or Monte Carlo methods. The shielding analysis control modules (SAS1, SAS2H and SAS4) provide a general procedure for cross-section preparation, fuel depletion/decay calculation and general onedimensional or multi-dimensional shielding analysis. The module SAS4 used in the analysis presented in this paper, is a three-dimensional Monte Carlo shielding analysis module, which uses an automated biasing procedure specialized for a nuclear fuel transport or storage container. The Spent Fuel Interim Storage Facility in our country is projected to be a parallelepiped concrete monolithic module, consisting of an external reinforced concrete structure with vertical storage cylinders (pits) arranged in a rectangular array. A pit is filled with sealed cylindrical baskets of stainless steel arranged in a stack, and with each basket containing spent fuel bundles in vertical position. The pit is closed with a concrete plug. The cylindrical geometry model is used in the shielding evaluation for a spent fuel storage structure (pit), and only the active parts of the superposed bundles is considered. The dose rates have been calculated in both the axial and radial directions using SAS4.(author)
Directory of Open Access Journals (Sweden)
Gholamzadeh Zohreh
2014-12-01
Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view
International Nuclear Information System (INIS)
Gomes, Renato G.; Rebello, Wilson F.; Cavaliere, Marcos Paulo; Vellozo, Sergio O.; Moreira Junior, Luis; Vital, Helio C.; Silva, Ademir X.
2013-01-01
Using the MCNPX code, the objective was to calculate by means of computer simulation spectroscopy range inside the irradiation chamber upper radiator gamma research irradiating facility Army Technology Center (CTEx). The calculations were performed in the spectral range usual 2 points for research purposes irradiating the energy spectra of gamma rays from the source of Cesium chloride 137. Sought the discretization of the spectrum in 100 channels at points of upper bound of 1cm higher and lower dose rates previously known. It was also conducted in the laboratory lifting the spectrum of Cesium-137 source using NaI scintillator detector and multichannel analyzer. With the source spectrum Cesium-137 contained in the literature and raised in the laboratory, both used as reference for comparison and analysis in terms of probability of emission maximum of 0.661 MeV The spectra were quite consistent in terms of the behavior of the energy distributions with scores. The position of maximum dose rate showed absorption detection almost maximum energy of 0.661 MeV photopeak In the spectrum of the position of minimum dosage rate, it was found that due to the removal of the source point of interest, some loss detection were caused by Compton scattering. (author)
International Nuclear Information System (INIS)
Nassini, H.; Meyder, R.
1988-11-01
The calculations were performed with SSYST-3 code system. The experiments have been carried out with shortened REBEKA simulators. The main goal was to verify whether the model for describing the Zircaloy cladding deformation and rupture at high temperatures (NORA2 model) is able to predict the Argentine cladding material behaviour under such conditions. A good agreement between experimental results and model predictions was found in the range of temperatures from 700 to 850 0 C (in α phase and first half of α+β transition phase). There were some discrepances at higher temperatures, probably due to the strong influence of heating rate on material properties. According to the results of these post-test calculations, it can be concluded that NORA2 model is able to well describe the Argentine Zircaloy cladding deformation and rupture under LOCA conditions, because there was a good agreement in the range of burst (ballooning) temperatures which are expected in this type of accident. (orig./HP) [de
Kasprzak, A A; Szmyt, M; Malkowski, W; Surdyk-Zasada, J; Przybyszewska, W; Szmeja, J; Helak-Łapaj, C; Seraszek-Jaros, A; Kaczmarek, E
2013-12-01
The study aimed at quantitative analysis of expression involving markers of mast cells (tryptase), monocytes/macrophages (CD68 molecule) and dendritic cells (S100 protein) in gallbladder mucosa with acute and chronic calculous cholecystitis. Routinely prepared tissue material from the patients with acute (ACC) (n = 16) and chronic calculous cholecystitis (CCC) (n = 55) was evaluated. Three cellular markers were localized by immunocytochemistry. Their expression was quantified using spatial visualization technique. The expression of tryptase was similar in acute and chronic cholecystitis. CD68 expression in ACC was significantly higher than in the CCC group. Expression of S100 protein was significantly higher in CCC as compared to the ACC group. No significant correlations were disclosed between expression of studied markers and grading in the gallbladder wall. A weak negative correlation was noted between expression of CD68 and number of gallstones in the CCC group. The spatial visualization technique allowed for a credible quantitative evaluation of expression involving markers of mast cells (MCs), monocytes/macrophages (Mo/Ma) and dendritic cells (DCs) in gallbladder mucosa with ACC and CCC. For the first time mucosal expression of S100 protein-positive DCs was evaluated in calculous cholecystitis. The results point to distinct functions of studied cell types in the non-specific immune response in calculous cholecystitis.
The calculation of deep levels in semiconductors by using a recursion method for super-cells
International Nuclear Information System (INIS)
Wong Yongliang.
1987-01-01
The paper presents the theory of deep levels in semiconductors, the super-cell approach to the theory of deep level impurities, the calculation of band structure by using the tight-binding method and the recursion method used to study the defects in the presence of lattice relaxation and extended defect complexes. 47 refs
International Nuclear Information System (INIS)
Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A.
1995-01-01
In conformity with the protocol of the Workshop under Contract open-quotes Assessment of RBMK reactor safety using modern Western Codesclose quotes VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core
Directory of Open Access Journals (Sweden)
Subrata Sarker
2014-01-01
Full Text Available Electrochemical impedance spectroscopy (EIS is one of the most important tools to elucidate the charge transfer and transport processes in various electrochemical systems including dye-sensitized solar cells (DSSCs. Even though there are many books and reports on EIS, it is often very difficult to explain the EIS spectra of DSSCs. Understanding EIS through calculating EIS spectra on spreadsheet can be a powerful approach as the user, without having any programming knowledge, can go through each step of calculation on a spreadsheet and get instant feedback by visualizing the calculated results or plot on the same spreadsheet. Here, a brief account of the EIS of DSSCs is given with fundamental aspects and their spreadsheet calculation. The review should help one to develop a basic understanding about EIS of DSSCs through interacting with spreadsheet.
Energy Technology Data Exchange (ETDEWEB)
De Carlan, L.; Bordy, J.M.; Gouriou, J. [CEA Saclay, LIST, Laboratoire National Henri Becquerel, Laboratoire de Metrologie de la Dose 91 - Gif-sur-Yvette (France)
2010-07-01
Within the frame of the CONRAD European project (Coordination Network for Radiation Dosimetry), and more precisely within a work group paying attention to uncertainty assessment in computational dosimetry and aiming at comparing different approaches, the authors report the simulation of an irradiator containing a caesium 137 source to calculate the kerma in air as well as its uncertainty due to different parameters. They present the problem geometry, recall the studied issues (kerma uncertainty, influence of capsule source, influence of the collimator, influence of the air volume surrounding the source). They indicate the codes which have been used (MNCP, Fluka, Penelope, etc.) and discuss the obtained results for the first issue
Investigation of the influence of the open cell foam models geometry on hydrodynamic calculation
Soloveva, O. V.; Solovev, S. A.; Khusainov, R. R.; Popkova, O. S.; Panenko, D. O.
2018-01-01
A geometrical model of the open cell foam was created as an ordered set of intersecting spheres. The proposed model closely describes a real porous cellular structure. The hydrodynamics flow was calculated on the basis of a simple model in the ANSYS Fluent software package. A pressure drop was determined, the value of which was compared with the experimental data of other authors. As a result of the conducted studies, we found that a porous structure with smoothed faces provides the smallest pressure drop with the same porosity of the package. Analysis of the calculated data demonstrated that the approximation of an elementary porous cell substantially distorts the flow field. This is undesirable in detailed modeling of the open cell foam.
Translation of ARAC computer codes
International Nuclear Information System (INIS)
Takahashi, Kunio; Chino, Masamichi; Honma, Toshimitsu; Ishikawa, Hirohiko; Kai, Michiaki; Imai, Kazuhiko; Asai, Kiyoshi
1982-05-01
In 1981 we have translated the famous MATHEW, ADPIC and their auxiliary computer codes for CDC 7600 computer version to FACOM M-200's. The codes consist of a part of the Atmospheric Release Advisory Capability (ARAC) system of Lawrence Livermore National Laboratory (LLNL). The MATHEW is a code for three-dimensional wind field analysis. Using observed data, it calculates the mass-consistent wind field of grid cells by a variational method. The ADPIC is a code for three-dimensional concentration prediction of gases and particulates released to the atmosphere. It calculates concentrations in grid cells by the particle-in-cell method. They are written in LLLTRAN, i.e., LLNL Fortran language and are implemented on the CDC 7600 computers of LLNL. In this report, i) the computational methods of the MATHEW/ADPIC and their auxiliary codes, ii) comparisons of the calculated results with our JAERI particle-in-cell, and gaussian plume models, iii) translation procedures from the CDC version to FACOM M-200's, are described. Under the permission of LLNL G-Division, this report is published to keep the track of the translation procedures and to serve our JAERI researchers for comparisons and references of their works. (author)
Non-coding RNAs as epigenetic regulator of glioma stem-like cell differentiation
Directory of Open Access Journals (Sweden)
Keisuke eKatsushima
2014-02-01
Full Text Available Glioblastomas show heterogeneous histological features. These distinct phenotypic states are thought to be associated with the presence of glioma stem cells (GSCs, which are highly tumorigenic and self-renewing sub-population of tumor cells that have different functional characteristics. Differentiation of GSCs may be regulated by multi-tiered epigenetic mechanisms that orchestrate the expression of thousands of genes. One such regulatory mechanism involves functional non-coding RNAs (ncRNAs, such as microRNAs (miRNAs; a large number of ncRNAs have been identified and shown to regulate the expression of genes associated with cell differentiation programs. Given the roles of miRNAs in cell differentiation, it is possible they are involved in the regulation of gene expression networks in GSCs that are important for the maintenance of the pluripotent state and for directing differentiation. Here, we review recent findings on ncRNAs associated with GSC differentiation and discuss how these ncRNAs contribute to the establishment of tissue heterogeneity during glioblastoma tumor formation.
Selective expression of long non-coding RNAs in a breast cancer cell progression model.
Tracy, Kirsten M; Tye, Coralee E; Page, Natalie A; Fritz, Andrew J; Stein, Janet L; Lian, Jane B; Stein, Gary S
2018-02-01
Long non-coding RNAs (lncRNAs) are acknowledged as regulators of cancer biology and pathology. Our goal was to perform a stringent profiling of breast cancer cell lines that represent disease progression. We used the MCF-10 series, which includes the normal-like MCF-10A, HRAS-transformed MCF-10AT1 (pre-malignant), and MCF-10CA1a (malignant) cells, to perform transcriptome wide sequencing. From these data, we have identified 346 lncRNAs with dysregulated expression across the progression series. By comparing lncRNAs from these datasets to those from an additional set of cell lines that represent different disease stages and subtypes, MCF-7 (early stage, luminal), and MDA-MB-231 (late stage, basal), 61 lncRNAs that are associated with breast cancer progression were identified. Querying breast cancer patient data from The Cancer Genome Atlas, we selected a lncRNA, IGF-like family member 2 antisense RNA 1 (IGFL2-AS1), of potential clinical relevance for functional characterization. Among the 61 lncRNAs, IGFL2-AS1 was the most significantly decreased. Our results indicate that this lncRNA plays a role in downregulating its nearest neighbor, IGFL1, and affects migration of breast cancer cells. Furthermore, the lncRNAs we identified provide a valuable resource to mechanistically and clinically understand the contribution of lncRNAs in breast cancer progression. © 2017 Wiley Periodicals, Inc.
SHARP: A Spatially Higher-order, Relativistic Particle-in-cell Code
Energy Technology Data Exchange (ETDEWEB)
Shalaby, Mohamad; Broderick, Avery E. [Department of Physics and Astronomy, University of Waterloo, 200 University Avenue West, Waterloo, ON, N2L 3G1 (Canada); Chang, Philip [Department of Physics, University of Wisconsin-Milwaukee, 1900 E. Kenwood Boulevard, Milwaukee, WI 53211 (United States); Pfrommer, Christoph [Leibniz-Institut für Astrophysik Potsdam (AIP), An der Sternwarte 16, D-14482 Potsdam (Germany); Lamberts, Astrid [Theoretical Astrophysics, California Institute of Technology, Pasadena, CA 91125 (United States); Puchwein, Ewald, E-mail: mshalaby@live.ca [Institute of Astronomy and Kavli Institute for Cosmology, University of Cambridge, Madingley Road, Cambridge, CB3 0HA (United Kingdom)
2017-05-20
Numerical heating in particle-in-cell (PIC) codes currently precludes the accurate simulation of cold, relativistic plasma over long periods, severely limiting their applications in astrophysical environments. We present a spatially higher-order accurate relativistic PIC algorithm in one spatial dimension, which conserves charge and momentum exactly. We utilize the smoothness implied by the usage of higher-order interpolation functions to achieve a spatially higher-order accurate algorithm (up to the fifth order). We validate our algorithm against several test problems—thermal stability of stationary plasma, stability of linear plasma waves, and two-stream instability in the relativistic and non-relativistic regimes. Comparing our simulations to exact solutions of the dispersion relations, we demonstrate that SHARP can quantitatively reproduce important kinetic features of the linear regime. Our simulations have a superior ability to control energy non-conservation and avoid numerical heating in comparison to common second-order schemes. We provide a natural definition for convergence of a general PIC algorithm: the complement of physical modes captured by the simulation, i.e., those that lie above the Poisson noise, must grow commensurately with the resolution. This implies that it is necessary to simultaneously increase the number of particles per cell and decrease the cell size. We demonstrate that traditional ways for testing for convergence fail, leading to plateauing of the energy error. This new PIC code enables us to faithfully study the long-term evolution of plasma problems that require absolute control of the energy and momentum conservation.
JH, Summerfield; MW, Manley
2016-01-01
A simple simulation of chemical species movement is presented. The species traverse a Nafion membrane in a fuel cell. Three cells are examined: direct methanol, direct ethanol, and direct glucose. The species are tracked using excess proton concentration, electric field strength, and voltage. The Matlab computer code is provided.
SWAT2: The improved SWAT code system by incorporating the continuous energy Monte Carlo code MVP
International Nuclear Information System (INIS)
Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi
2003-01-01
SWAT is a code system, which performs the burnup calculation by the combination of the neutronics calculation code, SRAC95 and the one group burnup calculation code, ORIGEN2.1. The SWAT code system can deal with the cell geometry in SRAC95. However, a precise treatment of resonance absorptions by the SRAC95 code using the ultra-fine group cross section library is not directly applicable to two- or three-dimensional geometry models, because of restrictions in SRAC95. To overcome this problem, SWAT2 which newly introduced the continuous energy Monte Carlo code, MVP into SWAT was developed. Thereby, the burnup calculation by the continuous energy in any geometry became possible. Moreover, using the 147 group cross section library called SWAT library, the reactions which are not dealt with by SRAC95 and MVP can be treated. OECD/NEA burnup credit criticality safety benchmark problems Phase-IB (PWR, a single pin cell model) and Phase-IIIB (BWR, fuel assembly model) were calculated as a verification of SWAT2, and the results were compared with the average values of calculation results of burnup calculation code of each organization. Through two benchmark problems, it was confirmed that SWAT2 was applicable to the burnup calculation of the complicated geometry. (author)
An electrostatic Particle-In-Cell code on multi-block structured meshes
Meierbachtol, Collin S.; Svyatskiy, Daniil; Delzanno, Gian Luca; Vernon, Louis J.; Moulton, J. David
2017-12-01
We present an electrostatic Particle-In-Cell (PIC) code on multi-block, locally structured, curvilinear meshes called Curvilinear PIC (CPIC). Multi-block meshes are essential to capture complex geometries accurately and with good mesh quality, something that would not be possible with single-block structured meshes that are often used in PIC and for which CPIC was initially developed. Despite the structured nature of the individual blocks, multi-block meshes resemble unstructured meshes in a global sense and introduce several new challenges, such as the presence of discontinuities in the mesh properties and coordinate orientation changes across adjacent blocks, and polyjunction points where an arbitrary number of blocks meet. In CPIC, these challenges have been met by an approach that features: (1) a curvilinear formulation of the PIC method: each mesh block is mapped from the physical space, where the mesh is curvilinear and arbitrarily distorted, to the logical space, where the mesh is uniform and Cartesian on the unit cube; (2) a mimetic discretization of Poisson's equation suitable for multi-block meshes; and (3) a hybrid (logical-space position/physical-space velocity), asynchronous particle mover that mitigates the performance degradation created by the necessity to track particles as they move across blocks. The numerical accuracy of CPIC was verified using two standard plasma-material interaction tests, which demonstrate good agreement with the corresponding analytic solutions. Compared to PIC codes on unstructured meshes, which have also been used for their flexibility in handling complex geometries but whose performance suffers from issues associated with data locality and indirect data access patterns, PIC codes on multi-block structured meshes may offer the best compromise for capturing complex geometries while also maintaining solution accuracy and computational efficiency.
International Nuclear Information System (INIS)
Abdallah, A. M.; Elsherbiny, E. M.; Sobhy, M.
1995-01-01
The P n -spatial expansion method has been used for calculating the one speed transport utilization factor in heterogenous slab cells in which neutrons may scatter anisotropically; by considering the P 1- approximation with a two-term scattering kernel in both the fuel and moderator regions, an analytical expression for the disadvantage factor has been derived. The numerical results obtained have been shown to be much better than those calculated by the usual P 1- and P 3- approximations and comparable with those obtained by some exact methods. 3 tabs
International Nuclear Information System (INIS)
Santos, Rubens Souza dos; Martinez, Aquilino Senra; Alvim, Antonio Carlos Marques
2002-01-01
In this work is presented a methodology which focuses the distribution of neutron absorber rods in nuclear reactor power plants, for utilizing in space kinetic calculations, principally in the cluster ejection transients of control rods. A numerical model for macroscopic constant calculations based on the knowledge of the neutron flux without the control rods is proposed, as alternative to the analytical models, based on the hypothesis of the null current on the cell super boundaries. The proposed model in this work has itself showed adequate to deal with problems with strong space dependence, once that the model showed consistence in the global average built in the analytical model. (author)
Calculation of cell face velocity of non-staggered grid system
Li, Wang
2012-07-28
In this paper, the cell face velocities in the discretization of the continuity equation, the momentum equation, and the scalar equation of a non-staggered grid system are calculated and discussed. Both the momentum interpolation and the linear interpolation are adopted to evaluate the coefficients in the discretized momentum and scalar equations. Their performances are compared. When the linear interpolation is used to calculate the coefficients, the mass residual term in the coefficients must be dropped to maintain the accuracy and convergence rate of the solution. © Shanghai University and Springer-Verlag Berlin Heidelberg 2012.
Spacecraft charging analysis with the implicit particle-in-cell code iPic3D
Energy Technology Data Exchange (ETDEWEB)
Deca, J.; Lapenta, G. [Centre for Mathematical Plasma Astrophysics, KU Leuven, Celestijnenlaan 200B bus 2400, 3001 Leuven (Belgium); Marchand, R. [Department of Physics, University of Alberta, Edmonton, Alberta T6G 2J1 (Canada); Markidis, S. [High Performance Computing and Visualization Department, KTH Royal Institute of Technology, Stockholm (Sweden)
2013-10-15
We present the first results on the analysis of spacecraft charging with the implicit particle-in-cell code iPic3D, designed for running on massively parallel supercomputers. The numerical algorithm is presented, highlighting the implementation of the electrostatic solver and the immersed boundary algorithm; the latter which creates the possibility to handle complex spacecraft geometries. As a first step in the verification process, a comparison is made between the floating potential obtained with iPic3D and with Orbital Motion Limited theory for a spherical particle in a uniform stationary plasma. Second, the numerical model is verified for a CubeSat benchmark by comparing simulation results with those of PTetra for space environment conditions with increasing levels of complexity. In particular, we consider spacecraft charging from plasma particle collection, photoelectron and secondary electron emission. The influence of a background magnetic field on the floating potential profile near the spacecraft is also considered. Although the numerical approaches in iPic3D and PTetra are rather different, good agreement is found between the two models, raising the level of confidence in both codes to predict and evaluate the complex plasma environment around spacecraft.
Energy Technology Data Exchange (ETDEWEB)
Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2013-10-15
The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)
Bobin, C; Thiam, C; Bouchard, J
2016-03-01
At LNE-LNHB, a liquid scintillation (LS) detection setup designed for Triple to Double Coincidence Ratio (TDCR) measurements is also used in the β-channel of a 4π(LS)β-γ coincidence system. This LS counter based on 3 photomultipliers was first modeled using the Monte Carlo code Geant4 to enable the simulation of optical photons produced by scintillation and Cerenkov effects. This stochastic modeling was especially designed for the calculation of double and triple coincidences between photomultipliers in TDCR measurements. In the present paper, this TDCR-Geant4 model is extended to 4π(LS)β-γ coincidence counting to enable the simulation of the efficiency-extrapolation technique by the addition of a γ-channel. This simulation tool aims at the prediction of systematic biases in activity determination due to eventual non-linearity of efficiency-extrapolation curves. First results are described in the case of the standardization (59)Fe. The variation of the γ-efficiency in the β-channel due to the Cerenkov emission is investigated in the case of the activity measurements of (54)Mn. The problem of the non-linearity between β-efficiencies is featured in the case of the efficiency tracing technique for the activity measurements of (14)C using (60)Co as a tracer. Copyright © 2015 Elsevier Ltd. All rights reserved.
Understanding the regulation of coding and noncoding transcription in cell populations.
Beilharz, Traude Helene
2016-05-01
Whole transcriptome analyses have unveiled the uncomfortable truth that we know less about how transcription is regulated then we thought. In addition to its role in classic promoter-driven transcription of coding RNA, it is now clear that RNA Pol II also drives abundant expression of noncoding RNA. For the majority of this the functional significance remains unclear. Moreover, its regulation and impact are hard to predict because it often proceeds in unexpected ways from cryptic promoters, including by driving convergent antisense transcription from within 3' UTRs. This review suggests that its time to rethink how we envisage gene expression by inclusion of the regulatory architecture of the full genetic locus, and expanding our thinking to encompass the fact that we generally study cells within heterogeneous populations.
Numerical simulation of Ge solar cells using D-AMPS-1D code
Energy Technology Data Exchange (ETDEWEB)
Barrera, Marcela, E-mail: barrera@tandar.cnea.gov.ar [Comision Nacional de Energia Atomica, Avenida General Paz 1499, San Martin 1650, Buenos Aires (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina); Rubinelli, Francisco [Instituto de Desarrollo Tecnologico para la Industria Quimica (INTEC)-CONICET, Gueemes 3450, Santa Fe 3000 (Argentina); Rey-Stolle, Ignacio [Instituto de Energia Solar, Universidad Politecnica de Madrid, Avenida Complutense 30, Madrid 28040 (Spain); Pla, Juan [Comision Nacional de Energia Atomica, Avenida General Paz 1499, San Martin 1650, Buenos Aires (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina)
2012-08-15
A solar cell is a solid state device that converts the energy of sunlight directly into electricity by the photovoltaic effect. When light with photon energies greater than the band gap is absorbed by a semiconductor material, free electrons and free holes are generated by optical excitation in the material. The main characteristic of a photovoltaic device is the presence of internal electric field able to separate the free electrons and holes so they can pass out of the material to the external circuit before they recombine. Numerical simulation of photovoltaic devices plays a crucial role in their design, performance prediction, and comprehension of the fundamental phenomena ruling their operation. The electrical transport and the optical behavior of the solar cells discussed in this work were studied with the simulation code D-AMPS-1D. This software is an updated version of the one-dimensional (1D) simulation program Analysis of Microelectronic and Photonic Devices (AMPS) that was initially developed at The Penn State University, USA. Structures such as homojunctions, heterojunctions, multijunctions, etc., resulting from stacking layers of different materials can be studied by appropriately selecting characteristic parameters. In this work, examples of cells simulation made with D-AMPS-1D are shown. Particularly, results of Ge photovoltaic devices are presented. The role of the InGaP buffer on the device was studied. Moreover, a comparison of the simulated electrical parameters with experimental results was performed.
Energy Technology Data Exchange (ETDEWEB)
Vasko, Marek; Daniska, Vladimir; Ondra, Frantisek; Bezak, Peter; Kristofova, Kristina; Tatransky, Peter; Zachar, Matej [DECOM Slovakia, spol. s.r.o., J. Bottu 2, SK-917 01 Trnava (Slovakia); Lindskog, Staffan [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)
2007-03-15
The presented report is focused on tentative calculations of basic decommissioning parameters such as costs, manpower and exposure of personnel for activities of older nuclear facility decommissioning in Sweden represented by Intermediate storage facility for spent fuel in Studsvik, by means of calculation code OMEGA. This report continuously follows up two previous projects, which described methodology of cost estimates of decommissioning with an emphasis to derive cost functions for alpha contaminated material and implementation of the advanced decommissioning costing methodology for Intermediate Storage facility for Spent Fuel in Studsvik. The main purpose of the presented study is to demonstrate the trial application of the advanced costing methodology using OMEGA code for Intermediate Storage Facility for Spent Fuel in Studsvik. Basic work packages presented in report are as follows: 1. Analysis and validation input data on Intermediate Storage Facility for Spent Fuel and assemble a database suitable for standardised decommissioning cost calculations including radiological parameters, 2. Proposal of range of decommissioning calculations and define an extent of decommissioning activities, 3. Defining waste management scenarios for particular material waste streams from Intermediate Storage Facility for Spent Fuel, 4. Developing standardised cost calculation structure applied for Intermediate Storage Facility for Spent Fuel decommissioning calculation and 5. Performing tentative decommissioning calculations for Intermediate Storage Facility for Spent Fuel by OMEGA code. Calculated parameters of decommissioning are presented in structure according to Proposed Standardized List of Items for Costing Purposes. All parameters are documented and summed up in both table and graphic forms in text and Annexes. The presented report documents availability and applicability of methodology for evaluation of costs and other parameters of decommissioning in a form implemented
International Nuclear Information System (INIS)
Vasko, Marek; Daniska, Vladimir; Ondra, Frantisek; Bezak, Peter; Kristofova, Kristina; Tatransky, Peter; Zachar, Matej; Lindskog, Staffan
2007-03-01
The presented report is focused on tentative calculations of basic decommissioning parameters such as costs, manpower and exposure of personnel for activities of older nuclear facility decommissioning in Sweden represented by Intermediate storage facility for spent fuel in Studsvik, by means of calculation code OMEGA. This report continuously follows up two previous projects, which described methodology of cost estimates of decommissioning with an emphasis to derive cost functions for alpha contaminated material and implementation of the advanced decommissioning costing methodology for Intermediate Storage facility for Spent Fuel in Studsvik. The main purpose of the presented study is to demonstrate the trial application of the advanced costing methodology using OMEGA code for Intermediate Storage Facility for Spent Fuel in Studsvik. Basic work packages presented in report are as follows: 1. Analysis and validation input data on Intermediate Storage Facility for Spent Fuel and assemble a database suitable for standardised decommissioning cost calculations including radiological parameters, 2. Proposal of range of decommissioning calculations and define an extent of decommissioning activities, 3. Defining waste management scenarios for particular material waste streams from Intermediate Storage Facility for Spent Fuel, 4. Developing standardised cost calculation structure applied for Intermediate Storage Facility for Spent Fuel decommissioning calculation and 5. Performing tentative decommissioning calculations for Intermediate Storage Facility for Spent Fuel by OMEGA code. Calculated parameters of decommissioning are presented in structure according to Proposed Standardized List of Items for Costing Purposes. All parameters are documented and summed up in both table and graphic forms in text and Annexes. The presented report documents availability and applicability of methodology for evaluation of costs and other parameters of decommissioning in a form implemented
Tang, Xing; Hou, Mei; Ding, Yang; Li, Zhaohui; Ren, Lichen; Gao, Ge
2013-01-01
While more and more long intergenic non-coding RNAs (lincRNAs) were identified to take important roles in both maintaining pluripotency and regulating differentiation, how these lincRNAs may define and drive cell fate decisions on a global scale are still mostly elusive. Systematical profiling and comprehensive annotation of embryonic stem cells lincRNAs may not only bring a clearer big picture of these novel regulators but also shed light on their functionalities. Based on multiple RNA-Seq datasets, we systematically identified 300 human embryonic stem cell lincRNAs (hES lincRNAs). Of which, one forth (78 out of 300) hES lincRNAs were further identified to be biasedly expressed in human ES cells. Functional analysis showed that they were preferentially involved in several early-development related biological processes. Comparative genomics analysis further suggested that around half of the identified hES lincRNAs were conserved in mouse. To facilitate further investigation of these hES lincRNAs, we constructed an online portal for biologists to access all their sequences and annotations interactively. In addition to navigation through a genome browse interface, users can also locate lincRNAs through an advanced query interface based on both keywords and expression profiles, and analyze results through multiple tools. By integrating multiple RNA-Seq datasets, we systematically characterized and annotated 300 hES lincRNAs. A full functional web portal is available freely at http://scbrowse.cbi.pku.edu.cn. As the first global profiling and annotating of human embryonic stem cell lincRNAs, this work aims to provide a valuable resource for both experimental biologists and bioinformaticians.
Directory of Open Access Journals (Sweden)
Lei Wang
2016-09-01
Full Text Available As the sole output neurons in the retina, ganglion cells play significant roles in transforming visual information into spike trains, and then transmitting them to the higher visual centers. However, coding strategies that retinal ganglion cells (RGCs adopt to accomplish these processes are not completely clear yet. To clarify these issues, we investigate the coding properties of three types of RGCs (repetitive spiking, tonic firing, and phasic firing by two different measures (spike-rate and spike-latency. Model results show that for periodic stimuli, repetitive spiking RGC and tonic RGC exhibit similar spike-rate patterns. Their spike-rates decrease gradually with increased stimulus frequency, moreover, variation of stimulus amplitude would change the two RGCs’ spike-rate patterns. For phasic RGC, it activates strongly at medium levels of frequency when the stimulus amplitude is low. While if high stimulus amplitude is applied, phasic RGC switches to respond strongly at low frequencies. These results suggest that stimulus amplitude is a prominent factor in regulating RGCs in encoding periodic signals. Similar conclusions can be drawn when analyzes spike-latency patterns of the three RGCs. More importantly, the above phenomena can be accurately reproduced by Hodgkin’s three classes of neurons, indicating that RGCs can perform the typical three classes of firing dynamics, depending on the distinctions of ion channel densities. Consequently, model results from the three RGCs may be not specific, but can also applicable to neurons in other brain regions which exhibit part(s or all of the Hodgkin’s three excitabilities.
Cell verification of parallel burnup calculation program MCBMPI based on MPI
International Nuclear Information System (INIS)
Yang Wankui; Liu Yaoguang; Ma Jimin; Wang Guanbo; Yang Xin; She Ding
2014-01-01
The parallel burnup calculation program MCBMPI was developed. The program was modularized. The parallel MCNP5 program MCNP5MPI was employed as neutron transport calculation module. And a composite of three solution methods was used to solve burnup equation, i.e. matrix exponential technique, TTA analytical solution, and Gauss Seidel iteration. MPI parallel zone decomposition strategy was concluded in the program. The program system only consists of MCNP5MPI and burnup subroutine. The latter achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, and data exchanging with MCNP5MPI. Also, the program was verified with the pressurized water reactor (PWR) cell burnup benchmark. The results show that it,s capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)
Particle-in-cell vs straight line Gaussian calculations for an area of complex topography
International Nuclear Information System (INIS)
Lange, R.; Sherman, C.
1977-01-01
Two numerical models for the calculation of time integrated air concentraton and ground deposition of airborne radioactive effluent releases are compared. The time dependent Particle-in-Cell (PIC) model and the steady state Gaussian plume model were used for the simulation. The area selected for the comparison was the Hudson River Valley, New York. Input for the models was synthesized from meteorological data gathered in previous studies by various investigators. It was found that the PIC model more closely simulated the three-dimensional effects of the meteorology and topography. Overall, the Gaussian model calculated higher concentrations under stable conditions. In addition, because of its consideration of exposure from the returning plume after flow reversal, the PIC model calculated air concentrations over larger areas than did the Gaussian model