SLAROM: a code for cell homogenization calculation of fast reactor
International Nuclear Information System (INIS)
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)
Calculations of WWER cells and assemblies by WIMS-7B code
International Nuclear Information System (INIS)
A study of the nuclear data libraries of the WIMS-7B code have been performed in calculations of computational benchmark problems. The benchmarks cover pin cell, single fuel assembly with several different fuel types, moderator densities. Fuel depletion is performed to a burnup of 60 MWd/kgNM in the WWER-1000 pin cell. The results of the analysis of the benchmark with different code systems have been compared and indicated good agreement among the different methods and data. (Authors)
HERMET: cell neutronic calculation code for MTR (materials testing reactors) fuels
International Nuclear Information System (INIS)
The HERMET neutronic calculation code was developed for resolution of systems, at a cell calculation level in one-dimensional plain geometry (MTR), preserving its heterogeneous character with or without reflecting boundary conditions and reducing the cost as regards time and machine-memory. This code also includes the burn-up calculation which may be performed with the critical spectra B0, B1 or the one improved by leakages corresponding to the buckling given by the user. The burn-up scheme may be carried out by a transport equation with intermediate stages without flux reevaluation or by a predictor-corrector scheme. (Author)
Kinetic parameters evaluation of PWRs using static cell and core calculation codes
International Nuclear Information System (INIS)
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 ((βeff)i)/((βeff)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 ((βeff)i)/((βeff)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.
CRX: a transport theory code for cell and assembly calculations based on characteristic method
International Nuclear Information System (INIS)
A transport theory code CRX based on characteristic method with a general geometric tracking routine for rectangular and hexagonal geometrical problems is developed and tested for heterogeneous cell and assembly calculations. Since the characteristic method treats explicitly (analytically) the streaming portion of the transport equation, CRX treats strong absorbers well and has no practical limitations placed on the geometry of the problem. To test the code, it was applied to three benchmark problems which consist of complex meshes and compared with other codes. (author)
SWAT, Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2
International Nuclear Information System (INIS)
1 - Description of program or function: SWAT evaluates isotopic composition of spent nuclear fuel, especially for burnup credit issues by driving codes SRAC95 and ORIGEN2.1 or ORIGEN2. SWAT is an automated driver code system. At the initial development phase, it was constructed by combining source programs of SRAC and ORIGEN2. To overcome the problem associated with code updates, SWAT chose to use system function of UNIX operating system to execute SRAC95 and ORIGEN2. So that, SWAT is independent of development and modification of SRAC95 and ORIGEN2.1. In SWAT, ORIGEN2(82) or ORIGEN2.1 is used for burnup calculations using the matrix exponential method. An updated decay library is included in the distribution. SWAT uses SRAC95 for neutron spectrum and effective cross section calculation in 107 groups, using the collision probability method for given geometry and isotopic composition. One or two dimensional cell geometries are supported in SRAC95. NEA-1698/02: The main purpose of new package is to run SWAT on several machines not supported in previous package (IA64 under Linux, Windows with cygwin and Sun,...) and several commercial FORTRAN compiler (Intel, PGI, Fujitsu). 2 - Methods: In calculating the problem-dependent cross section in SWAT, the total burnup history is divided into 'burnup steps'. Power, boric acid concentration, temperature of each region, and void ratio of coolant are given as history data. For each burnup step, the neutron spectrum and effective cross section are evaluated by SRAC95 using the information given in previous burnup calculation and cell geometry information. The user can select geometry options for the collision probability method in SRAC95. 3 - Restrictions on the complexity of the problem: Resonance absorption calculation with ultra-fine group cross section can not be directly applicable for 2D geometry
Application Of WIMS Code To Calculation Kartini Reactor Parameters By Pin-Cell And Cluster Method
International Nuclear Information System (INIS)
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
Dose Distribution Calculation Using MCNPX Code in the Gamma-ray Irradiation Cell
International Nuclear Information System (INIS)
60Co-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 60Co 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
BCG: a computer code for calculating neutron spectra and criticality in cells of fast reactors
International Nuclear Information System (INIS)
The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is discussed. The code solves the unidimensional neutron transport equation together with interface current relations at each energy point in an unionized energy grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstruced total microscopic cross sections of the atomic species in the cell. Results for a simplified fuel cell illustrate the high resolution and accuracy that can be obtained with the code. (author)
BCG: a code for calculating pointwise neutron spectra and criticality in fast reactor cells
International Nuclear Information System (INIS)
The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is presented. The code solves the unidimensional neutron transport equation together with interface current relations at each energy in an unionized grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstructed total microscopic cross sections of the atomic species in the cell. Results for a defined sample problem illustrate the high resolution and accuracy that can be obtained with the code. (author)
Calculation of anisotropic few-group constants in asymptotic cells: the code ANICELL
International Nuclear Information System (INIS)
The theoretical background of the ANICELL computer program together with a user's manual is presented. ANICELL is a nuclear reactor neutron transport code which solves the traditional asymptotic and the so-called tilted flux transport problems in one-dimensional cylindrical geometry using linearly anisotropic scattering. The method of solution used is the first flight collision probability technique. Few-group constants including radial and axial diffusion coefficients for the cell are also prepared by the program. (author)
PINSPEC. A Monte Carlo code for pin cell spectral calculations for educational applications
International Nuclear Information System (INIS)
Students in many reactor physics courses are exposed to canonical reactor physics concepts through theoretical problems simplified to allow for tractable analytical solutions. Such problems typically require tedious mathematical derivation which is often not the most effective approach to teaching basic reactor physics concepts. A new complementary methodology to introduce these concepts is made possible with PINSPEC, a pin cell Monte Carlo code for educational use. PINSPEC enables students to simulate pin cell models for various reactor types with a simple-to-use Python interface. PINSPEC uses point-wise cross section data and includes a module for Single-Level Breit-Wigner cross-section generation and Doppler broadening. The PINSPEC code supports a variety of tallies which students may use to compute resonance integrals, multi-group cross sections, and more for various materials and pin configurations. PINSPEC is undergoing review for open source release in the near future such that it will be a free and accessible tool for instructors developing reactor physics curricula with an applied and interactive approach to learning. (author)
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.
Particle optics in the TIT-RFQ calculated using a 3D particle-in-cell code
International Nuclear Information System (INIS)
Beam dynamics in an RFQ at the Tokyo Institute of Technology was analyzed using a 3D particle-in-cell computer code. In this calculation not only space charge force between each macroparticles but also 3D image charge field were included. Beam transmission performance was calculated for two types of vane-tip design with different tip curvature radii. These results are compared with ones obtained with the idealized linear two-term potential. The old vane tip design with a small tip curvature radius has given very poor beam transmission efficiency which cannot be accepted for the actual machine. (author)
Calculation code of the fission products activity
International Nuclear Information System (INIS)
The document describes the two codes for the calculation of the fission products activity. The ''Pepin le bref'' code gives the exact value of the beta and gamma activities of completely known fission products. The code ''Plus Pepin'' introduces the beta and gamma activities whose properties are partially known. (A.L.B.)
Two-dimensional sensitivity calculation code: SENSETWO
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and parameter survey using developed analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. Evaporation calculations using cathode processor calculation code with distillation process, which was developed in 2000, were evaluated. By selecting proper input data (time step, mesh size etc.), the results showed that the present code agreed well for the evaporation rate of cadmium., and the capability of the distillation process design and simulation with the code has been confirmed. Parameter surveys using developed analytical model were performed for the purpose of reflection of cooling system design of the pyrochemical process cell. 4 cases of cooling flow patterns were surveyed at the normal and low flow rate conditions. From the result of parameter surveys, it was shown that the cooling pattern with direct cooling for heating facilities in the lower cell and balk cooling for upper cell is desirable. (author)
Spectroscopic calculation code ASPECT and its application
International Nuclear Information System (INIS)
The Code ASPECT is available for calculations of electronic levels of atoms and ions by the intermediate coupling scheme. This scheme is characterized by the simultaneous diagonalization of Hamiltonians for electronic repulsion, spin orbit interaction and crystal field effect. ASPECT performs the sorting of microstates involved in the electronic configuration in problem, calculation of matrix elements of these Hamiltonians, and diagonalization of the summed matrix. As input data, the calculation needs only parameter values of Slater integrals. ASPECT is also applied to calculate transition probabilities between the electronic levels obtained by this code. ASPECT is particularly focused on complex configurations containing f-electrons as met in Lanthanides and Actinides, which are not easily treated by an algebraic method. For convenience of users, Slater integral values for configurations fn of Lanthanides and Actinides are installed in the code so that users may select merely the atomic number. This document is composed of three parts. The first part (Chapter 1-3) describes quantum mechanical principles to calculate matrix elements of each unperturbed Hamiltonian and transition probabilities. The second part (Chapter 4) explains the structure of the code, and the last part (Chapter 5) serves as the manual for applications of this code, in which some samples are included. The third part (Chapter 6) is added as supplement for users who will improve this code. (author)
ASME Code Calculations for the CC Cryostat
Energy Technology Data Exchange (ETDEWEB)
Luther, R.D.; /Fermilab
1987-11-04
This engineering note contains the ASHE Code calculations for the CC Cryostat prepared by the manufacturer, Richmond-Lox Equipment Company. Most of these were taken from calculations initially prepared by Fermilab personne1and pub1ished in Eng. Note 68.
International Nuclear Information System (INIS)
The designers of the innovative reactors have proposed a number of approaches to increasing resource efficiency. Adding thorium, a fertile material, to the fuel is considered in this report. Under this approach, a large portion of the reactor output is produced by fissioning of the 233U resulting from neutron capture by thorium, which results in reduced requirements for naturally-occurring fissile uranium (235U). The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The concept of using Th-233U as fuel has been applied to an existing LWR design as compare with another fuel cycles (UO2 and MOX). SRAC code is extensive used to investigate the lattice cell problem. (author)
The code system COROUT: Radioactive inventory calculations
International Nuclear Information System (INIS)
The code system COROUT is devoted to the evaluation of nuclear reactor out-of-core radioactive inventory for the sake of the nuclear power plant decommissioning problem. The code includes calculations of the neutron flux distributions and activation kinetics in the consistent way. Only thermal neutrons are taken into consideration in the present code version. Code is divided into three steps. The first step prepares the necessary data file containing data on reactor geometry, core flux, reactor operational history and data on elements in the out-of-core zones. The main part of calculations are performed during the second step. Here the thermal neutron flux distribution in the out-of-core area is calculated for two-dimensional cylindrical geometry and the system of gain-loss equations and the activation kinetics is solved for the elements in the different out-of-core shells. The Vladimirov's method of iterations on the spatial grid is used for the neutron flux calculations. The kinetic equations are solved by the operational method. The change of neutron field due to activation during reactor campaign is taken into account. The third part of COROUT code system allows to prepare plots of flux and activity distribution for different shells. All steps could be initiated independently using the results stored at the previous steps. The code is destined for the personal computers and has been written on the base of 32-bit FORTRAN language for IBM PC. 4 refs, 6 figs, 1 tab
TEA: A Code Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver
2016-07-01
We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. and Eriksson. It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature–pressure pairs. We tested the code against the method of Burrows & Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows & Sharp. Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is available under a reproducible-research, open-source license via https://github.com/dzesmin/TEA.
Presentation of the NABE calculation code
International Nuclear Information System (INIS)
The purpose of the NABE code is to follow up the physical and chemical parameters of a concrete cell when large amounts of sodium are released by accident in the reactor building. The code contains several modules: behaviour of the concrete, heating-up, water and CO2 release, reactions between liquid and vapor sodium with the concrete breakdown products, combustion of sodium and hydrogen (water), power released by the fission products in the liquid, the gas and the concrete, behaviour of the atmosphere of the cell (pressure, temperature, leaks), sodium vaporization and boiling, and condensation on the cold walls
Development of Fast running DNBR Calculation Code
Energy Technology Data Exchange (ETDEWEB)
Kwon, Hyuk; Seo, K. W.; Kim, S. J.; Hwang, D. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2010-10-15
SMART core adopted a core protection(SCOPS) and a supervising system(SCOMS) to satisfy the SAFDL for AOO and normal operation. Generally, the criteria is limited to the DNBR limit so that the DNBR calculation module is required in the protection and the supervising system of core. There are CPU time limit and calculation robustness as some requirements of the DNBR calculation module in SCOPS and SCOMS caused by hardware limitations. The non-iterative few channel methods are needed to satisfy the requirements. Non-iterative numerical method is similar to the CETOP algorithm originated from ref. 1. The method is known as the non-iterative prediction and correction method. An optimum number of channels for core lumping model is selected as 4- channel which is same channel number of CETOP model. A compensation model of lumped channel is needed to ensure that the 4-channel thermal hydraulic field is nearly equivalent to that field of 1/8-core model that is calculated by MATRA-S. The code called FAST that is fast running DNBR calculation is developed to satisfy the requirements of CPU time and calculation robustness. Present paper is described of characteristics and calculation results of developed FAST code
Burnup calculation code system COMRAD96
International Nuclear Information System (INIS)
COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)
Core-wide calculations by HELIOS code
International Nuclear Information System (INIS)
The transport method of HELIOS is called the CCCP method, because it is based on current coupling and collision probabilities. The system to be calculated consists of space elements that are coupled with each other and with the boundaries by interface currents. The angular dependence of the interface or coupling currents can be discretized in various ways. This is done by partitioning the directional half-sphere into a number of θ polar levels, and each θ level into a number of φ azimuthal intervals. In two dimensional calculations the discretization of the azimuthal level is dominant. In the last issue of the HELIOS code (version 1.10) the maximum value of azimuthal discretization is increased from 4 to 12. This gives the possibility to calculate large (core-wide or near core-wide) systems with appropriate accuracy, which extends the applicability of the HELIOS program. This paper presents the experience gained from HELIOS calculations of large systems having several WWER-440 assemblies. The examined parameter is the core-wide power distribution, which was inadequately calculated by former versions of HELIOS. The application of high azimuthal discretization gives substantial improvement in accuracy, compared to reference solutions calculated by MCNP Monte-Carlo code. Although HELIOS is designed to calculate assembly-wide systems, it is now applicable to core-wide systems. Using its possibilities, the area of application is extended to calculate reference solutions for core-wide programs or to examine spectral changes of few-group cross sections due to burnup in real situations. Some potential areas of application are presented in the paper, together with the limitations of those applications. (Author)
KENO-IV code benchmark calculation, (6)
International Nuclear Information System (INIS)
A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multigroup constants library MGCL. The present report describes the results of a benchmark test using criticality experiments about Plutonium fuel in various shape. In all, 33 cases of experiments have been calculated for Pu(NO3)4 aqueous solution, Pu metal or PuO2-polystyrene compact in various shape (sphere, cylinder, rectangular parallelepiped). The effective multiplication factors calculated for the 33 cases distribute widely between 0.955 and 1.045 due to wide range of system variables. (author)
RTP: Radionuclides inventories calculation using ORIGEN Code
International Nuclear Information System (INIS)
ORIGEN is a widely used computer code for calculating the buildup, decay, and processing of radioactive materials. The ORIGEN code was created by famous and reputable nuclear institution in United States, Oak Ridge National Laboratory (ORNL). For a nuclear reactor, either it is a nuclear power reactor or nuclear research reactor, the radionuclide inventories data is important. This data is acquired by performing source term modelling. A fresh nuclear fuel could not cause any harm to human. However, used nuclear fuel could pose danger threat to human. The fission products particularly long-lived radionuclides for example H-3, Co-60, Cs-137 that are generated inside the fuel yield a significance amount of radioactivity. Therefore, there is no doubt that for a facility having a nuclear reactor, it is vital to anticipate the amount of fission products inside the fuel together with the radioactivity that it may emit. Sufficient information on the radionuclide inventories allows the facility to provide adequate shielding protection and ensure safe transportation of nuclear fuel, when it is needed. This paper briefly describes application of ORIGEN code to calculate the radionuclides inventories of TRIGA-PUSPATI REACTOR (RTP) fuel. (author)
Calculation of doppler coefficient of reactivity by WIMS code
International Nuclear Information System (INIS)
The Doppler coefficient of reactivity is an important factor in prediction of several transients in light water reactors. Some of the past studies raised the question about the 10% uncertainty that traditionally was taken in calculations of Doppler coefficient by LWR lattice code. In order to bridge the gap of lack of accurate benchmark problem to evaluate the accuracy of Doppler effect, Mosteller et al. proposed a computational benchmark problem of Doppler coefficient to evaluate the accuracy and consistency of LWR lattice physics code. In this paper we present the results obtained from WIMS-D4 lattice code and compare it with those obtained by CELL-2 lattice code part of the EPRI-PRESS reactor physics package. The results obtained from the Monte Carlo code MCNP-3A served as reference for both cases, and was taken from ref 1. (authors). 4 refs., 2 figs., 1 tab
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)
Data calculation program for RELAP 5 code
International Nuclear Information System (INIS)
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)
KENO-IV code benchmark calculation, (4)
International Nuclear Information System (INIS)
A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multi-group constants library MGCL. The present paper describes the results of a test using criticality experiments about slab-cylinder system of uranium nitrate solution. In all, 128 cases of experiments have been calculated for the slab-cylinder configuration with and without plexiglass reflector, having the various critical parameters such as the number of cylinders and height of the uranium nitrate solution. It is shown among several important results that the code and library gives a fairly good multiplication factor, that is, k sub(eff) -- 1.0 for heavily reflected cases, whereas k sub(eff) -- 0.91 for the unreflected ones. This suggests the necessity of more advanced treatment of the criticality calculation for the system where neutrons can easily leak out during slowing down process. (author)
Methods and computer codes for nuclear systems calculations
Indian Academy of Sciences (India)
B P Kochurov; A P Knyazev; A Yu Kwaretzkheli
2007-02-01
Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space–time calculations. Computer code TRIFON solves space-energy problem in (, ) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space–time neutron processes simulation. Modification of TRIFON was developed for the simulation of space–time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.
Calculation code revised MIXSET for Purex process
International Nuclear Information System (INIS)
Revised MIXSET is a FORTRAN IV calculation code developed to simulate steady and transient behaviors of the Purex extraction process and calculate the optimum operating condition of the process. Revised MIXSET includes all the functions of MIXSET code as shown below. a) Maximum chemical system of eight components can be handled with or without mutual dependence of the distribution of components. b) The flowrate and concentration of feed can be renewed successively at any state, transient or steady, for searching optimum operating conditions. c) Optimum inputs of feed concentrations and flowrates can be calculated to satisfy both of specification and recovery rate of a product. d) Radioactive decay reactions can be handled on each component. Besides these functions, the following chemical reactions concerned in Purex process are newly-included in Revised MIXSET code and the quantitative changes of components such as H+, U(IV), U(VI), Pu(III), Pu(IV), NH2OH, N2H4 can be simulated. 1st Gr. (i) reduction of Pu(IV); U4+ + 2Pu4+ + 2H2O → UO22+ + 2Pu3+ + 4H+. (ii) oxidation of Pu(III); 2Pu3+ + 3H+ + NO3- → 2Pu4+ + HNO2 + H2O. (iii) oxidation of U(IV); U4+ + NO3- + H2O → UO22+ + H+ + HNO2 2U4+ + O2 + 2H2O → 2UO22+ + 4H+. (iv) decomposition of HNO2; HNO2 + N2H5+ → HN3 + 2H2O + H+. (author)
Electrical Conductivity Calculations from the Purgatorio Code
Energy Technology Data Exchange (ETDEWEB)
Hansen, S B; Isaacs, W A; Sterne, P A; Wilson, B G; Sonnad, V; Young, D A
2006-01-09
The Purgatorio code [Wilson et al., JQSRT 99, 658-679 (2006)] is a new implementation of the Inferno model describing a spherically symmetric average atom embedded in a uniform plasma. Bound and continuum electrons are treated using a fully relativistic quantum mechanical description, giving the electron-thermal contribution to the equation of state (EOS). The free-electron density of states can also be used to calculate scattering cross sections for electron transport. Using the extended Ziman formulation, electrical conductivities are then obtained by convolving these transport cross sections with externally-imposed ion-ion structure factors.
Integrated burnup calculation code system SWAT
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
The aim of this report is the comparison between two cell code systems which base on two different physical approaches. To this end wet and dry PROTEUS buffer cell reactivities are computed using the transport codes ONEDANT and SURCU together with self-shielded cross section ENDF/B-IV based libraries obtained using NMATXS module and postprocessor code TRANSX-CTR. TRANSX-CTR performs self-shielding of resonance cross sections using Bondarenko method. The same calculations are then repeated using new NJOY module MICROR and two region spectrum code MICROX-2, to perform more accurate pointwise resonance shielding solving slowing down equations. It is shown that the dry cell is not sensitive to the resonance cross sections, because the spectrum is too hard and energy of most neutrons is higher than resonance energies of fissionable isotopes. However the lattice is sensitive to the fission spectrum used. Also the computations pertaining to the water moderated cell show sensitivity to different weighting functions used to calculate fission spectra from fission matrices, because the spectrum is still too hard. But now large differencies in the reactivity of the cell arise between the shielding factor method and the computations based on the pointwise self-shielding. This is because the shielding factor method even if sophisticated (Bondarenko model) is not representative in lower resonance range where resonances are not narrow. (author)
Calculation code MIXSET for Purex process
International Nuclear Information System (INIS)
MIXSET is a FORTRAN IV calculation code for Purex process that simulate the dynamic behavior of solvent extraction processes in mixer-settlers. Two options permit terminating dynamic phase by time or by achieving steady state. These options also permit continuing calculation successively using new inputs from a arbitrary phase. A third option permits artificial rapid close to steady state and a fourth option permits searching optimum input to satisfy both of specification and recovery rate of product. MIXSET handles maximum chemical system of eight components with or without mutual dependence of the distribution of the components. The chemical system in MIXSET includes chemical reactions and/or decaying reaction. Distribution data can be supplied by third-power polynominal equations or tables, and kinetic data by tables or given constants. The fluctuation of the interfacial level height in settler is converted into the flow rate changes of organic and aqueous stream to follow dynamic behavior of extraction process in detail. MIXSET can be applied to flowsheet study, start up and/or shut down procedure study and real time process management in countercurrent solvent extraction processes. (auth.)
Analytical stress tensor and pressure calculations with the CRYSTAL code
Doll, Klaus
2010-01-01
Abstract The calculation of the stress tensor and related properties and its implementation in the CRYSTAL code are described. The stress tensor is obtained from the earlier implemented analytical gradients with respect to the cell parameters. Subsequently, the pressure and enthalpy is computed, and a test concerning the pressure driven phase transition in KI is used as an illustration. Finally, the possibility of applying external pressure is implemented. The ...
Code system BCG for gamma-ray skyshine calculation
International Nuclear Information System (INIS)
A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)
Fast reactor nuclear physics parameters calculation code system 'EXPARAM'
International Nuclear Information System (INIS)
The calculation code system ''EXPARAM'' was designed to analyze the experimental results systematically measured at the fast critical assembly (FCA) in Tokai research establishment of JAERI. Some calculation codes developed independently in JAERI and in US research institutes were collected and arranged as the fast reactor physics calculation code system. The multi-group core calculation code and the perturbation calculation code based on the diffusion theory and transport theory calculate the physics parameters such as eigenvalue, reaction rate, Doppler reactivity worth and sodium void worth. The dynamic physics parameters such as prompt neutron lifetime and effective delayed neutron fraction are also calculated. Input and Output data of calculation codes are transferred to each other using a direct access file on UNIX computer system. (author)
Development of transient neutron transport calculation code
International Nuclear Information System (INIS)
A transient neutron transport code for time-dependent analyses of neutronics systems, named DOT4-T, has been developed. The code is based on the Discrete Ordinates code DOT4.2, which solves the steady-state neutron transport equation in two dimensions. For the discretization of time variable, a direct method, the fully implicit and unconditionally stable time integration scheme, has been employed. The resulting code has been tested using several one-dimensional and two-dimensional benchmark problems, and the results obtained with DOT4-T shows very satisfactory agreement with the benchmark problem results. (authors)
DeCART code verifications by numerical benchmark calculations of HTTR
International Nuclear Information System (INIS)
DeCART code verifications have been performed through the numerical benchmark calculations of HTTR. The reference calculations have been carried out using the Monte Carlo McCARD code in which a double heterogeneity model was used. Verification results show that the DeCART code gives less negative MTC and RTC than the McCARD code does and thus the DeCART code underestimates the multiplication factors at states with high moderator and reflector temperatures. However, the DeCART code predicts more negative FTC than McCARD code does. In the depletion calculation for the HTTR single cell and single block, the error of the DeCART code increases with burnup. While the DeCART code error in a 2-dimensional core depletion calculation decreases with burnup up to around 500 FPD. (author)
A code to calculate multigroup constants for fast neutron reactor
International Nuclear Information System (INIS)
KQCS-2 code is a new improved version of KQCS code, which was designed to calculate multigroup constants for fast neutron reactor. The changes and improvements on KQCS are described in this paper. (author)
The VADMAP code to calculate the SAF of photon
International Nuclear Information System (INIS)
A computer code VADMAP has been developed to calculate the Specific Absorbed Fraction, SAF, of photon. The development of the code is aimed at efficient and systematic preparation of the SAF data files for several different human phantoms in a suitable form as a direct input data file to DOSimetric DAta Calculation system, DOSDAC, which is being developed at Japan Atomic Energy Research Institute, JAERI. This document describes the methodology used in the code, the code structure, user's information including the way of implementing the code on FACOM/M-380, and the performance through calculation and preparation of the SAF data file. In order to show the performance of the code, a set of the SAF values for an adult human phantom was calculated and was organized to prepare the SAF file. Comparing the calculated SAF values with those tabulated in ORNL-5000, the quality of the code was examined. (author)
MOx benchmark calculations by deterministic and Monte Carlo codes
International Nuclear Information System (INIS)
Highlights: ► MOx based depletion calculation. ► Methodology to create continuous energy pseudo cross section for lump of minor fission products. ► Mass inventory comparison between deterministic and Monte Carlo codes. ► Higher deviation was found for several isotopes. - Abstract: A depletion calculation benchmark devoted to MOx fuel is an ongoing objective of the OECD/NEA WPRS following the study of depletion calculation concerning UOx fuels. The objective of the proposed benchmark is to compare existing depletion calculations obtained with various codes and data libraries applied to fuel and back-end cycle configurations. In the present work the deterministic code NEWT/ORIGEN-S of the SCALE6 codes package and the Monte Carlo based code MONTEBURNS2.0 were used to calculate the masses of inventory isotopes. The methodology to apply the MONTEBURNS2.0 to this benchmark is also presented. Then the results from both code were compared.
Rotamak equilibrium calculations using the PEST code
International Nuclear Information System (INIS)
This report describes the use of the equilibrium part of the Princeton equilibrium and stability code PEST to model rotamak equilibria with an applied toroidal magnetic field. An overview of the code is provided, together with a list of required input data. The simulation of a range of equilibria measured in the ANSTO rotamak shows that the rotamak approximately satisfies magnetohydrodynamic equilibrium. Of particular interest is the presence of large diamagnetic poloidal current about the magnetic axis which produces a peak in the plasma pressure on the magnetic axis. For a low toroidal field, however, poloidal current of opposite direction is simultaneously driven on flux surfaces distant from the magnetic axis, producing paramagnetism
CRACKEL: a computer code for CFR fuel management calculations
International Nuclear Information System (INIS)
The CRACKLE computer code is designed to perform rapid fuel management surveys of CFR systems. The code calculates overall features such as reactivity, power distributions and breeding gain, and also calculates for each sub-assembly plutonium content and power output. A number of alternative options are built into the code, in order to permit different fuel management strategies to be calculated, and to perform more detailed calculations when necessary. A brief description is given of the methods of calculation, and the input facilities of CRACKLE, with examples. (author)
CONDOR: neutronic code for fuel elements calculation with rods
International Nuclear Information System (INIS)
CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)
Xenon poisoning calculation code for miniature neutron source reactor (MNSR)
International Nuclear Information System (INIS)
In line with the actual requirements and based upon the specific characteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poisoning Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data
Xenon poisoning calculation code for miniature neutron source reactor (MNSR)
Institute of Scientific and Technical Information of China (English)
无
2001-01-01
In line with the actual requirements and based upon the specific char acteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poison ing Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data.
CONSUL code package application for LMFR core calculations
International Nuclear Information System (INIS)
CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)
TEA: A Code for Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver
2015-01-01
We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. (1958) and Eriksson (1971). It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature-pressure pairs. We tested the code against the method o...
Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
Energy Technology Data Exchange (ETDEWEB)
Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)
2015-09-15
A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.
Code package of the physics calculation and fuel management of uranium hydride zirconium reactor
International Nuclear Information System (INIS)
The code package of uranium hydride zirconium reactor physics calculation is established. considering the thermalization of H in ZrH, the nuclear data of H in ZrH in WIMS library pattern are provided and WIMS-N2 library is obtained. The cell parameters are calculated using WIMS-D/4 code and SIMS-N2 library. The diffusion calculation is performed using CITATION code and SIXTUS-2 code. The in-core fuel management code XPR-ICFM is obtained on the basis of SIXTUS-2 code. To check the accuracy and reliability of the code package, the critical keff and the value of the control rod of abroad TRIGA, the pulsed reactor in china are calculated. The results are satisfied
Development and validation of a nodal code for core calculation
International Nuclear Information System (INIS)
The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency
Recent transportation calculation code systems and their accuracy evaluation
International Nuclear Information System (INIS)
In the field of shielding design, many studies have been carried out for the development of radiation transportation calculation codes (transportation codes) including Monte Carlo codes. The present report outlines major transportation codes used in Japan for design of shielding. Major one-dimensional codes include ANISN (Sn), PALLAS-PL and SP-Br (direct integration) whili two-dimensional ones include DOT-3.5 and TWOTRAN-II. All these transportation codes have been developed on the basis of numerical solution to the Boltzmann's transportation equation. These codes are roughly divided into two groups: discrete ordinates type and Monte Carlo type. The former include Sn-type codes and direct integration type codes. Sn-type codes are currently used most widely. The accuracy and other features of a code should be tested before applysing it to practical shielding design. One of the techniques for this purpose is the benchmark method, which consists of benchmark tests and analysis of the test results. The possible overall error involved in calculations can be determined from the benchmark tests. (Nogami, K.)
SRAC2006: A comprehensive neutronics calculation code system
International Nuclear Information System (INIS)
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, SN transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)
Calculations of angular momentum coupling coefficients on a computer code
International Nuclear Information System (INIS)
In this study, Clebsch-Gordan coefficients, 3j symbols, Racah coefficients, Wigner's 6j and 9j symbols were calculated by a prepared computer code of COEFF. The computer program COEFF is described which calculates angular momentum coupling coefficients and expresses them as quotient of two integers multiplied by the square root of the quotient of two integers. The program includes subroutines to encode an integer into its prime factors, to decode of prime factors back into an integer , to perform basic arithmetic operations on prime-coded numbers, as well as subroutines which calculate the coupling coefficients themselves. The computer code COEFF had been prepared to run on a VAX. In this study we rearranged the code to run on PC and tested it successfully. The obtained values in this study, were compared with the values of other computer programmes. A pretty good agreement is obtained between our prepared computer code and other computer programmes
TEA: A Code for Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Bowman, M Oliver
2015-01-01
We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. (1958) and Eriksson (1971). It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature-pressure pairs. We tested the code against the method of Burrows & Sharp (1999), the free thermochemical equilibrium code CEA (Chemical Equilibrium with Applications), and the example given by White et al. (1958). Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is ...
MCOR - Monte Carlo depletion code for reference LWR calculations
International Nuclear Information System (INIS)
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
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
Calculation codes in radiation protection, radiation physics and dosimetry
International Nuclear Information System (INIS)
These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)
Revisiting the calculation of effective free distance of turbo codes
Chatzigeorgiou, I; Wassell, I. J.
2008-01-01
The expression for the minimum Hamming weight of the output of a constituent convolutional encoder, when its input is a weight-2 sequence is revisited. The new expression particularly facilitates the calculation of the effective free distance of recently proposed schemes, namely non-systematic turbo codes and pseudo-randomly punctured turbo codes.
Burnup calculation methodology in the serpent 2 Monte Carlo code
International Nuclear Information System (INIS)
This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)
A computer code for calculating a γ-external dose from a randomly distributed radioactive cloud
International Nuclear Information System (INIS)
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)
Description of the CAREM Reactor Neutronic Calculation Codes
International Nuclear Information System (INIS)
In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included
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.)
Exposure calculation code module for reactor core analysis: BURNER
International Nuclear Information System (INIS)
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
Development and validation of Monte-Carlo burnup calculation code MCNTRANS
International Nuclear Information System (INIS)
A new nuclear fuel burnup calculation code MCNTRANS based on MCNP was introduced in this paper. The neutronics calculation parameter was extracted from the MCNP5 reaction rate tally result, while a graph theory algorithm was implemented to track the burnup chain and the analytic solution of the Bateman equation was given. At the same time, the detailed physical process was considered to improve the accuracy and serviceability of this code, and prediction-correction method was used to allow a large burnup step. The OECD/NEA and JAERI pin cell benchmark problems were used to validate the code MCNTRANS while a reference result was given by other code. It can be concluded that the calculation results of MCNTRANS are generally consistent with the experimental result and that of the other burnup codes, and part of the actinides and fission products calculation result show better accuracy. (authors)
Calculation of SPERT Reactor benchmarks using 3D diffusion code DIREN
International Nuclear Information System (INIS)
The three dimensional diffusion code DIREN was developed at Institute for Nuclear Research (INR) Pitesti for reactor physics calculations for natural uranium and advanced CANDU reactors. Cell codes used are WIMS (from NEA library) and DRAGON (available in open source system). The latter is used also for super cell modeling of reactor control devices. These codes and the auxiliary programs were linked together in a calculation system. In order to apply WIMS-DRAGON-DIREN system to LWR, first the reactor SPERT benchmarks problems were calculated. The core including the control rods was modeled in three dimensional geometry. Following the calculations of the critical height (Hcrit), three dimensional power and flux distributions were obtained. The standard procedure used for CANDU reactor calculations (incremental cross sections for reactivity devices) underestimated the worth of control rods. A simple procedure to obtain the internal boundary conditions was developed using the super cell code DRAGON. Also the DIREN 3D diffusion code was modified to apply inner boundary conditions at control rods assigned volumes. Applying the inner boundary conditions yielded results closer to the measured values (e.g. the measured Hcrit was 49.53 cm as compared to 53.15 cm, the calculated one on 7 groups for nominal temperature). The reactivity coefficients for temperature and density required in transient's simulations were also calculated. The sample test problem T83 (hot stand-by, fast transient) was simulated using the RELAP code. (authors)
Validation of IRBURN calculation code system through burnup benchmark analysis
International Nuclear Information System (INIS)
Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.
CONSUL code package application for LMFR core calculations
Energy Technology Data Exchange (ETDEWEB)
Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)
2008-07-01
CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)
Verification test calculations for the Source Term Code Package
International Nuclear Information System (INIS)
The purpose of this report is to demonstrate the reasonableness of the Source Term Code Package (STCP) results. Hand calculations have been performed spanning a wide variety of phenomena within the context of a single accident sequence, a loss of all ac power with late containment failure, in the Peach Bottom (BWR) plant, and compared with STCP results. The report identifies some of the limitations of the hand calculation effort. The processes involved in a core meltdown accident are complex and coupled. Hand calculations by their nature must deal with gross simplifications of these processes. Their greatest strength is as an indicator that a computer code contains an error, for example that it doesn't satisfy basic conservation laws, rather than in showing the analysis accurately represents reality. Hand calculations are an important element of verification but they do not satisfy the need for code validation. The code validation program for the STCP is a separate effort. In general the hand calculation results show that models used in the STCP codes (e.g., MARCH, TRAP-MELT, VANESA) obey basic conservation laws and produce reasonable results. The degree of agreement and significance of the comparisons differ among the models evaluated. 20 figs., 26 tabs
Progress on burnup calculation methods coupling Monte Carlo and depletion codes
Energy Technology Data Exchange (ETDEWEB)
Leszczynski, Francisco [Comision Nacional de Energia Atomica, San Carlos de Bariloche, RN (Argentina). Centro Atomico Bariloche]. E-mail: lesinki@cab.cnea.gob.ar
2005-07-01
Several methods of burnup calculations coupling Monte Carlo and depletion codes that were investigated and applied for the author last years are described. here. Some benchmark results and future possibilities are analyzed also. The methods are: depletion calculations at cell level with WIMS or other cell codes, and use of the resulting concentrations of fission products, poisons and actinides on Monte Carlo calculation for fixed burnup distributions obtained from diffusion codes; same as the first but using a method o coupling Monte Carlo (MCNP) and a depletion code (ORIGEN) at a cell level for obtaining the concentrations of nuclides, to be used on full reactor calculation with Monte Carlo code; and full calculation of the system with Monte Carlo and depletion codes, on several steps. All these methods were used for different problems for research reactors and some comparisons with experimental results of regular lattices were performed. On this work, a resume of all these works is presented and discussion of advantages and problems found are included. Also, a brief description of the methods adopted and MCQ system for coupling MCNP and ORIGEN codes is included. (author)
Coupled code calculation of rod withdrawal at power accident
Energy Technology Data Exchange (ETDEWEB)
Grgić, Davor, E-mail: davor.grgic@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Benčik, Vesna, E-mail: vesna.bencik@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Šadek, Siniša, E-mail: sinisa.sadek@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia)
2013-08-15
Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced.
Coupled code calculation of rod withdrawal at power accident
International Nuclear Information System (INIS)
Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced
Hot zero power reactor calculations using the Insilico code
Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.
2016-06-01
In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.
User effects on the transient system code calculations. Final report
International Nuclear Information System (INIS)
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)
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)
The best estimate codes applied to VVER calculations validation methodology
International Nuclear Information System (INIS)
The best estimate thermal hydraulic codes for PWRs and BWRs accident analysis were elaborated in 80-th years. The main goal of the best estimate codes (BEC) calculations was to obtain the real picture of the reactor facilities parameters changes during transient and accident regimes. This codes were based on 5--6 equations mathematical models taking into account separate flows of water and vapor. Validation of the best estimate thermal hydraulic codes applied VVER calculations is the main objective of the article. The concept of CSNI test matrices application to the codes validation is presented. The ways of the matrices improvement for this purpose is outlined. Taking into account deficiency of the operating integral test facilities a special attention is focused on the original separate effects experimental data obtained in Ukraine and which was not included to the CSNI matrices. To facilitate application of the obtained data to the codes validation the numerical experiment method was elaborated. The example of the method application is described
Core design calculations of IRIS reactor using modified CORD-2 code package
International Nuclear Information System (INIS)
Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)
Verification of the COCAGNE core code using cluster depletion calculations
International Nuclear Information System (INIS)
EDF/R and D is developing a new calculation scheme based on the transport- Simplified Pn (SPn) approach. The lattice code used is the deterministic code APOLLO2, developed at CEA with the support of EDF and AREVA-NP. The core code is the code COCAGNE, developed at EDF R and D. The latter can take advantage of a microscopic depletion solver which improves the treatment of spectral history effects. This solver can resort to a specific correction based on the use of the Pu239 concentration as a spectral indicator. In order to evaluate the improvements brought by this Pu239 correction model, one uses (3x3 assemblies) cluster depletion calculations as test-cases. UOX and UOX/MOX clusters are both considered. As a reference, APOLLO2 depletion calculations of these clusters, using a critical boron (CB) search scheme at each calculation step, are performed. This choice of methodology (using CB search instead of a fixed average CB) enables to highlight historical spectral effects related to the boron concentration. This methodology is also more consistent with the depletion calculation of real cores. Pin by pin COCAGNE calculations are performed and compared with the APOLLO2 results. The analysis of the results obtained shows that the boron concentration computed by COCAGNE gets more consistent with APOLLO2 when the Pu239 corrector is used, especially for UOX/MOX clusters. As for pin power distribution, the use of the Pu239 model also enables to reduce slightly the gap between APOLLO2 and COCAGNE. This work will be extended to clusters with gadolinium-poisoned fuel assemblies and reflector regions. (author)
WIPP Benchmark calculations with the large strain SPECTROM codes
International Nuclear Information System (INIS)
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
Decay heat calculation an international nuclear code comparison
International Nuclear Information System (INIS)
The results of an international code comparison on decay heat are presented and discussed. Participants from more than ten laboratories calculated, using the same input data, decay heat for thirteen cooling times between 1 and 1013 sec. Two irradiation cases were proposed: fission pulse and 3x107 seconds of irradiation of 235U fuel. The results are analysed and compared. This inter-comparison shows that, if the same input data are given, most of the codes give very similar results for the decay heat and consequently also for the fission product contribution
Verification calculations for the WWER version of the TRANSURANUS code
International Nuclear Information System (INIS)
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
Introduction to reactor lattice calculations by the WIMSD code
International Nuclear Information System (INIS)
The present report is based on lectures delivered at the Workshop on Nuclear Reaction Data and Nuclear Reactors: Physics, Design and Safety hold in International Centre of Theoretical Physics, Trieste, in March 1998. The main goal of the set of lectures was to give the basis of reactor physics calculations for participants working on nuclear data.The last lectures, devoted to WIMS including the WIMSD code users. Following this general line the material is divided into three parts: The first part includes a short description of neutron transport phenomena limited to those definitions that are necessary to understand the approach to practical solution of the problem given in the second part on reactor lattice transport calculations. The detailed discussion of the neutron cross sections has been skipped as this subject has been treated in detail by other lectures. In the third part those versions of the well-known WIMSD code which are distributed by NEA Data Bank are described. The general structure of the code is given supplied in a more detailed description of aspects being the most common points of misunderstanding for the code users. (author)
SCRAM calculations with the KIKO3D code
International Nuclear Information System (INIS)
Discrepancies between calculated static reactivities and measured reactivities evaluated with reactivity meters led to investigating SCRAM with the KIKO3D dynamic code. The spatial effects near to the position of ionisation chambers were studied. As was expected, reactivities calculated from the flux curves of different nodes using inverse point kinetics are in a wide range. The dynamic and static reactivities in case of asymmetric SCRAM differ considerably as a result of the slow flux shape redistribution. The effect of source neutrons from spontaneous fission and the node-wise delayed neutron fraction on the results is also presented. (Authors)
Wake field calculations with three-dimensional BCI code
International Nuclear Information System (INIS)
The new MAFIA code T3 is introduced, belonging to a family of fully 3-D codes for computer-aided design of RF cavities developed by the MAFIA collaboration at DESY, KFA Juelich and LANL. T3 is a 3-D extension of TBCI, solving Maxwell's equations in the time domain using the FIT ansatz, allowing the use of structures of arbitrary shape and dielectric material insertions, and integrating the wake potential at arbitrary positions. An open boundary condition was implemented to simulate infinite beam pipes. An IBM 3081 with 5 Mbytes of main storage can handle problems up to 80,000 mesh points while a window option enables the treatment of very long structures using up to 1,000,000 mesh points. Together with its postprocessors W3COR and W3OUT and the MAFIA mesh generator, this code was used to calculate wake potentials for several beam pipe components (i.e. vacuum chambers, vacuum chamber junctions etc.) which required fully 3-D calculations. Comparison of T3 results with TBCI calculations in the case of a cylindrically symmetric structure (pillbox) showed the agreement within a few percent
Calculation code PULCO for Purex process in pulsed column
International Nuclear Information System (INIS)
The calculation code PULCO, which can simulate the Purex process using a pulsed column as an extractor, has been developed. The PULCO is based on the fundamental concept of mass transfer that the mass transfer within a pulsed column occurs through the interface of liquid drops and continuous phase fluid, and is the calculation code different from conventional ones, by which various phenomena such as the generation of liquid drops, their rising and falling, and the unification of liquid drops actually occurring in a pulsed column are exactly reflected and can be correctly simulated. In the PULCO, the actually measured values of the fundamental quantities representing the extraction behavior of liquid drops in a pulsed column are incorporated, such as the mass transfer coefficient of each component, the diameter and velocity of liquid drops in a pulsed column, the holdup of dispersed phase, and axial turbulent flow diffusion coefficient. The verification of the results calculated with the PULCO was carried out by installing a pulsed column of 50 mm inside diameter and 2 m length with 40 plate stage in a glove box for unirradiated uranium-plutonium mixed system. The results of the calculation and test were in good agreement, and the validity of the PULCO was confirmed. (Kako, I.)
Comparison of computer code calculations with FEBA test data
International Nuclear Information System (INIS)
The FEBA forced feed reflood experiments included base line tests with unblocked geometry. The experiments consisted of separate effect tests on a full-length 5x5 rod bundle. Experimental cladding temperatures and heat transfer coefficients of FEBA test No. 216 are compared with the analytical data postcalculated utilizing the SSYST-3 computer code. The comparison indicates a satisfactory matching of the peak cladding temperatures, quench times and heat transfer coefficients for nearly all axial positions. This agreement was made possible by the use of an artificially adjusted value of the empirical code input parameter in the heat transfer for the dispersed flow regime. A limited comparison of test data and calculations using the RELAP4/MOD6 transient analysis code are also included. In this case the input data for the water entrainment fraction and the liquid weighting factor in the heat transfer for the dispersed flow regime were adjusted to match the experimental data. On the other hand, no fitting of the input parameters was made for the COBRA-TF calculations which are included in the data comparison. (orig.)
Simulator validation of calculation code in REDNET upgrade system
International Nuclear Information System (INIS)
The reactor data network (REDNET) is a computer-based data acquisition, display and archival system which acquires data from the National Research Universal (NRU) reactor's 'fuelled sites', and several experimental loop facilities in support of CANDU technology development (e.g., fuel, fuel behaviour, and materials research programs). The system supports the processing of data collected for subsequent display at the respective experimental facilities, and in the NRU control room. REDNET was installed in the 1980s based on the 1970s computer technology. The computer hardware is obsolete and spare parts are either extremely hard to find or are now unavailable. The Upgrade system is intended to replace the REDNET and eliminate the risk of losing the data acquisition of important experimental data needed in support of the CANDU Fuel Development Program. An important goal of the Upgrade system is to improve the accuracy in the measurement and calculation of thermal power. Calculations in REDNET are performed in FORTRAN code with some in-house macros. The same calculations are re-implemented in the Upgrade system in structured-text and function-block languages. To ensure that there is no deviation or loss of accuracy in the calculations of the Upgrade system compared to those in REDNET, software validation is performed on calculation code in the Upgrade system. The validation consists of a two-stage and three-point check (at ∼0%, 50% and ∼100% signal level) process for every data type and data point in the Upgrade system. This paper presents the purpose, the major tools and process, and the results of the validation. It is concluded, based on the validation results, that the Upgrade system achieves at least the same, and in many cases better, accuracy in all the calculations. (author)
Calculation of reactor pressure vessel fluence using TORT code
International Nuclear Information System (INIS)
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 1.784x1018n/cm2. The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60 cm below the midplane at zero degree
Tokamak plasma power balance calculation code (TPC code) outline and operation manual
International Nuclear Information System (INIS)
This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (author)
ANIGAM: a computer code for the automatic calculation of nuclear group data
International Nuclear Information System (INIS)
The computer code ANIGAM consists mainly of the well-known programmes GAM-I and ANISN as well as of a subroutine which reads the THERMOS cross section library and prepares it for ANISN. ANIGAM has been written for the automatic calculation of microscopic and macroscopic cross sections of light water reactor fuel assemblies. In a single computer run both were calculated, the cross sections representative for fuel assemblies in reactor core calculations and the cross sections of each cell type of a fuel assembly. The calculated data were delivered to EXTERMINATOR and CITATION for following diffusion or burn up calculations by an auxiliary programme. This report contains a detailed description of the computer codes and methods used in ANIGAM, a description of the subroutines, of the OVERLAY structure and an input and output description. (oririg.)
Adjoint Monte Carlo techniques and codes for organ dose calculations
International Nuclear Information System (INIS)
Adjoint Monte Carlo simulations can be effectively used for the estimation of doses in small targets when the sources are extended in large volumes or surfaces. The main features of two computer codes for calculating doses at free points or in organs of an anthropomorphic phantom are described. In the first program (REBEL-3) natural gamma-emitting sources are contained in the walls of a dwelling room; in the second one (POKER-CAMP) the user can specify arbitrary gamma sources with different spatial distributions in the environment: in (or on the surface of) the ground and in the air. 3 figures
Saphyr: a code system from reactor design to reference calculations
International Nuclear Information System (INIS)
In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels
Saphyr: a code system from reactor design to reference calculations
Energy Technology Data Exchange (ETDEWEB)
Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)
2003-07-01
In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.
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
Burnup calculations using serpent code in accelerator driven thorium reactors
International Nuclear Information System (INIS)
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232Th and mixed 233U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
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.)
Computer code for shielding calculations of x-rays rooms
International Nuclear Information System (INIS)
The building an effective barrier against ionizing radiation present in radiographic rooms requires consideration of many variables. The methodology used for thickness specification of primary and secondary, barrier of a traditional radiographic room, considers the following factors: Use Factor, Occupational Factor, distance between the source and the wall, Workload, Kerma in the air and distance between the patient and the source. With these data it was possible to develop a computer code, which aims to identify and use variables in functions obtained through graphics regressions provided by NCRP-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) report, for shielding calculation of room walls, and the walls of the dark room and adjacent areas. With the implemented methodology, it was made a code validation by comparison of results with a study case provided by the report. The obtained values for thickness comprise different materials such as concrete, lead and glass. After validation it was made a case study of an arbitrary radiographic room.The development of the code resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in september/2011. (authors)
The PHREEQE Geochemical equilibrium code data base and calculations
International Nuclear Information System (INIS)
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)
Improvements of the subgroup resonance calculation code SUGAR
International Nuclear Information System (INIS)
Highlights: • Improvement of subgroup resonance code SUGAR is undertaken to increase efficiency. • Resonant nuclides are grouped for complex isotope compositions. • An in-house matrix MOC solver is employed to replace the AutoMOC solver. • These techniques speed 5–32 without any loss of accuracy or geometric flexibility. - Abstract: Due to its accuracy and geometric flexibility, the subgroup method is becoming a more and more attractive resonance calculation approach dedicated to obtaining resonance group macroscopic cross sections from multi-group libraries. In order to increase the efficiency of our subgroup code SUGAR, this paper contributes to the development of the code from four aspects. Firstly, subgroup parameters were proved to be problem-independent and the number of subgroups can be chosen automatically. This motivated us to produce a new multi-group library. Secondly, what subgroup method really needs is the relative subgroup flux within each multi-group instead of the relative multi-group flux between different groups. Thus, it is unnecessary to iteratively calculate in the whole energy range if the connections between different energy ranges can be approximated by a simple method. Thirdly, for problems with complex isotope compositions, resonant nuclides could be grouped according to their resonance characteristics. By this grouping, computational effort could be significantly reduced since nominal resonant nuclides turn out to be these nuclide groups rather than the actual nuclides. Finally, considering that most of the computational effort is spent on solving the subgroup neutron transport equation, an in-house matrix MOC solver is employed to replace the AutoMOC solver. In this way, the higher speed of the matrix MOC solver can be fully utilized by our subgroup code. To verify these theories and to prove the improvements, a series of benchmark problems were solved. It is demonstrated by these numerical results that these techniques can
International Nuclear Information System (INIS)
Three computer codes MATHEW, ADPIC of LLNL and GAMPUL of JAERI for prediction of wind field, concentration and external exposure rate of airborne radioactive materials are vectorized and the results are presented. Using the continuous equation of incompressible flow as a constraint, the MATHEW calculates the three dimensional wind field by a variational method. Using the particle-in -cell method, the ADPIC calculates the advection and diffusion of radioactive materials in three dimensional wind field and terrain, and gives the concentration of the materials in each cell of the domain. The GAMPUL calculates the external exposure rate assuming Gaussian plume type distribution of concentration. The vectorized code MATHEW attained 7.8 times speedup by a vector processor FACOM230-75 APU. The ADPIC and GAMPUL are estimated to attain 1.5 and 4 times speedup respectively on CRAY-1 type vector processor. (author)
Miniature neutron source reactor burnup calculations using IRBURN code system
International Nuclear Information System (INIS)
Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.
Electronic transport calculations in the ONETEP code: Implementation and applications
Bell, Robert A.; Dubois, Simon M.-M.; Payne, Michael C.; Mostofi, Arash A.
2015-08-01
We present an approach for computing Landauer-Büttiker ballistic electronic transport for multi-lead devices containing thousands of atoms. The method is implemented in the ONETEP linear-scaling density-functional theory code and uses matrix elements calculated from first-principles. Using a compact yet accurate basis of atom-centred non-orthogonal generalised Wannier functions that are optimised in situ to their unique local chemical environment, the transmission and related properties of very large systems can be calculated efficiently and accurately. Other key features include the ability to simulate devices with an arbitrary number of leads, to compute eigenchannel decompositions, and to run on highly parallel computer architectures. We demonstrate the scale of the calculations made possible by our approach by applying it to the study of electronic transport between aligned carbon nanotubes, with system sizes up to 2360 atoms for the underlying density-functional theory calculation. As a consequence of our efficient implementation, computing electronic transport from first principles in systems containing thousands of atoms should be considered routine, even on relatively modest computational resources.
TRANS-I: A fast calculating computer code for the calculation of reactivity transients
International Nuclear Information System (INIS)
In literature is shown that the adiabatic and the quasistatic approximation to space time neutron kinetics are generally fast and conservative methods for calculating reactivity transients. Nevertheless if a feedback reactivity is considered these methods predict too high values of peak flux, energy production and temperature. It is demonstrated, that the deficiency of adiabatic and quasistatic method can be removed, if the mean fuel temperature is multiplied by a weighting factor to get a corrected temperature for calculating Doppler-feedback. The code TRANS-I including this modification is presented. (author)
Codes complex for quick transport 3D neutron calculations of WWER
International Nuclear Information System (INIS)
Complex Surface Values System that capable to carry out the all stages of neutron physical stationary calculations for reactors WWER and PWR is presented. This complex based oneself on using the Surface Values Methods. The approach consists in maximum possible account of specific features in reactor calculations. There is a certain amount of specific features in reactor problems due attention to which is crucial for attaining the main goal-organizing of reactor code.s complexes those provide economical and safe reactors' operating. It is important for the estimation of a code quality to fix of the methodical component but it is necessary to have in view of scale of other ones of result uncertainties. We mention in passing it afterwards. Complex Surface Values System present the approach of building computational reactor models that account for the specifics of reactor problems outlined above. This approach is called the methods of surface values. It utilizes method of surface pseudo-sources for calculating cells within active cores and method of surface harmonics for calculating the whole core or certain sub-assemblies that contain several cells. The part of total complex - the complex Surface Values Lattices - is destined for cells, assembly and sub-assembly calculations of thermal water reactors. The complex Surface Values Lattices use micro constant libraries prepared in the format WIMS. There are libraries in such format based on different initial files of valued data (ENDBF, JEFF, UKNDL, JENDL) on site IAEA. We used these libraries for comparing and choused the one similar to recommended IAEA library. The code Surface Values Core is contained in the total complex Surface Values System besides the complex Surface Values Lattice. This code provide the total scale calculations of reactor's cores. The equations of the Surface Harmonics Method are suitable well for inside reactor core. Other approximations are necessary for neutron description inside reflector
Fragmentation calculation by intranuclear-cascade-evaporation code
Energy Technology Data Exchange (ETDEWEB)
Shigyo, Nobuhiro; Iga, Kiminori; Ishibashi, Kenji [Kyushu Univ., Fukuoka (Japan)
1997-03-01
High Energy Transport Code (HETC) based on the intranuclear-cascade-evaporation model is modified for calculating the fragmentation cross section. For the intranuclear-cascade process, nucleon-nucleon cross sections are used for collision computation; effective in-medium-corrected cross sections are adopted instead of the original free-nucleon collision. The exciton model is adopted for improvement of backward nucleon-emission cross section for low-energy nucleon-incident events. The fragmentation reaction is incorporated into the original HETC as a subroutine set by the use of the systematics of the reaction. The modified HETC (HETC-3STEP/FRG) reproduces experimental fragment yields to a reasonable degree. (author)
Some benchmark calculations for VVER-1000 assemblies by WIMS-7B code
International Nuclear Information System (INIS)
Our aim in this report is to compare of calculation results, obtained with the use of different libraries, which are in the variant of the WIMS7B code. We had the three libraries: the 1986 library is based on the UKNDL files, the two 1996 libraries are based on the JEF-2.2 files, the one having the 69 group approximation, the other having the 172 group approximation. We wanted also to have some acquaintance with the new option of WIMS-7B - CACTUS. The variant of WIMS-7B was placed at our disposal by the code authors for a temporal use for 9 months. It was natural to make at comparisons with analogous values of TVS-M, MCU, Apollo-2, Casmo-4, Conkemo, MCNP, HELIOS codes, where the other different libraries were used. In accordance with our aims the calculations of unprofiled and profiled assemblies of the VVER-1000 reactor have been carried out by the option CACTUS. This option provides calculations by the characteristics method. The calculation results have been compared with the K∞ values obtained by other codes in work. The conclusion from this analysis is such: the methodical parts of errors of these codes have nearly the same values. Spacing for Keff values can be explained of the library microsections differences mainly. Nevertheless, the more detailed analysis of the results obtained is required. In conclusion the calculation of a depletion of VVER-1000 cell has been carried out. The comparison of the dependency of the multiply factor from the depletion obtained by WIMS-7B with different libraries and by the TVS-M, MCU, HELIOS and WIMS-ABBN codes in work has been performed. (orig.)
Relative efficiency calculation of a HPGe detector using MCNPX code
International Nuclear Information System (INIS)
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 60Co 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 60Co 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)
Computer code for calculating reliability/availability of technical systems
International Nuclear Information System (INIS)
Three computer codes are reviewed, which can be applied to reliability analyses of technical systems. They are based on the fault tree and the laws of probability theory. The codes can be used for both non-repairable and repairable systems. The simulation code REMO 79 and the analytical code RELAV are based on the conception that a failure of system components is immediately detected and repaired. The model of the FUPRO2 code provides for failures to be detected and repaired only in periodic functional tests. Apart from code descriptions experience and far-reaching aspects resulting from modularization of the fault trees are summarized. (author)
Benchmark calculation of nuclear design code for HCLWR
International Nuclear Information System (INIS)
In the calculation of the lattice cell for High Conversion Light Water Reactors, big differences of nuclear design parameters appear between the results obtained by various methods and nuclear data libraries. The validity of the calculation can be verified by the critical experiment. The benchmark calculation is also efficient for the estimation of the validity in wide range of lattice parameters and burnup. As we do not have many measured data. The benchmark calculations were done by JAERI and MAPI, using SRAC and WIMS-E respectively. The problem covered the wide range of lattice parameters, i.e., from tight lattice to the current PWR lattice. The comparison was made on the effective multiplication factor, conversion ratio, and reaction rate of each nuclide, including burnup and void effects. The difference of the result is largest at the tightest lattice. But even at that lattice, the difference of the effective multiplication factor is only 1.4 %. The main cause of the difference is the neutron absorption rate U-238 in resonance energy region. The difference of other nuclear design parameters and their cause were also grasped. (author)
Physical Model and Calculation Code for Fuel Coolant Interactions
International Nuclear Information System (INIS)
The base of the physical model of the FCI in discussion are the shock tube experiments performed with UO2 and Na. The experiments demonstrated that the process of 'vapour explosion' in constrained configurations consists of successive cycles. According to the experimental results, the interaction model described here divides each cycle into three phases: Fuel-Coolant-Contact (Phase A), Ejection and reentry of the coolant (Phase B), Impact and Fragmentation (Phase C). The results of some calculations, performed for different values of the system pressure resp. of the coolant bulk temperature, are compiled in plots displaying the series of successive ejection events of an interaction; the plots represent the ejection height and the corresponding pressure in the vapour volume versus time. The pressure peaks mark the impact pressure pulse resp. the vapour pressure immediately after contact, which-ever is greater; the acoustic pressure peak of Phase A has not been plotted. The third and the following cycles reveal the oscillating reentry behaviour. A physical model has been proposed to describe fuel coolant interactions in shock-tube geometry. The corresponding code is presently in an advanced state of development. A principal feature, of the code is the consistent application of the Fourier-equation throughout the whole interaction process; this reveals some peculiarities which do not become evident otherwise. The two dimensional representation of the coolant flow provides the basis for a fragmentation mechanism, thus relieving from the necessity to introduce the fragmentation as an input parameter. These features may indicate a step towards a more realistic comprehension of the subject
Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment
International Nuclear Information System (INIS)
Highlights: • MCNP and ORIGEN are coupled to perform nuclides depletion and decay calculation. • Coupled system MCORE uses “modified predictor corrector” approach. • MCORE can use different depletion schemes and simulate fuel shuffling. • MCORE is assessed by a “VVER-1000 LEU Assembly Computational Benchmark”. • MCORE is also assessed by a fast reactor benchmark problem. - Abstract: An MCNP–ORIGEN burn-up calculation code system, named MCORE (MCNP and ORIGEN burn-up Evaluation code), is developed in this work. MCORE makes use of the Monte Carlo neutron and photon transport code MCNP4C and nuclides depletion and decay calculation code ORIGEN2.1. MCNP and ORIGEN are coupled by data processing and linking subroutines. In MCORE, a so called “modified predictor corrector” approach is used. MCORE provides the capability of using different depletion calculation schemes and simulating fuel shuffling. Total nuclide density changes in active cells are considered in MCORE. The validity and applicability of the developed code are tested by investigating and predicting the neutronic and isotopic behavior of a “VVER-1000 LEU Assembly Computational Benchmark” at lattice level and a “Physics of Plutonium Recycling” fast reactor at core level (OECD-NEA). The comparison results show that the MCORE code predicts the nuclide composition within 5% accuracy and k∞ within 800 pcm at the end of the burn-up for LEU assembly (40 MWD/kg HM). For a fast reactor, the results obtained by MCORE are in the range of reported results except for 243Am. In general, MCORE results show a good agreement with the benchmark values
Fuel rod under power oscillations; calculations with the ENIGMA code
International Nuclear Information System (INIS)
Power oscillations in a BWR may result from a series of events starting from a re-circulation pump trip or can be initiated during start-up at low-flow conditions by other perturbations. Whole core and regional oscillations have been observed. Severe consequences may be anticipated if the instability diverges and the reactor protection system fails (no scram) in all phases of the incident (ATWS). Power peaks higher than ten times of the pre-transient power level have been speculated to appear. Low-magnitude oscillations have been observed at the TVO plant, Olkiluoto 1987, and at the Lasalle-2 plant, 1988, and in other BWRs world-wide. Typically, a boiling water reactor has an unstable operational point at low flow and high power conditions. The physical phenomenon behind the instability is density wave oscillations leading to boiling boundary oscillations and void fraction fluctuations across the heated channel. These in turn, make the fission power vary. The typical frequency of the oscillations seems to be of the order of 0.5 Hz, and thus the power peak for a fuel rod is considerably wider than a RIA-pulse, for instance. Large oscillations can result in elevated fuel temperatures, accelerated fission gas release and additional internal loads on the cladding. These effects may be more severe for a high burnup rod with a large fission gas inventory and a closed gap. Therefore, an experiment has been proposed to be conducted at Halden reactor for simulating the fuel rod response under power oscillations. As there is lack of knowledge also on the relevant boundary conditions, pre-calculations with various input options have been performed and are further suggested. Calculations with FRAPTRAN code have shown the importance of the cladding-coolant heat transfer to the fuel temperature. The applicability of the ENIGMA code to this kind of transients was confirmed. To support the planning of the proposed Halden test, estimates on fuel and cladding temperatures as well as
Calculation of cell volumes and surface areas in MCNP
International Nuclear Information System (INIS)
MCNP is a general Monte Carlo neutron-photon particle transport code which treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces. It is necessary to calculate cell volumes and surface areas so that cell masses, fluxes, and other important information can be determined. The volume/area calculation in MCNP computes cell volumes and surface areas for cells and surfaces rotationally symmetric about any arbitrary axis. 5 figures, 1 table
14 CFR 234.8 - Calculation of on-time performance codes.
2010-01-01
... (AVIATION PROCEEDINGS) ECONOMIC REGULATIONS AIRLINE SERVICE QUALITY PERFORMANCE REPORTS § 234.8 Calculation of on-time performance codes. (a) Each reporting carrier shall calculate an on-time performance code... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Calculation of on-time performance...
Calculations of neutron penetration through graphite medium with Monte Carlo code MCNP
International Nuclear Information System (INIS)
Experiments for fast neutron penetration through graphite are analysed with the continuous energy Monte Carlo code MCNP. Reaction rates and energy spectra obtained with the MCNP are compared with measured values and calculated ones with McBEND code. And validity of penetration calculation with the MCNP is comfirmed. In addition, it is revealed that the MCNP code using Weight-Window method is well applicable to calculations of neutron penetration through graphite up to 70 cm in depth. (author)
Computer codes used in the calculation of high-temperature thermodynamic properties of sodium
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
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)
Model Calculation of Fission Product Yields Data using GEF Code
International Nuclear Information System (INIS)
Fission yields data are classified with spontaneous fission data and neutron induced fission data. The fission product yields data at several energy points for the limited actinides are included in nuclear data libraries such as ENDF/B, JEFF and JENDL because production of those is based mainly on experimental results and it is very difficult to conduct experiments for all actinides and continuous energies. Therefore, in order to obtain fission yields data without experimental data, a theoretical fission model should be introduced to produce the yields data. GEneral Fission model (GEF) is developed to predict the properties for fissioning systems that have not been measured and that are not accessible to experiment. In this study, the fission yields data generated from GEF code are compared with the measured data and the recently available nuclear data libraries. The GEF code is very powerful tool to generate fission yields without measurements. Also, it can produce the distribution of fission product yields for continuous neutron energy while measured data are given only at several energies. The fission yields data of 235U have been tentatively generated with GEF code in this work. Comparing GEF results with measurements and recently released evaluated fission yields data, it is confirmed that GEF code can successfully predict the fission yields data. With its sophisticated model, GEF code is playing a significant role in nuclear industry
NUFACE: An interface code for the calculation of nuclear responses
International Nuclear Information System (INIS)
The NUFACE interface code computes nuclear responses for use in the nuclear analysis of a given tokamak reactor design. The NUFACE code operates on the neutron and gamma fluxes provided by the one-dimensional neutral-particle transport code ONEDANT. Zonewise and zone-boundary responses are computed to obtain both zone-integrated values and maximum surface values. Information on each material mixture within a zone and on each element or isotope constituent of each material is computed. This feature allows for a detailed analysis of the reactor whereby one can easily identify the fractional contribution to the response of interest from each material and each element or isotope. 4 refs., 4 figs., 3 tabs
NUFACE: An interface code for the calculation of nuclear responses
Energy Technology Data Exchange (ETDEWEB)
Henderson, D.L. (Oak Ridge National Lab., TN (USA)); Gomes, I.C. (Tennessee Univ., Knoxville, TN (USA))
1990-01-01
The NUFACE interface code computes nuclear responses for use in the nuclear analysis of a given tokamak reactor design. The NUFACE code operates on the neutron and gamma fluxes provided by the one-dimensional neutral-particle transport code ONEDANT. Zonewise and zone-boundary responses are computed to obtain both zone-integrated values and maximum surface values. Information on each material mixture within a zone and on each element or isotope constituent of each material is computed. This feature allows for a detailed analysis of the reactor whereby one can easily identify the fractional contribution to the response of interest from each material and each element or isotope. 4 refs., 4 figs., 3 tabs.
ERINNI an optical model Fortran IV code for the calculation of multiple cascading particle emissions
International Nuclear Information System (INIS)
ERINNI is an extension of the CERBERO code to the calculations of compound nucleus decay cascades. Up to three successive decays are considered. Optical model transmission coefficients are internally calculated. (authors0
Revised SWAT. The integrated burnup calculation code system
International Nuclear Information System (INIS)
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
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)
Computer code for double beta decay QRPA based calculations
International Nuclear Information System (INIS)
The computer code developed by our group some years ago for the evaluation of nuclear matrix elements, within the QRPA and PQRPA nuclear structure models, involved in neutrino-nucleus reactions, muon capture and β± processes, is extended to include also the nuclear double beta decay
MUXS: a code to generate multigroup cross sections for sputtering calculations
International Nuclear Information System (INIS)
This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc
Improvement and test calculation on basic code or sodium-water reaction jet
Energy Technology Data Exchange (ETDEWEB)
Saito, Yoshinori; Itooka, Satoshi [Advanced Reactor Engineering Center, Hitachi Works, Hitachi Ltd., Hitachi, Ibaraki (Japan); Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo [Consulting Engineering Dept., Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan)
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{center_dot}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)
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)
KARATE - a code for VVER-440 core calculation
Energy Technology Data Exchange (ETDEWEB)
Gado, J.; Hegedus, Cs.J.; Hegyi, Gy.; Kereszturi, A.; Makai, M.; Maraczi, Cs.; Telbisz, M.
1994-12-31
A modular calculation system has been elaborated at the KFKI Atomic Energy Research Institute for VVER-440 cores. The purpose of KARATE is the calculation of neutron physical and thermal-hydraulic processes in the core at normal, startup, and slow transient conditions. KARATE is under validation and verification (V&V) against mathematical, experimental, and operational data.
International Nuclear Information System (INIS)
The Carmen code (theory and user's manual) is described. This code for assembly and core calculations uses diffusion theory (Citation), with feedback in the cross sections by zone due to the effects of burnup, water density, fuel temperature, Xenon and Samarium. The burnup calculation of a full cycle is solved in only an execution of Carmen, and in a reduced computer time. (auth.)
International Nuclear Information System (INIS)
A fast running computer code SHETEMP has been developed for analysis of reactivity initiated accidents under constant core cooling conditions such as coolant temperature and heat transfer coefficient on fuel rods. This code can predict core power and fuel temperature behaviours. A control rod movement can be taken into account in power control system. The objective of the code is to provide fast running capability with easy handling of the code required for audit and design calculations where a large number of calculations are performed for parameter surveys during short time period. The fast running capability of the code was realized by neglection of fluid flow calculation. The computer code SHETEMP was made up by extracting and conglomerating routines for reactor kinetics and heat conduction in the transient reactor thermal-hydraulic analysis code ALARM-P1, and by combining newly developed routines for reactor power control system. As ALARM-P1, SHETEMP solves point reactor kinetics equations by the modified Runge-Kutta method and one-dimensional transient heat conduction equations for slab and cylindrical geometries by the Crank-Nicholson methods. The model for reactor power control system takes into account effects of PID regulator and control rod drive mechanism. In order to check errors in programming of the code, calculated results by SHETEMP were compared with analytic solution. Based on the comparisons, the appropriateness of the programming was verified. Also, through a sample calculation for typical modelling, it was concluded that the code could satisfy the fast running capability required for audit and design calculations. This report will be described as a code manual of SHETEMP. It contains descriptions on a sample problem, code structure, input data specifications and usage of the code, in addition to analytical models and results of code verification calculations. (author)
Code-B-1 for stress/strain calculation for TRISO fuel particle (Contract research)
International Nuclear Information System (INIS)
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)
Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation
International Nuclear Information System (INIS)
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)
A new assembly-level Monte Carlo neutron transport code for reactor physics calculations
International Nuclear Information System (INIS)
This paper presents a new assembly-level Monte Carlo neutron transport code, specifically intended for diffusion code group-constant generation and other reactor physics calculations. The code is being developed at the Technical Research Centre of Finland (VTT), under the working title 'Probabilistic Scattering Game', or PSG. The PSG code uses a method known as Woodcock tracking to simulate neutron histories. The advantages of the method include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. The main drawback is the inability to calculate reaction rates in optically thin volumes. This narrows the field of application to calculations involving parameters integrated over large volumes. The main features of the PSG code and the Woodcock tracking method are introduced. The code is applied in three example cases, involving infinite lattices of two-dimensional LWR fuel assemblies. Comparison calculations are carried out using MCNP4C and CASMO-4E. The results reveal that the code performs quite well in the calculation cases of this study, especially when compared to MCNP. The PSG code is still under extensive development and there are both flaws in the simulation of the interaction physics and programming errors in the source code. The results presented here, however, seem very encouraging, especially considering the early development stage of the code. (author)
Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)
International Nuclear Information System (INIS)
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.)
Computer code ANISN multiplying media and shielding calculation 2. Code description (input/output)
International Nuclear Information System (INIS)
The new code CCC-0514-ANISN/PC is described, as well as a ''GENERAL DESCRIPTION OF ANISN/PC code''. In addition to the ANISN/PC code, the transmittal package includes an interactive input generation programme called APE (ANISN Processor and Evaluator), which facilitates the work of the user in giving input. Also, a 21 group photon cross section master library FLUNGP.LIB in ISOTX format, which can be edited by an executable file LMOD.EXE, is included in the package. The input and output subroutines are reviewed. 6 refs, 1 fig., 1 tab
SENVAR: a code for handling chemical uncertainties in solubility calculations
International Nuclear Information System (INIS)
In the planning for a repository for spent nuclear fuel it is important to know the solubility of some important solid phases in order to, for example, predict migration of radionuclides from the repository. The method presented in the present paper investigates the effect of uncertainties in thermodynamical data, i.e. stability and solubility constants, for the calculated solubility of a solid phase. The adopted approach is simple Monte Carlo sampling. The investigation is mainly made in three steps. First a preliminary sensitivity analysis where the important parameters are determined. This is done by holding each of the parameters at a fixed value for a given number of solubility calculations. During this time all other parameters are varied according to a pre-set random matrix. The variance for each stationary parameter is then calculated and the one with the smallest variance is deemed the most important one and so on. The parameters that are deemed important are then transferred into the uncertainty analysis. There, each parameter may be given a separate interval for the uncertainty and then a couple of thousand solubility calculations are made where the values of the parameters are varied according to the Monte Carlo method. The results from these calculations are used to estimate the effect of the uncertainties in a plot showing the density function of the solubility and some statistical estimators. The solubility calculations are also used to give data to a stepwise regression program which estimates the importance of each parameter entered into the uncertainty analysis. The regression error is also shown in order to make it easy to determine which values may be correct or not
Aspects of cell calculations in deterministic reactor core analysis
Energy Technology Data Exchange (ETDEWEB)
Varvayanni, M. [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Savva, P., E-mail: savvapan@ipta.demokritos.gr [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N. [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)
2011-02-15
{Tau}he capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available
The use of the codes from MCU family for calculations of WWER type reactors
International Nuclear Information System (INIS)
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
Energy Technology Data Exchange (ETDEWEB)
Leal, L.C.; Larson, N.M.
1995-09-01
The SAMDIST computer code has been developed to calculate distribution of resonance parameters of the Reich-Moore R-matrix type. The program assumes the parameters are in the format compatible with that of the multilevel R-matrix code SAMMY. SAMDIST calculates the energy-level spacing distribution, the resonance width distribution, and the long-range correlation of the energy levels. Results of these calculations are presented in both graphic and tabular forms.
Blockage calculation of LMFBR core subassembly with subchannel code-SOBOS
International Nuclear Information System (INIS)
Sodium-boiling is a very important subject to be considered in the Liquid Metal Fast Breeder Reactor (LMFBR) design. Blockage is one of the most important causes of sodium boiling. The author shows the calculation results with subchannel code 'SOBOS' with the advanced subchannel model. And the results of calculations match that of experiments very well, indicating that the subchannel code could be used to calculate the blockage boiling
Validation of capture yield calculations in the Resolved Resonance Energy Range with CONRAD code
Litaize, Olivier; Archier, Pascal; Becker, Bjorn; Schillebeeckx, Peter; Kopecky, Stefan
2013-03-01
This paper deals with the validation of the multiple scattering corrections developed in the CONRAD code for the capture yield calculations in the Resolved Resonance energy Range (RRR). In order to calculate the capture yields, analytic and stochastic calculation schemes implemented in CONRAD are described and compared with the analysis code SAMMY/SAMSMC. The results are in excellent agreement for a variety of samples. We concentrate the discussion here on 238U, 197Au and 55Mn.
Validation of capture yield calculations in the Resolved Resonance Energy Range with CONRAD code
Directory of Open Access Journals (Sweden)
Schillebeeckx Peter
2013-03-01
Full Text Available This paper deals with the validation of the multiple scattering corrections developed in the CONRAD code for the capture yield calculations in the Resolved Resonance energy Range (RRR. In order to calculate the capture yields, analytic and stochastic calculation schemes implemented in CONRAD are described and compared with the analysis code SAMMY/SAMSMC. The results are in excellent agreement for a variety of samples. We concentrate the discussion here on 238U, 197Au and 55Mn.
Numerical identification of bacteria with a hand-held calculator as an alternative to code books.
Schindler, J; Schindler, Z
1982-01-01
The Hewlett-Packard HP 41C hand-held calculator can be used for the numerical identification of bacteria. The dimensions of the identification matrix are limited to about 30 by 22; however, many groups of clinically important bacteria can be numerically identified by this method. Hand-held calculators can be used as an alternative to code books. At present, these calculators and additional tests can help solve identification problems in profiles not contained in code books.
Benchmark calculations by the nuclear criticality safety analysis code system JACS(MGCL, KENO-IV)
International Nuclear Information System (INIS)
Since 1980, as many as 1394 cases of benchmark calculations on criticality problems have been performed by the KENO-IV Monte Carlo calculation code with the MGCL cross section data library. The code system is a part of the criticality safety evaluation code system JACS developed at JAERI. The code validation results have been published in a series of JAERI-M reports and others. This report summarizes these results and the reliability of the code system systematically. The number of the calculated cases briefly described in this report together with their experimental systems and data are 502 for 17 kinds of homogeneous single-unit systems, 331 for 8 kinds of homogeneous multi-unit systems and 561 for 16 kinds of heterogeneous systems. Discussions and interpretations are made on the calculated keff's (neutron multiplication factors) with their bias errors. The factors related to the bias errors are confirmed together with their causes and trends. (author)
A CFD validation methodology for containment code calculations of hydrogen mixing and recombination
International Nuclear Information System (INIS)
In the frame of ANSALDO activities on containment hydrogen accident events, a simulation procedure was developed to qualify and verify the calculations performed by simplified containment computer codes through the use of a full 3-D Navier Stokes solver. The methodology aims to reduce the computational time usually associated with a general purpose CFD code complete simulation of the containment transient, limiting on the other hand the loss of accuracy typical of the use of a simplified Containment Code. This goal has been fulfilled by the development of a calculation procedure organised in several different steps able to verify the calculated transient parameters by GOTHIC3.4 (the simplified code) with specific calculations performed with CFX4.2 (the CFD code). The paper describes the main milestones of the methodology development and summarizes main results, findings, as well the possible direction of use of the performed work. (authors)
Quasiparticle GW calculations within the GPAW electronic structure code
DEFF Research Database (Denmark)
Hüser, Falco
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...... 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...... respect to the system one wants to investigate by choosing a certain functional or by tuning parameters. A succesful alternative is the so-called GW approximation. It is mathematically precise and gives a physically well-founded description of the complicated electron interactions in terms of screening...
Full-core pin-power calculations using Monte Carlo codes
International Nuclear Information System (INIS)
Pin wise calculations of core power distribution have been performed for a criticality mock up installation that models a WWER-1000 reactor. Two Monte Carlo codes have been applied for solving of this problem: the MCNP4B code and the KENO-VI code from the SCALE 4.4 system. The codes use different kinds of neutron cross section data: pointwise continuous-energy ENDF/B-VI data and multigroup ENDF/B-V data. Comparisons of calculated results show that the MCNP4B and KENO-VI results are in good agreement. (authors)
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.
International Nuclear Information System (INIS)
There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0. (author)
International Nuclear Information System (INIS)
The authors report calculations performed using the MNCP and PENELOPE codes to determine the Hp(3)/K air conversion coefficient which allows the Hp(3) dose equivalent to be determined from the measured value of the kerma in the air. They report the definition of the phantom, a 20 cm diameter and 20 cm high cylinder which is considered as representative of a head. Calculations are performed for an energy range corresponding to interventional radiology or cardiology (20 keV-110 keV). Results obtained with both codes are compared
A transport based one-dimensional perturbation code for reactivity calculations in metal systems
Energy Technology Data Exchange (ETDEWEB)
Wenz, T.R.
1995-02-01
A one-dimensional reactivity calculation code is developed using first order perturbation theory. The reactivity equation is based on the multi-group transport equation using the discrete ordinates method for angular dependence. In addition to the first order perturbation approximations, the reactivity code uses only the isotropic scattering data, but cross section libraries with higher order scattering data can still be used with this code. The reactivity code obtains all the flux, cross section, and geometry data from the standard interface files created by ONEDANT, a discrete ordinates transport code. Comparisons between calculated and experimental reactivities were done with the central reactivity worth data for Lady Godiva, a bare uranium metal assembly. Good agreement is found for isotopes that do not violate the assumptions in the first order approximation. In general for cases where there are large discrepancies, the discretized cross section data is not accurately representing certain resonance regions that coincide with dominant flux groups in the Godiva assembly. Comparing reactivities calculated with first order perturbation theory and a straight {Delta}k/k calculation shows agreement within 10% indicating the perturbation of the calculated fluxes is small enough for first order perturbation theory to be applicable in the modeled system. Computation time comparisons between reactivities calculated with first order perturbation theory and straight {Delta}k/k calculations indicate considerable time can be saved performing a calculation with a perturbation code particularly as the complexity of the modeled problems increase.
A transport based one-dimensional perturbation code for reactivity calculations in metal systems
International Nuclear Information System (INIS)
A one-dimensional reactivity calculation code is developed using first order perturbation theory. The reactivity equation is based on the multi-group transport equation using the discrete ordinates method for angular dependence. In addition to the first order perturbation approximations, the reactivity code uses only the isotropic scattering data, but cross section libraries with higher order scattering data can still be used with this code. The reactivity code obtains all the flux, cross section, and geometry data from the standard interface files created by ONEDANT, a discrete ordinates transport code. Comparisons between calculated and experimental reactivities were done with the central reactivity worth data for Lady Godiva, a bare uranium metal assembly. Good agreement is found for isotopes that do not violate the assumptions in the first order approximation. In general for cases where there are large discrepancies, the discretized cross section data is not accurately representing certain resonance regions that coincide with dominant flux groups in the Godiva assembly. Comparing reactivities calculated with first order perturbation theory and a straight Δk/k calculation shows agreement within 10% indicating the perturbation of the calculated fluxes is small enough for first order perturbation theory to be applicable in the modeled system. Computation time comparisons between reactivities calculated with first order perturbation theory and straight Δk/k calculations indicate considerable time can be saved performing a calculation with a perturbation code particularly as the complexity of the modeled problems increase
Validation of the Monteburns code for criticality calculation of TRIGA reactors
Energy Technology Data Exchange (ETDEWEB)
Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Jeraj, Robert [Jozef Stafan Institute, Ljubljana (Slovenia)
2002-07-01
Use of Monte Carlo methods in burnup calculations of nuclear fuel has become practical due to increased speed of computers. Monteburns is an automated computational tool that links the Monte Carlo code MCNP with the burnup and decay code ORIGEN2.1. This code system was used to simulate a criticality benchmark experiment with burned fuel on a TRIGA Mark II research reactor. Two core configurations were simulated and k{sub eff} values calculated. The comparison between the calculated and experimental values shows good agreement, which indicates that the MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulations of TRIGA reactors. (author)
Radiative Transfer Code: Application to the calculation of PAR
Indian Academy of Sciences (India)
D Emmanuel; D Phillippe; C Malik
2000-12-01
The production of carbon in the ocean, the so-called primary production, depends on various physico- biological parameters: the biomass and nutrient amounts in oceans, the salinity and temperature of the water and the light available in the water column. We focus on the visible spectrum of the solar radiation defined as the Photosynthetically Active Radiation (PAR). We developed a model (Chami et al. 1997) to simulate the behavior of the solar beam in the atmosphere and the ocean. We first describe the theoretical basis of the code and the method we used to solve the radiative transfer equation (RTE): the successive orders of scattering (SO). The second part deals with a sensitivity study of the PAR just above and below the sea surface for various atmospheric conditions. In a cloudy sky, we computed a ratio between vector fluxes just above the sea surface and spherical fluxes just beneath the sea surface. When the optical thickness of the cloud increases this ratio remains constant and around 1.29. This parameter is convenient to convert vector flux at the sea surface as retrieved from satellite to PAR. Subsequently, we show how solar radiation as vector flux rather than PAR leads to an underestimate of the primary production up to 40% for extreme cases.
Development of leak rate calculation model and code in piping
International Nuclear Information System (INIS)
Background: With the development of fracture mechanics, Leak-Before-Break (LBB) is widely used in nuclear power plant piping design. Purpose: In order to support the application of LBB, leak rate through crack need to be calculated. Methods: In this text, an analytical flow model is developed based on homogeneous non-equilibrium model, a computer program is also developed based on this analytical model. Results: Comparison between the results from the above program and test results shows that deviations of calc. results to test results are within ±50%. Conclusions: Conclusions can be got that this analytical model and computer program meet the requirement of engineering application. (authors)
Burnup calculations using the ORIGEN code in the CONKEMO computing system
International Nuclear Information System (INIS)
This article describes the CONKEMO computing system for kinetic multigroup calculations of nuclear reactors and their physical characteristics during burnup. The ORIGEN burnup calculation code has been added to the system. The results of an international benchmark calculation are also presented. (author)
Code of hybrid calculation for the study of heat exchangers
International Nuclear Information System (INIS)
A series integration method has been chosen, machine time being similar to space, the computer storing the solutions passed to allow for an approximation at the finished differences of the timed derivatives. Space distributions of the functions of state (enthalpy and temperature) of the two fluids are calculated by analogical integration in the direction of the circulation of the fluids. This is made possible by the disconnection of the two fluids thanks to a method of prediction-correction for the equation of the wall. The great rapidity of the analogical integrations and the absence of iterations make the calculation in actual time possible (assuming that 200 points are taken for the storage of the solutions passed, the time required for a full calculation is 0.4 second). It is, therefore, possible to integrate this simulation in an actual (for example Nuclear Station) or simulated loop. The exchanger may or may not be with counter-current, with or without change of phase. Thermodynamic tables with two variables are introduced into the computer, interpolation is effected in analogical and this allows, inter alia, enthalpy, temperature, pressure and volumic mass to be perfectly in phase one with the other. The work bears most particularly on the accuracy and stability of simulation of the timed derivatives which are obtained by difference between an analogical function and the same function stored in the computer at the preceding step. Particular emphasis is made to bear on the operation of the model, a monitor program controlling the various phases of utilization in function of the type of exchanger, experiments to be made; a program of control of the results enables the operator to obtain the results either immediately in printed form, curve tracer or oscilloscope, put to scale, or in deferred a more elaborate processing of the results. The same program allows for the simulation of any exchanger (different fluid, variable geometry, etc.) or any number of
Method of tallying adjoint fluence and calculating kinetics parameters in Monte Carlo codes
International Nuclear Information System (INIS)
A method of using iterated fission probability to estimate the adjoint fluence during particles simulation, and using it as the weighting function to calculate kinetics parameters βeff and A in Monte Carlo codes, was introduced in this paper. Implements of this method in continuous energy Monte Carlo code MCNP and multi-group Monte Carlo code MCMG are both elaborated. Verification results show that, with regardless additional computing cost, using this method, the adjoint fluence accounted by MCMG matches well with the result computed by ANISN, and the kinetics parameters calculated by MCNP agree very well with benchmarks. This method is proved to be reliable, and the function of calculating kinetics parameters in Monte Carlo codes is carried out effectively, which could be the basement for Monte Carlo codes' utility in the analysis of nuclear reactors' transient behavior. (authors)
Study on the GEANT4 code applications to the dose calculation using imaging data
Lee, JeongOk; Kim, JhinKee; Kwon, HyeongCheol; Kim, JungSoo; Kim, BuGil; Jeong, DongHyeok
2015-01-01
The use of GEANT4 code has increased in the medical field. There are various studies to calculate the patient dose distributions with the GEANT4 code using the imaging data. In present study, the Monte Carlo simulations based on the DICOM data were performed to calculate absorbed dose in the patient's body. Various visualization tools were equipped in the GEANT4 code to display the detector construction, however there are limitations to display the DICOM images. In addition, it is difficult to display the dose distributions on the imaging data of the patient. Recently, gMocren code, volume visualization tool for GEANT4 simulation, has been developed and used in volume visualization of image files. In this study, the imaging data based absorbed dose distributions in patient were performed by using the gMocren code. The dosimetric evaluations with TLD and film dosimetry methods were carried out to verify the calculation results.
Study on GEANT4 code applications to dose calculation using imaging data
Lee, Jeong Ok; Kang, Jeong Ku; Kim, Jhin Kee; Kwon, Hyeong Cheol; Kim, Jung Soo; Kim, Bu Gil; Jeong, Dong Hyeok
2015-07-01
The use of the GEANT4 code has increased in the medical field. Various studies have calculated the patient dose distributions by users the GEANT4 code with imaging data. In present study, Monte Carlo simulations based on DICOM data were performed to calculate the dose absorb in the patient's body. Various visualization tools are installed in the GEANT4 code to display the detector construction; however, the display of DICOM images is limited. In addition, to displaying the dose distributions on the imaging data of the patient is difficult. Recently, the gMocren code, a volume visualization tool for GEANT4 simulation, was developed and has been used in volume visualization of image files. In this study, the imaging based on the dose distributions absorbed in the patients was performed by using the gMocren code. Dosimetric evaluations with were carried out by using thermo luminescent dosimeter and film dosimetry to verify the calculated results.
International Nuclear Information System (INIS)
A short review is given for models using in thermohydraulic code HYDRA-IBRAE/LM and the results of verification of calculational code on the problems of liquid metal coolant flow and heat transfer. It is shown that developed version of code HYDRA-IBRAE/LM simulates one-phase flow of lead, sodium and lead-bismuth coolants with high accuracy and the processes of sodium boiling with good one. The results of applied calculations of once-through steam generators are considered. It is pointed out that code HYDRA-IBRAE/LM represents correctly physics of the processes and phenomena taking place in the steam generator. The results of cross-verification calculations of lead steam generator by codes HYDRA-IBRAE/LM and TRIANA-4 show satisfactory agreement of results on temperatures of coolants and materials of channel walls in Field tube
Chromaticity calculations and code comparisons for x-ray lithography source XLS and SXLS rings
International Nuclear Information System (INIS)
This note presents the chromaticity calculations and code comparison results for the (x-ray lithography source) XLS (Chasman Green, XUV Cosy lattice) and (2 magnet 4T) SXLS lattices, with the standard beam optic codes, including programs SYNCH88.5, MAD6, PATRICIA88.4, PATPET88.2, DIMAD, BETA, and MARYLIE. This analysis is a part of our ongoing accelerator physics code studies. 4 figs., 10 tabs
Chromaticity calculations and code comparisons for x-ray lithography source XLS and SXLS rings
Energy Technology Data Exchange (ETDEWEB)
Parsa, Z.
1988-06-16
This note presents the chromaticity calculations and code comparison results for the (x-ray lithography source) XLS (Chasman Green, XUV Cosy lattice) and (2 magnet 4T) SXLS lattices, with the standard beam optic codes, including programs SYNCH88.5, MAD6, PATRICIA88.4, PATPET88.2, DIMAD, BETA, and MARYLIE. This analysis is a part of our ongoing accelerator physics code studies. 4 figs., 10 tabs.
TEMP: a computer code to calculate fuel pin temperatures during a transient
International Nuclear Information System (INIS)
The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method
User effects on the thermal-hydraulic transient system code calculations
International Nuclear Information System (INIS)
Large thermalhydraulic 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 dealt by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. During these comparisons of the code results, there has been a continuous debate on the way how the code user influences the predicted system behaviour. This rather subjective element might become a crucial point with respect to the quantitative evaluation of the code uncertainties which is essential if the 'best estimate codes are used for licensing procedures'. The International Standard Problem Exercises (ISPs) proposed by the OECD-Committee for the Safety of Nuclear Installations (CSNI) and by IAEA (International Atomic Energy Agency) and thermalhydraulic code assessment activity undertaken by US.NRC under International Code Assessment and Application Program (ICAP) demonstrate the large effort put in this framework by organisations all over the world. 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 the influence of the code user on the calculated results is directly accounted. In this paper, the results of the investigations on the user effects for the thermalhydraulic transient system codes will be 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 which affect the results of the calculations and are hidden under user effects. These reasons and user effects will be discussed in detail and general recommendations and conclusions will be presented to control and limit them. In summary
Fission gas activities in the fuel-to-clad gap calculated with the code FUROM
International Nuclear Information System (INIS)
The fuel behaviour code FUROM (FUel ROd Model) has been in use and under improvement for several years at the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute. Several new features are added to it each year. In the present paper an extended fission gas release model is introduced. This model is suitable for the calculation of the release of not only stable but also radioactive isotopes. Code calculations are compared to international results. (authors)
Full Core, Heterogeneous, Time Dependent Neutron Transport Calculations with the 3D Code DeCART
Hursin, Mathieu
2010-01-01
The current state of the art in reactor physics methods to assess safety, fuel failure, and operability margins for Design Basis Accidents (DBAs) for Light Water Reactors (LWRs) rely upon the coupling of nodal neutronics and one-dimensional thermal hydraulic system codes. The neutronic calculations use a multi-step approach in which the assembly homogenized macroscopic cross sections and kinetic parameters are first calculated using a lattice code for the range of conditions (temperatures, bu...
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.
RADIATION DOSE CALCULATION FOR FUEL HANDLING FACILITY CLOSURE CELL EQUIPMENT
International Nuclear Information System (INIS)
This calculation evaluates the energy deposition rates in silicon, gamma and neutron flux spectra at various locations of interest throughout FHF closure cell. The physical configuration features a complex geometry, with particle flux attenuation of many orders of magnitude that cannot be modeled by computer codes that use deterministic methods. Therefore, in this calculation the Monte Carlo method was used to solve the photon and neutron transport. In contrast with the deterministic methods, Monte Carlo does not solve an explicit transport equation, but rather obtain answers by simulating individual particles, recording the aspects of interest of their average behavior, and estimates the statistical precision of the results
Application of Monte Carlo code EGS4 to calculate gamma exposure buildup factors
International Nuclear Information System (INIS)
Exposure buildup factors up to 40 mean free paths ranging from 0.015 MeV to 15 MeV photon energy were calculated by using the Monte Carlo simulation code EGS4 for ordinary concrete. The calculation involves PHOTX cross section library, a point isotropic source, infinite uniform medium model and a particle splitting method and considers the Bremsstrahlung, fluorescent effect, correlative (Rayleigh) scatter. The results were compared with the relevant data. Results show that the data of the buildup factors calculated by the Monte Carlo code EGS4 was reliable. The Monte Carlo method can be used widely to calculate gamma-ray exposure buildup factors. (authors)
An integrated multi-functional neutronics calculation and analysis code system: VisualBUS
International Nuclear Information System (INIS)
Neutronics calculation and analysis are the bases of reactor physics design, radiation protection, fuel management optimization, nuclear safety analysis, etc. After surveying and evaluating the status and trend of development of neutronics calculation and analysis codes, a network-based integrated multi-functional neutronics calculation and analysis code system has been designed and developed for applications in fusion, fission and various hybrid systems based on the adoption of advanced neutronics calculating approaches and modern computer' software technologies. A series of benchmark tests and applications have shown the maturity and effectiveness of the system. This paper gives a brief overview about main technical features of the system, the benchmark tests and applications. (authors)
International Nuclear Information System (INIS)
The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested
Point reactivity burnup code DELIGHT-4 for high temperature, gas-cooled reactor cells
International Nuclear Information System (INIS)
The code DELIGHT-4 has been developed for analizing burnup characteristics of the graphite moderated reactor cells and producing the few-group constants. Calculation models for the code are as follows: (1) The number of neutron energy groups is 61 for fast neutrons (10 MeV -- 2.38 eV) and 50 for thermal neutrons (2.38 eV -- 0 eV). (2) The doubly space-heterogeneous effect of fuel (dispersion of coated fuel particles in fuel compacts and regular array of fuel rods in graphite blocks) is considered in the calculation of resonance absorption. (3) The double heterogenity of burnable poison (dispersion of absorber grains in rods) can be considered. (4) The chemical binding effect of graphite is introduced in the scattering of thermal neutrons. (5) The calculations of criticality and burnup are by a few-energy-group models (up to 10 groups for both fast and thermal neutrons), and nuclide chains of thorium-uranium and uranium-plutonium are used for burnup calculation. (6) Neutron streaming effect through holes and gaps in cells can be considered in criticality calculation. (7) The flux distribution in cells can be calculated. The cell-averaged few group constants can be produced in card form for 1-D transport approximation code SLALOM, 2-D S sub( n) code TWOTRAN, 1-D diffusion code BRIQUET, 2-D diffusion code ZADOC-3 and 3-D diffusion code CITATION-DEGA. (author)
International Nuclear Information System (INIS)
In a D-T burning fusion reactor, the radioactivity induced by the 14 MeV neutrons causes many problems. It limits personnel access to the reactor during shutdown, generates decay heat and produces radwastes. A code system THIDA had been developed in 1978 to calculate the radioactivity and dose rate around a fusion device. The THIDA system consisted of the followings: one- and two-dimensional discrete ordinates radiation transport codes; induced activity calculation code; three libraries for transmutation and decay chain data, transmutation cross sections and delayed gamma-ray emission data. The present report gives a complete description of THIDA-2, a new advanced version of the THIDA system which has the following major improvements: 1. Capability to treat three-dimensional calculation models by the use of a Monte Carlo transport code. 2. Accurate decay heat calculation following the transport of delayed gamma rays. 3. Simplification of the data input process by the use of free format scheme and closer coupling between the radiation transport codes and the induced activity calculation code. 4. Self-descriptive output format and additional plotter output. 5. Capability to calculate problems requiring larger core memory by the use of variable dimension. (author)
Using deterministic codes to accelerate continuous energy Monte-Carlo standards calculations
International Nuclear Information System (INIS)
Deterministic codes are usually used for critical parameters or one dimension geometry calculations. Advantages of the use of deterministic codes are speed of the calculation and the absence of standard deviation on the keff results. Nevertheless, the deterministic results are affected by several intrinsic uncertainties as energetic condensation or self-shielding. So the way to proceed at CEA expert criticality group (CEA/SERMA/CP2C) is to always check the main results (minimum critical or maximal permissible values and un-moderated values) with a punctual Monte Carlo calculation. These last years, in particular cases (pure actinide fissile media, exotic reflectors), large discrepancies have been observed between the keff calculated by the CRISTAL V1 route reference (continuous energy Monte Carlo code TRIPOLI-4) and the keff target (by the standard route APOLLO2-Sn). The problematic for these cases was how to transpose the keff discrepancies observed between standard and reference routes to the dimensions (mass, thickness...) or how to reduce the keff discrepancies using optimized options of the deterministic code. One solution to transpose discrepancies is to iterate on dimensions using a punctual Monte Carlo code to achieve the desired keff eigenvalue. But, the amount of time for obtaining a good standard deviation and also the desired keff eigenvalue inside the Monte Carlo calculation uncertainty can quickly increase. The principle of the method presented in this paper is that the discrepancy between deterministic code and Monte-Carlo code, calculated at the same dimension, is low variable with the dimension. Therefore, correcting the keff eigenvalue on which the deterministic code converge with the discrepancy observed, leads to a dimension nearer to the true dimension (i.e. the dimension where Monte-Carlo code keff calculation is close to the keff eigenvalue). If the keff eigenvalue is outside the Monte Carlo uncertainty, the discrepancy is recalculated and
International Nuclear Information System (INIS)
The paper aims to present the main physical principles for selection of design characteristics of the fast reactor control rods (CR) system. The brief analysis of problems of CR physical calculations is given. Four components are described for the correction to the control rod worth calculated by the routine method based on the few - group three - dimensional diffusion code (TRIGEX) in hexagonal geometry. Principle considerations are given for the choice of the original task discretization methods implemented in this code to minimize the total error. Brief information is given about methods and codes used for the evaluation of error components of control rod worths calculated in a standard way. The results of experimental and calculational investigations of control rod physical characteristics are presented. These results were obtained at BFS critical assemblies simulating LMFBR cores. The investigations have been carried out for different types of core configurations. The experimental and calculated values are given on the distortion of power distribution due to the control rod insertion in the core. (author). 51 refs, 9 figs, 5 tabs
Refuelling design and core calculations at NPP Paks: codes and methods
International Nuclear Information System (INIS)
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)
User effects on the thermal-hydraulic transient system code calculations
International Nuclear Information System (INIS)
Large thermalhydraulic 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 dealt by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. During these comparisons of the code results, there has been a continuous debate on the way how the code user influences the predicted system behaviour. This rather subjective element might become a crucial point with respect to the quantitative evaluation of the code uncertainties which is essential if the ''best estimate codes are used for licensing procedures''. The International Standard Problem Exercises (ISPs) proposed by the OECD-Committee for the Safety of Nuclear Installations (CSNI) and by IAEA (International Atomic Energy Agency) and thermalhydraulic code assessment activity undertaken by US.NRC under International Code Assessment and Application Program (ICAP) demonstrate the large effort put in this framework by organisations all over the world. 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 the influence of the code user on the calculated results is directly accounted. In this paper, the results of the investigations on the user effects for the thermalhydraulic transient system codes will be 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 which affect the results of the calculations and are hidden under user effects. These reasons and user effects will be discussed in detail and general recommendations and conclusions will be presented to control and limit them. (authors). 15
The GSCAN code: From GNASH reaction calculations to ENDF nuclear data files
International Nuclear Information System (INIS)
The GSCAN code, which is enhanced version of the GNXS code included in the GNASH code package presented at the IAEA Trieste Workshops, utilizes an output made by the Hauser-Feshbach and preequilibrium GNASH reaction model code. Its main purposes are: (1) To convert the calculated cross sections into ENDF-6 format; (2) To calculate the emission spectra of A ≥ 5 secondary particles (recoils) and represent them in ENDF-6 format; and (3) To display all the exclusive reaction channels that contributed to a given inclusive emission channel (production cross section). This code has been widely used at Los Alamos in the production of the high-energy data files that extend up to 150 MeV for incident neutrons and protons, for enhanced radiation transport simulations of accelerator-driven systems. (author)
Boltzmann-Fokker-Planck calculations using standard discrete-ordinates codes
International Nuclear Information System (INIS)
The Boltzmann-Fokker-Planck (BFP) equation can be used to describe both neutral and charged-particle transport. Over the past several years, the author and several collaborators have developed methods for representing Fokker-Planck operators with standard multigroup-Legendre cross-section data. When these data are input to a standard S/sub n/ code such as ONETRAN, the code actually solves the Boltzmann-Fokker-Planck equation rather than the Boltzmann equation. This is achieved wihout any modification to the S/sub n/ codes. Because BFP calculations can be more demanding from a numerical viewpoint than standard neutronics calculations, we have found it useful to implement new quadrature methods ad convergence acceleration methods in the standard discrete-ordinates code, ONETRAN. We discuss our BFP cross-section representation techniques, our improved quadrature and acceleration techniques, and present results from BFP coupled electron-photon transport calculations performed with ONETRAN. 19 refs., 7 figs
Comparison of code calculations with experiments on containment response during LOCA conditions
International Nuclear Information System (INIS)
A series of experiments were performed on a one-tenth scale model of PHWR containment, incorporating pressure suppression system. The pressure-temperature transients in the model containment observed during simulated LOCA (Loss of Coolant) blowdown conditions were compared against calculated results form computer code PACSR, for purposes of verification of the code. Comparison of results indicated that calculated values of peak pressure in various compartment were significantly higher than observed ones. This disagreement was attributed mainly to modelling for energy absorption from containment atmosphere to structural surfaces, this effect being particularly important in a scaled down model. Good agreement between calculation and experiment was obtained after heat transfer correlation for energy absorption on surfaces were modified in the code. The study demonstrates the conservatism of the results from the code. (author). 6 refs., 1 tab., 9 figs
Code Betal to calculation Alpha/Beta activities in environmental samples
International Nuclear Information System (INIS)
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)
A benchmark problem was proposed to reproduce an experiment for target membrane structure cooling of Accelerator Driven System at the 10th meeting of IWGAR (International Working Group of Advanced Nuclear Reactors Thermal Hydraulic) by the Fluid Phenomena in Energy Exchanges Section of IAHR (International Association of Hydraulic Engineering and Research). The benchmark calculation has been carried out with AQUA and FLUENT codes to estimate the code validity for liquid metal thermal-hydraulics application. As a result of comparison between numerical analyses and experiment, it is concluded as follows: Inlet flow rate at the distributing grid much affects a coolant temperature and temperature pulsation near the membrane. The coolant temperature decreases and the pulsation decays rapidly as the flow rate toward the membrane center increases. On downstream of the distributing grid, numerical results agree with experimental data except that numerical analysis tends to overestimate the coolant temperature pulsation. Numerical results show that the decrease of coolant temperature and the dissipation of pulsation tend to be underestimated when the flow rate toward the membrane center increases. In FLUENT code, the dissipation of coolant temperature is underestimated more than in AQUA code because FLUENT code tends to overestimate the flow rate toward the membrane center. But the same tendency of the dissipation behavior is shown in AQUA code. A turbulent model is less influenced on the coolant behavior in this benchmark analysis. Because Prandtl (Pr) number of liquid metal is low and the turbulent flow is not developed sufficiently in the conditions of the experiment. (author)
Optimal Population Codes for Space: Grid Cells Outperform Place Cells
Mathis, Alexander; Herz, Andreas V. M.; Stemmler, Martin B
2012-01-01
Rodents use two distinct neuronal coordinate systems to estimate their position: place fields in the hippocampus and grid fields in the entorhinal cortex. Whereas place cells spike at only one particular spatial lo- cation, grid cells fire at multiple sites that correspond to the points of an imaginary hexagonal lattice. We study how to best construct place and grid codes, taking the probabilistic nature of neural spiking into account. Which spatial encoding properties of individu...
Burn up calculations for ETRR 1 and ETRR 2 reactors with wims and origen codes
International Nuclear Information System (INIS)
For ETRR -1 and ETRR - 2 research reactor, the 235 U depletion is determined with wims and origen codes the two calculated results show good agreement with each other. The buildup of different fission products (important from both the safety and protection point of view) is also calculated. The radioactivity and decay heat of the spent fuel is determined up to 30 years
Burnup calculation capability in the PSG2 / Serpent Monte Carlo reactor physics code
International Nuclear Information System (INIS)
The PSG continuous-energy Monte Carlo reactor physics code has been developed at VTT Technical Research Centre of Finland since 2004. The code is mainly intended for group constant generation for coupled reactor simulator calculations and other tasks traditionally handled using deterministic lattices physics codes. The name was recently changed from acronym PSG to 'Serpent', and the capabilities have been extended by implementing built-in burnup calculation routines that enable the code to be used for fuel cycle studies and the modelling of irradiated fuels. This paper presents the methodology used for burnup calculation. Serpent has two fundamentally different options for solving the Bateman depletion equations: 1) the Transmutation Trajectory Analysis method (TTA), based on the analytical solution of linearized depletion chains and 2) the Chebyshev Rational Approximation Method (CRAM), an advanced matrix exponential solution developed at VTT. The first validation results are compared to deterministic CASMO-4E calculations. It is also shown that the overall running time in Monte Carlo burnup calculation can be significantly reduced using specialized calculation techniques, and that the continuous-energy Monte Carlo method is becoming a viable alternative to deterministic assembly burnup codes. (authors)
Manual of Nucost 1.0 - code for calculation of nuclear power generation costs
International Nuclear Information System (INIS)
Nucost is a computer code developed at CDTN to perform cost calculation of electric power generated in PWR nuclear power plants, based on present worth cost method. The Nucost version 1.0 performs calculations of nuclear fuel cost cycle by cycle during the time life of the power plant. That calculation is performed with enough details permitting optimization and minimization. The code is also a tool to aid reload projects and economic operation of PWR reactors. This manual presents a description of Nucost version 1.0, instruction to enter data preparation and description of the Nucost output. (M.I.)
Using the ORIGEN-2 computer code for near core activation calculations
International Nuclear Information System (INIS)
The ORIGEN2 computer code is a useful tool for calculating radionuclide inventories resulting from irradiation of materials in a reactor. It is widely used to calculate activation products in irradiated metals that form the structural portion of fuel assemblies. The code is straightforward for materials within the active fuel region of a reactor core, which are subject to core average conditions. For materials outside the active core, ORIGEN2 cannot be used directly. However, ORIGEN2 can be used with the appropriate methodology to calculate the activation of materials in near core locations. This paper presents the background and a methodology for estimating radionuclide inventories in activated metals in near core locations
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)
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)
The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented
International Nuclear Information System (INIS)
Transporting and processing of radioisotopes and irradiated targets inside hot cells generate a significant contamination. The majority of contamination comes from dispersion of radioactive materials during processing the samples after irradiation. Processing includes opening, extracting the irradiated samples, and preparing the samples in a shield prior to transportation. A model of dispersion of radioactive products inside the cell is postulated. Before decontaminating the cell, the expected dose received by the worker must be evaluated. A RESRAD-BUILD code is used in this study to calculate the dose and the corresponding risk. The calculated dose received during the decontamination process is more than the permissible dose and many proposals are presented in the study to decrease the level of received doses
Code accuracy evaluation of ISP 35 calculations based on NUPEC M-7-1 test
International Nuclear Information System (INIS)
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)
Development of the Joyo MK-II core bowing reactivity calculation code
International Nuclear Information System (INIS)
The study on the passive safety test by using the Experimental Fast Reactor Joyo has been performed to demonstrate the inherent safety of fast breeder reactors. In this study, emphasis was placed on the improvement on the accuracy of the feedback reactivity analysis. As a bowing reactivity might play a significant roll in ATWS analysis because of its effectively short time constant and relatively large magnitude, an emphasis was placed upon the evaluation of the analysis precision of bowing reactivity. Taking into account of the refueling and irradiation history of the individual core component, the core bowing behavior in Joyo has been analyzed by using the MK-II core management code system MAGI, the interface code TETRAS which interpolate neutron flux and coolant temperature at the position of wrapper tube, and the core bowing calculation code BEACON. Calculation accuracy of above mentioned system was evaluated through the comparison of calculated and measured permanent distortion of subassemblies. In 1996, core bowing reactivity was calculated by AURORA code using the above calculated bowing behavior of individual core component as input. But because an approximate two dimensional material reactivity worth map was utilized in AURORA, it was made clear that some amount of error caused by extrapolation could not be neglected. Therefore calculation code ARCHCOM (Analysis of Reactivity Change due to Core Mechanics) which utilize three dimensional material reactivity worth map as input was developed for the Joyo MK-II core bowing reactivity calculation. This code reduces above mentioned extrapolation error that used to be occurred at isolated core component, such as control rod or irradiation rig and at the interface region between fuel and reflector which had sharp bowing reactivity worth gradient. (author)
Methods, algorithms and computer codes for calculation of electron-impact excitation parameters
Bogdanovich, P; Stonys, D
2015-01-01
We describe the computer codes, developed at Vilnius University, for the calculation of electron-impact excitation cross sections, collision strengths, and excitation rates in the plane-wave Born approximation. These codes utilize the multireference atomic wavefunctions which are also adopted to calculate radiative transition parameters of complex many-electron ions. This leads to consistent data sets suitable in plasma modelling codes. Two versions of electron scattering codes are considered in the present work, both of them employing configuration interaction method for inclusion of correlation effects and Breit-Pauli approximation to account for relativistic effects. These versions differ only by one-electron radial orbitals, where the first one employs the non-relativistic numerical radial orbitals, while another version uses the quasirelativistic radial orbitals. The accuracy of produced results is assessed by comparing radiative transition and electron-impact excitation data for neutral hydrogen, helium...
International Nuclear Information System (INIS)
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)
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.
A FORTRAN computer code for calculating flows in multiple-blade-element cascades
Mcfarland, E. R.
1985-01-01
A solution technique has been developed for solving the multiple-blade-element, surface-of-revolution, blade-to-blade flow problem in turbomachinery. The calculation solves approximate flow equations which include the effects of compressibility, radius change, blade-row rotation, and variable stream sheet thickness. An integral equation solution (i.e., panel method) is used to solve the equations. A description of the computer code and computer code input is given in this report.
Off-site dose calculation computer code based on ICRP-60(II) - liquid radioactive effluents -
International Nuclear Information System (INIS)
The development of computer code for calculating off-site doses(K-DOSE60) was based on ICRP-60 and the dose calculationi equations of Reg. Guide 1.109. In this paper, the methodology to compute dose for liquid effluents was described. To examine reliability of the K-DOSE60 code the results obtained from K-DOSE60 were compared with analytic solutions. For liquid effluents. The results by K-DOSE60 are in agreement with analytic solution
POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author)
Development of calculation code of fission products specific activity in primary coolant
International Nuclear Information System (INIS)
Based on an assumption of that there is a design basis fuel defect level from reactor startup, calculation method of fission products specific activities in primary coolant is studied. Time-dependent nuclide activities in defect fuel are calculated by ORIGEN code, and nuclide releases from the defect fuel are considered. After processed by interface codes, data are used by PCFPA code which is used to calculate nuclide activities in the coolant. PCFPA solves differential equations by unit of decay chain, and totally considers decay's contribution to nuclide activities, and considers different system design between secondary and third generation plants such as AP1000. The method could provide the maximum of specific activity during plant operation and their results are consistent with data in AP1000 DCD(Rev.16). The method could be applicable to shielding design in secondary and third generation plants such as AP1000. (authors)
An Efficient Group Key Management Using Code for Key Calculation for Simultaneous Join/Leave: CKCS
Directory of Open Access Journals (Sweden)
Melisa Hajyvahabzadeh
2012-08-01
Full Text Available This paper presents an efficient group key management protocol, CKCS (Code for Key Calculation in Simultaneous join/leave for simultaneous join/leave in secure multicast. This protocol is based on logical key hierarchy. In this protocol, when new members join the group simultaneously, server sends only thegroup key for those new members. Then, current members and new members calculate the necessary keys by node codes and one-way hash function. A node code is a random number which is assigned to each key to help users calculate the necessary keys. Again, at leave, the server just sends the new group key to remaining members. The results show that CKCS reduces computational and communication overhead, and also message size in simultaneous join/leave.
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.
Development of neutral transport lattice code DENT-2D and benchmark calculation
International Nuclear Information System (INIS)
We developed new transport lattice code called DENT-2D (Deterministic Neutral Particle Transport Code in 2-D imensional Space)primarily to generate few- group constants for the reactor physics analysis diffusion codes. This code is designed to be coupled with KAERI reactor analysis nodal code, MASTER [1] ,to complete the design system package. CASMO-3 and HELIOS have been used in generating the few- group constant for MASTER. Currently DENT-2D includes only neutron particle transport calculation in 2-dimensional Cartesian geometry. The characteristics method is adopted for the spatial discretization, which is advantageous for the treatment of the complicated geometry structure and the highly anisotropic scattering. The subgroup method is used for the resonance treatment. B1 approximation has been used to obtain the criticality spectrum considering the leakage effect in the real core situation. The exponential matrix method has been used for the depletion calculation. The results of benchmark calculations show that the prediction capability of DENT-2D is comparable to the other lattice codes such as HELIOS and CASMO-3
The FLUFF code for calculating finned surface heat transfer -description and user's guide
International Nuclear Information System (INIS)
FLUFF is a computer code for calculating heat transfer from finned surfaces by convection and radiation. It can also represent heat transfer by radiation to a partially emitting and absorbing medium within the fin cavity. The FLUFF code is useful not only for studying the behaviour of finned surfaces but also for deriving heat fluxes which can be applied as boundary conditions to other heat transfer codes. In this way models of bodies with finned surfaces may be greatly simplified since the fins need not be explicitly represented. (author)
Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide
International Nuclear Information System (INIS)
The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems
Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors
International Nuclear Information System (INIS)
The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d4 code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)
VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment
International Nuclear Information System (INIS)
Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)
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)
RAP-4A Computer code for thermohydraulic calculation of liquid metal cooled fuel clusters
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RAP-4A is a programme for calculating the fuel clusters thermal-hydraulic parameters in a fast liquid metal-cooled reactor. The code gives the possibility to calculate steady state axial distribution temperature, enthalpy, pressure drop and mass velocity . A monodimensional mathematical model along the cluster allowing the study of the single and two phase flow is used by taking into account the mixing between adjacent subchannels. Physical and mathematical models, general features and an example are presented. RAP-4A code is written in FORTRAN-IV language on IBM 370/135 computer
International Nuclear Information System (INIS)
Four calculational benchmarks have been selected to compare various nuclear data libraries based on both ENDF/B-IV and V, and to compare results from various transport codes. Discrepancies up to 20% in tritium production from 7Li were found and have been attributed mainly to differences in current ENDF/B-IV and V evaluations, while approx.4% is attributed to differences in the group structure of the libraries used. Results from MCNP and VIP Monte Carlo codes are in good agreement, but MORSE calculations show good agreement only for high threshold reactions
ITP.FOR: A code to calculate thermal transients in High Level Waste Tanks
International Nuclear Information System (INIS)
A variety of processing operations for high level radioactive waste occur in the High Level Waste Tanks in the H-Area of the Savannah River Site. Thermal design constraints exist on these processes, principally to limit the amount of corrosion inhibitor which must be added to protect the tank and cooling coil materials. The required amount of corrosion inhibitor, which must subsequently be removed prior to trapping the waste in borosilicate glass, increases exponentially with temperature over a fairly narrow range (some tens of degrees Celsius). For this reason, there is a need to model the thermal-hydraulic processes occurring in the waste tanks. A FORTRAN computer code, called ITP.FOR, was written to provide a simple but reasonably accurate analysis tool for plant operation design. The code was specifically written to model Tank 48, in which the In-Tank Precipitation (ITP) process of precipitating radioactive cesium will be initiated. Although the ITP.FOR code was written as personal-use software for scoping design calculations for Tank 48, the current intent is to extend the code's applicability to other H-Area waste tanks, and to certify the code in accordance with the NRTSC Quality Assurance requirements for critical-use software (1Q-34, 1991). Since the code's capabilities have generated some interest to date, the present report is presented as interim documentation of the code's mathematical models. This documentation will eventually be supplanted by the formal documentation of the expanded and benchmarked code
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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
International Nuclear Information System (INIS)
This report describes a development of a wind field calculation code and an atmospheric dispersion and dose calculation code which can be used for real-time prediction in an emergency. Models used in the computer codes are a mass-consistent model for wind field and a particle diffusion model for atmospheric dispersion. In order to attain quick response even when the codes are used in a small-scale computer, high-speed iteration method (MILUCR) and kernel density method are applied to the wind field model and the atmospheric and dose calculation model, respectively. In this report, numerical models, computational codes, related files and calculation examples are shown. (author)
International Nuclear Information System (INIS)
TEMPUL is one dimensional computer code for calculating radial fuel temperature distribution in a fuel immediately after the pulse. Implementation of TEMPUL code was performed to calculate of radial temperature distribution on TRIGA fuel element. The gap between fuel element and cladding is treated to be in contact (without gap), gap is filled with air and gap is filled with helium gas, respectively. Equilateral triangular arrangement coolant channel is assumed. The calculated results on calculation of radial temperature distribution in TRIGA fuel element immediately after the pulse occur relatively high ascending tendency in zirconium rod (radius 0.3175 cm) and fuel element-cladding interface (radius 1.82245 cm) at the first second after pulse with no gap and gap filled with helium gas treatment. Rising of cladding and interface between cladding and coolant average temperature reach up to 500 oC drastically occur in the first second after the pulse. (author)
PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation
International Nuclear Information System (INIS)
The computer code PEGASUS was developed to calculate neutron-induced reaction cross sections on the basis of the closed form exciton model preequilibrium theory and the multi-step evaporation theory. The cross sections and emitted particle spectra are calculated for the compound elastic scattering, (n,γ), (n,n'), (n,p), (n,α), (n,d), (n,t), (n,3He), (n,2n), (n,n'p), (n,n'α), (n,n'd), (n,n't), (n,2p) and (n,3n) reactions. The double differential cross sections of emitted particles are also calculated. The calculated results are written on a magnetic disk in the ENDF format. Parameter files and/or systematics formulas are provided for level densities, mass excess, radiation widths and inverse cross sections so that the input data to the code are made minimum. (author)
Study of magnetic island using a 3D MHD equilibrium calculation code
International Nuclear Information System (INIS)
Coupling the magnetic diagnostics and a 3D MHD equilibrium calculation code, the magnetic island is studied in the Large Helical Device (LHD) experiment. In an experiment, the collapse in the plasma core was observed in a configuration, which has large magnetic island produced by external perturbation coils. At the collapse, the temperature profile was flattened. This suggests the magnetic island evolved. The magnetic island was observed by the magnetic diagnostics. The magnetic diagnostics also suggests evolving the magnetic island. A 3D MHD equilibrium is calculated by the 3D MHD equilibrium code then signals of the magnetic diagnostics are simulated. Since the comparison of observed and calculated signals is comparable, the magnetic island in calculated equilibrium is similar to one of the experiment. (author)
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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)
Lazarakis, P.; Bug, M. U.; Gargioni, E.; Guatelli, S.; Rabus, H.; Rosenfeld, A. B.
2012-03-01
The concept of nanodosimetry is based on the assumption that initial damage to cells is related to the number of ionizations (the ionization cluster size) directly produced by single particles within, or in the close vicinity of, short segments of DNA. The ionization cluster-size distribution and other nanodosimetric quantities, however, are not directly measurable in biological targets and our current knowledge is mostly based on numerical simulations of particle tracks in water, calculating track structure parameters for nanometric target volumes. The assessment of nanodosimetric quantities derived from particle-track calculations using different Monte Carlo codes plays, therefore, an important role for a more accurate evaluation of the initial damage to cells and, as a consequence, of the biological effectiveness of ionizing radiation. The aim of this work is to assess the differences in the calculated nanodosimetric quantities obtained with Geant4-DNA as compared to those of the ad hoc particle-track Monte Carlo code ‘PTra’ developed at Physikalisch-Technische Bundesanstalt (PTB), Germany. The comparison of the two codes was made for incident electrons of energy in the range between 50 eV and 10 keV, for protons of energy between 300 keV and 10 MeV, and for alpha particles of energy between 1 and 10 MeV as these were the energy ranges available in both codes at the time this investigation was carried out. Good agreement was found for nanodosimetric characteristics of track structure calculated in the high-energy range of each particle type. For lower energies, significant differences were observed, most notably in the estimates of the biological effectiveness. The largest relative differences obtained were over 50%; however, generally the order of magnitude was between 10% and 20%.
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The MGPRAKTINETs computer code for the BESM-6 computer intended for calculation of zone average trmal neutron group fluxes and functionals is described. The neutron spatial-energy distribution in a multizone cyllindrically-symmetric reactor cell is calculated by the operator splitting method. For the solution of the spatial part of the problem the method of surface pseudosources (Gsub(N)-approximation) in approximation of plane derivatives from the energy neutron current is employed. The energy part of the problem is solved in a multigroup approximation. Computer code efficiency has been demonstrated by calculation of two-zone cells with internal and external sources of the cell with on additional absorber and RBMK cell with reduction of the latter to cylindrical geometry. It is shown that the approximation of plane derivatives of neutron energy current allows calculating reactor cell characteristics with a sufficient for design calculations accuracy
Performance of independent dose calculation in helical tomotherapy: implementation of the MCSIM code
International Nuclear Information System (INIS)
Currently, a software-based second check dose calculation for helical tomotherapy (HT) is not available. The goal of this study is to evaluate the dose calculation accuracy of the in-house software using EGS4 /MCSIM Monte Carlo environment against the treatment planning system calculations. In-house software was used to convert HT treatment plan information into a non-helical format. The MCSIM dose calculation code was evaluated by comparing point dose calculations and dose profiles against those from the HT treatment plan. Fifteen patients, representing five treatment sites, were used in this comparison. Point dose calculations between the HT treatment planning system and the EGS4 /MCSIM Monte Carlo environment had percent difference values below 5 % for the majority of this study. Vertical and horizontal planar profiles also had percent difference values below 5 % for the majority of this study. Down sampling was seen to improve speed without much loss of accuracy. EGS4 /MCSIM Monte Carlo environment showed good agreement with point dose measurements, compared to the HT treatment plans. Vertical and horizontal profiles also showed good agreement. Significant time saving may be obtained by down-sampling beam projections. The dose calculation accuracy of the in-house software using the MCSIM code against the treatment planning system calculations was evaluated. By comparing point doses and dose profiles, the EGS4 /MCSIM Monte Carlo environment was seen to provide an accurate independent dose calculation.
International Nuclear Information System (INIS)
Highlights: • Benchmark study performed for the neutronic calculations of TRIGA research reactors. • WIMSD-5B/CITATION is the utilized code system along with the WIMSD-IAEA-69 library. • The studied condensed spectra are five and seven energy groups spectra. • Analyzed: lattice parameters, reactivities, CR worth, flux and power distribution. • The lattice and neutronic parameters showed the accuracy of both condensed spectra. - Abstract: The objective of this paper is to assess the suitability and accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA Mark-III research reactors using the WIMSD/CITATION code system in proposed condensed energy spectra of five and seven energy groups with one and three thermal groups respectively. The utilized cell transport calculations code and core diffusion calculations code are the WIMSD-5B and the CITVAP v3.1 codes respectively, along with the WIMSD-IAEA-69 nuclear data library. Firstly, the assessment goes through analyzing the integral parameters – keff, ρ238, δ235, δ238, and C* – of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark-III Thai research reactor at the condensed energy spectra. The results include the excess reactivities of the cores and the worth of selected control rods which were compared with reference Monte Carlo results and experimental values. The results show good agreement with the references at both energy spectra and the better accuracy are attainable in the five energy groups spectrum. The results also include neutron flux distributions which are evaluated for future comparisons with other calculational techniques even they are comparable to reactors and fuels of the same type. The
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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
TMRBAR: a code to calculate plasma parameters for tandem-mirror reactors operating in the MARS mode
International Nuclear Information System (INIS)
The purpose of this report is to document the plasma power balance model currently used by LLNL to calculate steady state operating points for tandem mirror reactors. The code developed from this model, TMRBAR, has been used to predict the performance and define supplementary heating requirements for drivers used in the Mirror Advanced Reactor Study (MARS) and for the Fusion Power Demonstration (FPD) study. The equations solved included particle and energy balance for central cell and end cell species, quasineutrality at several cardinal points in the end cell region, as well as calculations of volumes, densities and average energies based on given constraints of beta profiles and fusion power output. Alpha particle ash is treated self-consistently, but no other impurity species is treated
Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels
International Nuclear Information System (INIS)
Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)
Some questions of using coding theory and analytical calculation methods on computers
International Nuclear Information System (INIS)
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
A FACSIMILE code for calculating void swelling and creep, with vacancy loops present: version VS4
International Nuclear Information System (INIS)
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)
Energy Technology Data Exchange (ETDEWEB)
Ash, J E; Marciniak, T J
1977-01-01
The Liquid Metal Fast Breeder Reactor (LMFBR) core structure consists of a matrix of hexagonal subassembly ducts. Evaluation of the safety aspects of the core structure requires that reliable computational procedures be available to predict the deformation response of the subassembly configuration to postulated local energy releases. Finite-element computer codes have been developed to calculate deflections and strains of a hexcan subassembly wrapper subjected to internal and external dynamic pressure loadings over a wide range of material-property conditions. An experimental and analytical program has been undertaken to validate and extend the codes for describing the core structural mechanics under reactor operating conditions, including, in particular, descriptions of possible subassembly-to-subassembly damage propagation. This report describes results of the first phase of the experimental program in which single hexcan sections were internally and externally hydrostatically pressurized out-of-pile at room temperature. The experimental data are compared with calculations from a two-dimensional finite-element structural-dynamics code, STRAW. Some additional comparisons were also made with calculations from a three-dimensional code, SADCAT. The correlations obtained between the computations and the hydrostatic experimental results were sufficiently good to validate the STRAW code and proceed to the next phase of the program involving the dynamic structural response.
KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code
International Nuclear Information System (INIS)
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
Erosion corrosion in power plant piping systems - Calculation code for predicting wall thinning
International Nuclear Information System (INIS)
Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the operator of the plant are taken into consideration. Together with a continuously updated erosion corrosion data base the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to minimize material loss due to erosion corrosion, reduce non-destructive testing and curtail monitoring programs for piping systems, recommend life-extending measures. (author). 12 refs, 17 figs
WOLF: a computer code package for the calculation of ion beam trajectories
International Nuclear Information System (INIS)
The WOLF code solves POISSON'S equation within a user-defined problem boundary of arbitrary shape. The code is compatible with ANSI FORTRAN and uses a two-dimensional Cartesian coordinate geometry represented on a triangular lattice. The vacuum electric fields and equipotential lines are calculated for the input problem. The use may then introduce a series of emitters from which particles of different charge-to-mass ratios and initial energies can originate. These non-relativistic particles will then be traced by WOLF through the user-defined region. Effects of ion and electron space charge are included in the calculation. A subprogram PISA forms part of this code and enables optimization of various aspects of the problem. The WOLF package also allows detailed graphics analysis of the computed results to be performed
Computer codes for the calculation of vibrations in machines and structures
International Nuclear Information System (INIS)
After an introductory paper on the typical requirements to be met by vibration calculations, the first two sections of the conference papers present universal as well as specific finite-element codes tailored to solve individual problems. The calculation of dynamic processes increasingly now in addition to the finite elements applies the method of multi-component systems which takes into account rigid bodies or partial structures and linking and joining elements. This method, too, is explained referring to universal computer codes and to special versions. In mechanical engineering, rotary vibrations are a major problem, and under this topic, conference papers exclusively deal with codes that also take into account special effects such as electromechanical coupling, non-linearities in clutches, etc. (orig./HP)
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.
Burnup calculations of TR-2 Research Reactor with Monteburns Monte Carlo Code
International Nuclear Information System (INIS)
Full text: In this study, some neutronic calculations of first and second core cycles of 5 MW pool type TR-2 Research Reactor have been performed using Multi-Step Monte Carlo Burnup Code System MONTEBURNS and the results were compared with the values of experiments and other codes. Time dependent keff distribution and burnup ratios belong to first and second core cycles of TR-2 Research Reactor were compared and quite good consistence in the results were observed. After modeling the first and second core cycles of TR-2 with MCNP5 Monte Carlo code, MCNP5 used in MONTEBURNS code has been parallelized in 8 HP ProLiant BL680C G5 systems with 4 quad-core Intel Xeon E7330 CPU, utilizing the MPI parallel protocol and simulations were performed on the 128 cores Linux parallel computing machine system. The computation time was reduced by parallelization of MONTEBURNS which uses MCNP in many steps. (authors)
CPS: a continuous-point-source computer code for plume dispersion and deposition calculations
Energy Technology Data Exchange (ETDEWEB)
Peterson, K.R.; Crawford, T.V.; Lawson, L.A.
1976-05-21
The continuous-point-source computer code calculates concentrations and surface deposition of radioactive and chemical pollutants at distances from 0.1 to 100 km, assuming a Gaussian plume. The basic input is atmospheric stability category and wind speed, but a number of refinements are also included.
Energy Technology Data Exchange (ETDEWEB)
Shibata, C.S.; Montes, A. [Instituto de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil); Galvao, R.M.O. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica
1994-04-01
This paper describes the `FLINESH` computer code for magnetic fields calculation developed for the simulation of field configurations in plasma magnetic confinement devices. The expressions for the poloidal field and flux, the program structure and the input parameters description are presented, and also the analysis of the graphic output possibilities. (L.C.J.A.). 12 refs, 14 figs, 2 tabs.
Calculation of double differential cross sections for structural materials by PEGASUS code
International Nuclear Information System (INIS)
The neutron induced neutron and proton emission double differential cross sections were calculated with PEGASUS code for Cr, Fe and Ni and their isotopes. Results are in fair agreement with experimental data for neutron energy near 14 MeV, confirming that PEGASUS may be applied successfully to produce the double differential cross section data for JENDL. (author)
International Nuclear Information System (INIS)
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
A 3D coarse-mesh time dependent code for nuclear reactor kinetic calculations
International Nuclear Information System (INIS)
A course-mesh code for time-dependent multigroup neutron diffusion calculation based on a direct integration scheme for the time dependence and a low order nodal flux expansion approximation for the space variables has been implemented as a fast tool for transient analysis. (Author)
Process of cross section generation for radiation shielding calculations, using the NJOY code
International Nuclear Information System (INIS)
The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author)
The solution of the LEU and MOX WWER-1000 calculation benchmark with the CARATE - multicell code
International Nuclear Information System (INIS)
Preparations for disposition of weapons grade plutonium in WWER-1000 reactors are in progress. Benchmark: Defined by the Kurchatov Institute (S. Bychkov, M. Kalugin, A. Lazarenko) to assess the applicability of computer codes for weapons grade MOX assembly calculations. Framework: 'Task force on reactor-based plutonium disposition' of OECD Nuclear Energy Agency. (Authors)
Calculate Some Characteristic Parameters Of VVER-1000's Fuel Assembly By MCNP4C2 Code
International Nuclear Information System (INIS)
This report presents the descriptions of parameters characteristics of the LEU and MOX Fuel Assemblies of VVER-1000 reactor, and calculation results such as infinite neutron multiplication factor kinf, two groups energies constants, neutron flux distribution by using Monte Carlo code MCNP. (author)
Calculation capability of NETFLOW++ code for natural circulation in sodium cooled fast reactor
International Nuclear Information System (INIS)
The present paper describes the simulation of the natural circulation in the secondary heat transport system (HTS) after an intentional plant trip of the experimental fast reactor 'Joyo' with the 140 MWt irradiation core using the plant dynamics analysis code NETFLOW++. This code is an integrated network code to calculate the nuclear steam supply system (NSSS) and the balance of the plant (BOP), i.e., turbine/feedwater system. Up to now, the code has been validated using transient data of the experimental sodium facility PLANDTL, experimental fast reactor 'Joyo' and the prototype fast breeder reactor 'Monju'. These validations are steps to evaluate the natural circulation transient of a large-scale fast breeder reactor. Therefore, the former validation results are introduced to show the degree of agreement. In order to consolidate the applicability of the code to the evaluation of the natural circulation, the present test was selected and simulated using the NETFLOW++ code. Major plant parameters are simulated with good agreement such a similar accuracy as the Mimir-N2 exclusive code for 'Joyo'. As a result, it is concluded that the NETFLOW++ is applicable to the natural circulation analysis of sodium-cooled fast reactors with the similar scale of the prototype reactor 'Monju'. (author)
Calculation of Plutonium content in RSG-GAS spent fuel using IAFUEL computer code
International Nuclear Information System (INIS)
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
2D Resistive Magnetohydrodynamics Calculations with an Arbitrary Lagrange Eulerian Code
Rousculp, C. L.; Gianakon, T. A.; Lipnikov, K. N.; Nelson, E. M.
2015-11-01
Single fluid resistive MHD is useful for modeling Z-pinch configurations in cylindrical geometry. One such example is thin walled liners for shock physics or HEDP experiments driven by capacitor banks such as the LANL's PHELIX or Sandia-Z. MHD is also useful for modeling high-explosive-driven flux compression generators (FCGs) and their high-current switches. The resistive MHD in our arbitrary Lagrange Eulerian (ALE) code operates in one and two dimensions in both Cartesian and cylindrical geometry. It is implemented as a time-step split operator, which consists of, ideal MHD connected to the explicit hydro momentum and energy equations and a second order mimetic discretization solver for implicit solution of the magnetic diffusion equation. In a staggered grid scheme, a single-component of cell-centered magnetic flux is conserved in the Lagrangian frame exactly, while magnetic forces are accumulated at the nodes. Total energy is conserved to round off. Total flux is conserved under the ALE relaxation and remap. The diffusion solver consistently computes Ohmic heating. Both Neumann and Dirichlet boundary conditions are available with coupling to external circuit models. Example calculations will be shown.
IPEN/MB-01 heavy reflector benchmark calculations using Serpent code
International Nuclear Information System (INIS)
A series of critical experiments with water-moderated square-pitched lattices with low-enriched uranium fuel rods was conducted at the IPEN/MB-01 research reactor facility, in 2005. Later, this data become some benchmarks. In one of these experiments the west face of the reactor core was covered with a set of thin SS-304 plates to simulate a heavy reflector as used in the EPR reactor (LEU-COMP-HERM-043). The plates are 3 mm thick and their width and axial length were large enough to cover one whole side of the active core of the reactor. The critical configurations were found as a function of the number of plates. Fuel rods containing UO2 with uranium enriched to 4.3% 235U were arranged in specific geometric configurations to be as close as possible to the critical state. In this work, these benchmark configurations with heavy reflectors were modeled using the Serpent Monte Carlo Code. Serpent uses a universe-based geometry model, which allows the description of practically any three-dimensional fuel or reactor configuration. Neutron transport is based on a combination of surface-to-surface ray-tracing and the Woodcock delta-tracking method. Woodcock method is many times faster than ray-tracing, so compared to MCNP code, Serpent code can bring huge gains in processing time of reactor calculations and reaction rate calculations. The results of these calculations were compared with experimental data and calculations with codes MCNP5 and SCALE6 (KENO-VI) using ENDF/B-VII.0 as cross-section input data. The codes performances are compared in terms of CPU calculation time and agreement with experimental data. Additional y, sensitivity on keff of Serpent woodcock threshold parameter was analyzed. (author)
Decay heat calculations with the CEA radioactivity data bauk and the code PEPIN
International Nuclear Information System (INIS)
The CEA radioactivity data bank, has been updated mainly from ENSDF and from some recent experimental results. This library contains the decay data for about 700 fission products (F.P.), 220 actinides and more than 1400 other nuclides. A comparison between our data and ENDF/B5 is shown for the fission products. The fission products part of this library is currently used for shielding and decay heat calculations with the PEPIN code. Calculations and spectral comparisons of the available experiments (Dickens, Lott, Yarmell ...) and other recent calculations is made for thermal fission of 235U and 239Pu using our data bank as input
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Two methods of calculating criticality are available in the 3D generalised geometry Monte Carlo particle transport code SPARTAN (Bending and Heffer, 1975). The first is a matrix technique in which the multiplication constant and source distribution of the system under study are calculated from estimates of fission probabilities and the second a method in which the multiplication constant is inferred from estimates of changes in neutron population over a number of neutron generations. Modifications are described which have been made to the way in which these methods are used in SPARTAN in order to improve the efficiency of criticality calculations. (author)
International Nuclear Information System (INIS)
PEGASUS, a preequilibrium and evaporation theory code, was developed which calculates 17 neutron reaction cross sections, the particle spectra and the double differential cross sections. The code is suited to a rapid and scoping calculation. Theoretical model and the some results of calculation are presented. (author)
International Nuclear Information System (INIS)
Highlights: • We present a new Monte Carlo method to perform sensitivity/perturbation calculations. • Sensitivity of keff, reaction rates, point kinetics parameters to nuclear data. • Fully continuous implicitly constrained Monte Carlo sensitivities to scattering distributions. • Implementation of the method in the continuous energy Monte Carlo code SERPENT. • Verification against ERANOS and TSUNAMI generalized perturbation theory results. - Abstract: In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the effects of nuclear data perturbation on several response functions: the effective multiplication factor, reaction rate ratios and bilinear ratios (e.g., effective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators
YNOGK: A New Public Code for Calculating Null Geodesics in the Kerr Spacetime
Yang, Xiaolin; Wang, Jiancheng
2013-07-01
Following the work of Dexter & Agol, we present a new public code for the fast calculation of null geodesics in the Kerr spacetime. Using Weierstrass's and Jacobi's elliptic functions, we express all coordinates and affine parameters as analytical and numerical functions of a parameter p, which is an integral value along the geodesic. This is the main difference between our code and previous similar ones. The advantage of this treatment is that the information about the turning points does not need to be specified in advance by the user, and many applications such as imaging, the calculation of line profiles, and the observer-emitter problem, become root-finding problems. All elliptic integrations are computed by Carlson's elliptic integral method as in Dexter & Agol, which guarantees the fast computational speed of our code. The formulae to compute the constants of motion given by Cunningham & Bardeen have been extended, which allow one to readily handle situations in which the emitter or the observer has an arbitrary distance from, and motion state with respect to, the central compact object. The validation of the code has been extensively tested through applications to toy problems from the literature. The source FORTRAN code is freely available for download on our Web site http://www1.ynao.ac.cn/~yangxl/yxl.html.
Blind Calculation of RD-14M Small Break LOCA Tests by CATHENA Code
International Nuclear Information System (INIS)
KAERI participated with the computer code CATHENA, which is used to analyze Pressurized Heavy Water Reactors (PHWRs), in an IAEA International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate thermal-hydraulic computer code against qualified data for Small Break Loss of Coolant Accident (SBLOCA) scenario generated on RD-14M Test Facility. Two specific SBLOCA tests selected for this ICSP titled 'Comparison of HWR Code Predictions with SBLOCA Experimental Data', are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow down. This report presents the blind calculation results for these tests conducted by CATHENA code before the test data are distributed to participants. For B9006 test, CATHENA code simulated all the phases of the transient such as blowdown, high-pressure ECI, secondary pressure ramp, refill, switch from high pressure ECI to low pressure ECI, exponential pump ramp, and natural circulation. For B9802 test, CATHENA calculation was intended to predict temperature rise of the FES sheath due to channel boiling, and power supply trip on high FES sheath temperature (600 .deg. C) process protection trip
International Nuclear Information System (INIS)
The computer codes BROHR and SYSFIT are presented. Both codes are based on the first-order matrix formalism of ion optics. By means of the code BROHR the trajectories of ions and electrons inside of any inclined field accelerating tubes can be calculated. The influence of the stripping process at tandem accelerators is included by changing of the mass and the charge of the ions and by increasing the beam emittance. The code SYSFIT is used for calculation of any beam transport systems and of the transported beam. Special requested imaging properties can be realized by parameter variation. Calculated examples are given for both codes. (author)
International Nuclear Information System (INIS)
Calculation of personnel exposure is a one of the main parameters being evaluated within the pre-decommissioning plans together with other decommissioning drivers such as costs, manpower, amounts of RAW and conventional waste and amount of discharged gaseous and liquid effluents. Alongside with manpower, the exposure is an indicator of the decommissioning process for need of staff, and quantifies impact of decommissioning on personnel from the radio hygienic point of view. At the same time it indicates suitability of individual work procedures use for decommissioning activities. For this reason it is important to estimate as precise as possible demands on personnel exposure even during preparatory decommissioning phase to quantify impact of decommissioning on personnel and eventually optimize the decommissioning process, if needed. The most appropriate way of staff exposure estimation during decommissioning preparatory phases is its calculation based on radiological and physical characteristics of equipment to be decommissioned and also quantitative and qualitative characterisation of typical decommissioning activities. On one hand, the methodology of exposure calculation should allow as much as possible realistic description and algorithmisation of exposure ways during decommissioning activities. On the other hand the calculation have to be systematic, well-arranged and clearly definable by appropriate mathematic relations. Calculation can be made by various approaches using more or less sophisticated software solutions from classic MS Excel sheets up to the complex calculation codes. In this paper, a methodology used for personnel exposure calculation and optimization implemented within the complex computer code OMEGA developed at DECOM, a.s. is described. (author)
Stepšys, A.; Mickevicius, S.; Germanas, D.; Kalinauskas, R. K.
2014-11-01
This new version of the HOTB program for calculation of the three and four particle harmonic oscillator transformation brackets provides some enhancements and corrections to the earlier version (Germanas et al., 2010) [1]. In particular, new version allows calculations of harmonic oscillator transformation brackets be performed in parallel using MPI parallel communication standard. Moreover, higher precision of intermediate calculations using GNU Quadruple Precision and arbitrary precision library FMLib [2] is done. A package of Fortran code is presented. Calculation time of large matrices can be significantly reduced using effective parallel code. Use of Higher Precision methods in intermediate calculations increases the stability of algorithms and extends the validity of used algorithms for larger input values. Catalogue identifier: AEFQ_v4_0 Program summary URL: http://cpc.cs.qub.ac.uk/summaries/AEFQ_v4_0.html Program obtainable from: CPC Program Library, Queen’s University of Belfast, N. Ireland Licensing provisions: GNU General Public License, version 3 Number of lines in programs, including test data, etc.: 1711 Number of bytes in distributed programs, including test data, etc.: 11667 Distribution format: tar.gz Program language used: FORTRAN 90 with MPI extensions for parallelism Computer: Any computer with FORTRAN 90 compiler Operating system: Windows, Linux, FreeBSD, True64 Unix Has the code been vectorized of parallelized?: Yes, parallelism using MPI extensions. Number of CPUs used: up to 999 RAM(per CPU core): Depending on allocated binomial and trinomial matrices and use of precision; at least 500 MB Catalogue identifier of previous version: AEFQ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 181, Issue 2, (2010) 420-425 Does the new version supersede the previous version? Yes Nature of problem: Calculation of matrices of three-particle harmonic oscillator brackets (3HOB) and four-particle harmonic oscillator brackets (4HOB) in a more
Development of an effective delayed neutron fraction calculation code, BETA-K
Energy Technology Data Exchange (ETDEWEB)
Kim, Taek Kyum; Song, Hoon; Kim, Young Il; Kim, Young In; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-08-01
BETA-K, an effective delayed neutron fraction calculation code consistent with Nodal Expansion Method (NEM), has been developed. By using relevant output files of DIF3D code, it can calculate the effective delayed neutron fraction({beta}{sub eff}), neutron lifetime(l{sub eff}), fission spectrum ({chi}-bar) and fission yield data({nu}) for each fissionable isotope, composition of fuels and over the whole core. BETA-K code has been validated by comparing the calculated values to the measured ones of effective delayed neutron fraction in two critical experiments, BFS73-1 and BFS55-1. BFS73-1 is a metal uranium core and BFS55-1 is a metal plutonium core. The C/E values, 1.007 and 0.992 for BFS73-1 and BFS55-1 respectively, agreed well with the experimental values within the experiment errors. BETA-K code predicts 0.00709 and 0.356 {mu}sec as the effective delayed neutron fraction and neutron life time for the uranium metallic fueled equilibrium core of 150MWe KALIMER. (author). 9 refs., 6 figs., 12 tabs.
Linear calculations of edge current driven kink modes with BOUT++ code
International Nuclear Information System (INIS)
This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density
Development of an effective delayed neutron fraction calculation code for hexagonal core
International Nuclear Information System (INIS)
BETA-K, an effective delayed neutron fraction calculation code consistent with Nodal Expansion Method(NEM) of hexagonal geometric core, has been developed. By using relevant output files of DIF3D code, it can calculate the effective delayed neutron fraction(betaeff) and neutron lifetime(leff) for each fissionable isotope, composition of fuels and over the whole core. BETA-K code has been validated by comparing the calculated values to the measured ones of effective delayed neutron fraction in two critical experiments, BFS73-1 and BFS55-1. BFS73-1 is a metal uranium core and BFS55-1 is a metal plutonium core. The C/E values, 1.007 and 0.992 for BFS73-1 and BFS55-1 respectively, agreed well with the experimental values within the experiment errors. BETA-K code predicts 0.00709 and 0.356μ sec as the effective delayed neutron fraction and neutron life time for the uranium metallic fueled equilibrium core of 150MWe KALIMER
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.
WIMS-ANL 4.0, Deterministic Code System for Lattice Calculation
International Nuclear Information System (INIS)
1 - Description of program or function: The WIMS-ANL code is an extension of the Winfrith WIMS-D4 code for lattice cell computations. This code has been tailored to address some of the problem areas encountered in dealing with research reactor fuels, experiment, reflector and control regions. The SUPERCELL option eliminates some of the limitations of the traditional SPECTROX solution and supports the solution of more complex geometries with a more detailed spatial mesh and multiple resonance materials. The code generates both macroscopic and microscopic cross sections in the ISOTXS format with any selected number of energy groups. The user can specify which fission product isotopes are to be explicitly included in the microscopic burnup dependent ISOTXS library. Fission product library data can be generated for use with the MCNP code and burnup dependent applications. The cross section library data provided are based on ENDF/B version VI (69 group) and V (69 and 172 group) data. A revised 172 group library based on ENDF/B-VI is being generated with newer data and additional isotopes. This library will be made available at a later time. The code is variably dimensioned so that other group structures could be used. The source code and output format have been completely revised to reflect current coding practices and to permit display of the results on typical desk top monitors. The content of the output displayed is completely under the user's control. 2 - Methods:The methods of solution in WIMS-ANL remain unchanged from those used in the original WIMS-D4 code with the same resonance treatment and a choice of collision probability and DSN solutions for the simple lattice cell. The SUPERCELL option provides for the selection of supporting auxiliary cells that might represent the various different elements and varying spectra of the final SUPERCELL model. The resonance treatments where applicable are carried out in the auxiliary cells. These data are combined in the
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.
International Nuclear Information System (INIS)
The paper presents the computer code Mitra (Multicomponent isotope transport) which has been constructed to calculate the release of radioactive fission products from nuclear fuels under non-stationary conditions. The code is based on a new integration method fo the mass transport equation in the presence of precipitation, re-solution and radioactive decay. The starting equations and the assumed physical models are briefly described in the main part of the report. A very detailed description of the formulae used and of the Mitra subprograms are presented in extended appendices
The spectral code Apollo2: from lattice to 2D core calculations
International Nuclear Information System (INIS)
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
FLAME3: a three-dimensional nodal code for calculating core reactivity and power distributions
International Nuclear Information System (INIS)
The FLAME3 nodal code calculates core reactivity and three-dimensional power distributions with thermal-hydraulic feedback effects. It employs variable dimensioning, which permits the user to size his own problem subject to the total core storage of the computer. Lagrange interpolation is used for fitting variable data. This allows any input-dependent variable to be fit versus as many as three independent core variables. A transient xenon capability is included, which enhances the code's usefulness in performing maneuvering analyses. Control rod data are input by node, permitting the treatment of partial-length control rods. The various models, including the programmed equation, are described
Comparison between CAREB code calculations and LOCA test results in the FUMEX III project
Energy Technology Data Exchange (ETDEWEB)
Horhoianu, Grigore; Ionescu, Dragos Victor; Pauna, Eduard Ionut [Institute for Nuclear Research, Pitesti (Romania)
2011-05-15
The IAEA initiated a Coordinated Research Project (CRP) on improvement of computer codes used for fuel behaviour simulation under the name: FUMEX III. The Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes. Recently, both codes have been improved with new models in order to extend their capabilities. The behaviour of fuel elements during high-temperature transients like LOCA is of importance to safety and licensing of power reactors. CAREB was developed for fuel transients analyses, such as LOCA and RIA. In this paper a comparison between CAREB code calculations and measured data from FIO-131 LOCA tests is presented. Several parameters were considered, including fuel sheath strains, internal element gas pressure, fuel centerline and sheath temperature, thicknesses of ZrO{sub 2} on the sheath. Fuel behavior during high-temperature transient was reasonably well modeled by CAREB code. New LOCA tests are planed to be performed in the C2-LOCA facility of the TRIGA research reactor at INR Pitesti in order to extend the experimental data base used for transient code validation. (orig.)
Comparison between CAREB code calculations and LOCA test results in the FUMEX III project
International Nuclear Information System (INIS)
The IAEA initiated a Coordinated Research Project (CRP) on improvement of computer codes used for fuel behaviour simulation under the name: FUMEX III. The Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes. Recently, both codes have been improved with new models in order to extend their capabilities. The behaviour of fuel elements during high-temperature transients like LOCA is of importance to safety and licensing of power reactors. CAREB was developed for fuel transients analyses, such as LOCA and RIA. In this paper a comparison between CAREB code calculations and measured data from FIO-131 LOCA tests is presented. Several parameters were considered, including fuel sheath strains, internal element gas pressure, fuel centerline and sheath temperature, thicknesses of ZrO2 on the sheath. Fuel behavior during high-temperature transient was reasonably well modeled by CAREB code. New LOCA tests are planed to be performed in the C2-LOCA facility of the TRIGA research reactor at INR Pitesti in order to extend the experimental data base used for transient code validation. (orig.)
Calculation of fuel and moderator temperature coefficients in APR1400 nuclear reactor by MVP code
International Nuclear Information System (INIS)
In this project, these fuel and moderator temperature coefficients were calculated in APR1400 nuclear reactor by MVP code. APR1400 is an advanced water pressurized reactor, that was researched and developed by Korea Experts, its electric power is 1400 MW. The neutronics calculations of full core is very important to analysis and assess a reactor. Results of these calculation is input data for thermal-hydraulics calculations, such as fuel and moderator temperature coefficients. These factors describe the self-safety characteristics of nuclear reactor. After obtaining these reactivity parameters, they were used to re-run the thermal hydraulics calculations in LOCA and RIA accidents. These thermal-hydraulics results were used to analysis effects of reactor physics parameters to thermal hydraulics situation in nuclear reactors. (author)
Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system
International Nuclear Information System (INIS)
The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)
SYSMOD: user-interface for data processing, calculation codes and analysis of PWR lattices
International Nuclear Information System (INIS)
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)
The conversion coefficients, H'(d,α)/φ, for monoenergetic positrons and positron-emitting radionuclides were calculated by using the user code UCICRPM of the Monte Carlo code EGS5 to estimate the radiation dose for medical staff involved in positron emission tomography examinations. From these coefficients, the dose equivalent rates per unit activity at 0.07 and 10 mm depths in a soft tissue for a straight-line source of 2-deoxy-2-[18F]fluoro-d-glucose (18F-FDG) were calculated by using the developed user code UCF18DOSE. The dose equivalent rates per unit activity at 0.07 and 10 mm depths were measured by using a personal dosemeter (DOSE 3) under the same conditions as those considered in the calculation. The calculated dose equivalent rates per unit activity at 0.07 and 10 mm depths were 0.116 and 0.0352 pSv min-1 Bq-1, respectively, at 20 cm from the 18F-FDG injection tube. (authors)
Calculation of Gamma-ray Responses for HPGe Detectors with TRIPOLI-4 Monte Carlo Code
Lee, Yi-Kang; Garg, Ruchi
2014-06-01
The gamma-ray response calculation of HPGe (High Purity Germanium) detector is one of the most important topics of the Monte Carlo transport codes for nuclear instrumentation applications. In this study the new options of TRIPOLI-4 Monte Carlo transport code for gamma-ray spectrometry were investigated. Recent improvements include the gamma-rays modeling of the electron-position annihilation, the low energy electron transport modeling, and the low energy characteristic X-ray production. The impact of these improvements on the detector efficiency of the gamma-ray spectrometry calculations was verified. Four models of HPGe detectors and sample sources were studied. The germanium crystal, the dead layer of the crystal, the central hole, the beryllium window, and the metal housing are the essential parts in detector modeling. A point source, a disc source, and a cylindrical extended source containing a liquid radioactive solution were used to study the TRIPOLI-4 calculations for the gamma-ray energy deposition and the gamma-ray self-shielding. The calculations of full-energy-peak and total detector efficiencies for different sample-detector geometries were performed. Using TRIPOLI-4 code, different gamma-ray energies were applied in order to establish the efficiency curves of the HPGe gamma-ray detectors.
Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh
International Nuclear Information System (INIS)
The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)
Emergency Doses (ED) - Revision 3: A calculator code for environmental dose computations
International Nuclear Information System (INIS)
The calculator program ED (Emergency Doses) was developed from several HP-41CV calculator programs documented in the report Seven Health Physics Calculator Programs for the HP-41CV, RHO-HS-ST-5P (Rittman 1984). The program was developed to enable estimates of offsite impacts more rapidly and reliably than was possible with the software available for emergency response at that time. The ED - Revision 3, documented in this report, revises the inhalation dose model to match that of ICRP 30, and adds the simple estimates for air concentration downwind from a chemical release. In addition, the method for calculating the Pasquill dispersion parameters was revised to match the GENII code within the limitations of a hand-held calculator (e.g., plume rise and building wake effects are not included). The summary report generator for printed output, which had been present in the code from the original version, was eliminated in Revision 3 to make room for the dispersion model, the chemical release portion, and the methods of looping back to an input menu until there is no further no change. This program runs on the Hewlett-Packard programmable calculators known as the HP-41CV and the HP-41CX. The documentation for ED - Revision 3 includes a guide for users, sample problems, detailed verification tests and results, model descriptions, code description (with program listing), and independent peer review. This software is intended to be used by individuals with some training in the use of air transport models. There are some user inputs that require intelligent application of the model to the actual conditions of the accident. The results calculated using ED - Revision 3 are only correct to the extent allowed by the mathematical models. 9 refs., 36 tabs
Wall-touching kink mode calculations with the M3D code
International Nuclear Information System (INIS)
This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicated by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall
Calculation of the RSG-GAS core using computer code citation-3D
International Nuclear Information System (INIS)
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 Keff are less and greater than unity (Keffeff>1)
GOBLIN computer code. Comparison between calculations and TLTA small break test
International Nuclear Information System (INIS)
GOBLIN calcuations have been performed for two simulation tests of the boiling water reactor (BWR) small break loss-of-coolant accidents (LOCAs) which were conducted in the two loop test apparatus (TLTA). The first test investigated the small break with nondegraded emergency core coolant (ECC) systems and the second test studied the same small break but with degraded ECC systems in which the high pressure core spray (HPCS) was assumed unavailable. Very good agreement between test data and calculations is achieved. The second test is the most challenging from code comparison point of view and the code prediction of the complicated mass distribution pattern which changes with time is very satisfactory. In the first test and to some extent late in the second test multidimensional subchannel effects are evident in the core bundle region. These are not and cannot be reproduced by the code since the bundle model of GOBLIN is strictly one-dimensional. (Author)
Wall-touching kink mode calculations with the M3D code
Energy Technology Data Exchange (ETDEWEB)
Breslau, J. A., E-mail: jbreslau@pppl.gov; Bhattacharjee, A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08542 (United States)
2015-06-15
This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicated by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.
VizieR Online Data Catalog: ynogkm: code for calculating time-like geodesics (Yang+, 2014)
Yang, X.-L.; Wang, J.-C.
2013-11-01
Here we present the source file for a new public code named ynogkm, aim on calculating the time-like geodesics in a Kerr-Newmann spacetime fast. In the code the four Boyer-Lindquis coordinates and proper time are expressed as functions of a parameter p semi-analytically, i.e., r(p), μ(p), φ(p), t(p), and σ(p), by using the Weiers- trass' and Jacobi's elliptic functions and integrals. All of the ellip- tic integrals are computed by Carlson's elliptic integral method, which guarantees the fast speed of the code.The source Fortran file ynogkm.f90 contains three modules: constants, rootfind, ellfunction, and blcoordinates. (3 data files).
International Nuclear Information System (INIS)
The radioactivity induced in PWR cluster control rods during reactor operation has been calculated using the computer programme ORIGEN. Neutron fluxes and spectrum conditions as well as the strongly shielded cross sections for the absorber materials Ag, In and Cd have been obtained by running the cell and assembly code CASMO for a couple of typical cases. The results show that Ag-110m, Fe-55 and Co-60 give the largest activity contributions in the interval 1-10 years after the end of irradiation, and Ni-63 and Cd-113m in a longer time perspective. (author)
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...
Shielding evaluation for e-Linac - Inter-comparison of Monte Carlo codes and analytical calculations
International Nuclear Information System (INIS)
Estimation of optimum shielding thickness is an important aspect in radiation protection as well as in assessment of cost effectiveness of any upcoming accelerator facility. Analytical calculations for shielding estimates are fast and being frequently used even though they are very approximate. Estimates by Monte Carlo codes, on the other hand is accurate, provided used in a judicious manner, but they are very time consuming and require high end computational hardware. The purpose of this work is to compare the results from various available Monte Carlo codes, such as FLUKA and EGSmc. The estimated output was also compared with the analytical techniques. For the work, an e-Linac facility of 50 MeV electron beam was used and calculations were carried out with 1 mA beam current. (author)
CHARADE: A characteristic code for calculating rate-dependent shock-wave response
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.N.; Tonks, D.L.
1991-01-01
In this report we apply spatially one-dimensional methods and simple shock-tracking techniques to the solution of rate-dependent material response under flat-plate-impact conditions. This method of solution eliminates potential confusion of material dissipation with artificial dissipative effects inherent in finite-difference codes, and thus lends itself to accurate calculation of elastic-plastic deformation, shock-to-detonation transition in solid explosives, and shock-induced structural phase transformation. Equations are presented for rate-dependent thermoelastic-plastic deformation for (100) planar shock-wave propagation in materials of cubic symmetry (or higher). Specific numerical calculations are presented for polycrystalline copper using the mechanical threshold stress model of Follansbee and Kocks with transition to dislocation drag. A listing of the CHARADE (for characteristic rate dependence) code and sample input deck are given. 26 refs., 11 figs.
Comprehensive nuclear model calculations: Introduction to the theory and use of the GNASH code
International Nuclear Information System (INIS)
A user's manual describing the theory and operation of the GNASH nuclear reaction computer code is presented. This work is based on a series of lectures describing the statistical Hauser-Feshbach plus preequilibrium version of the code with full angular momentum conservation. This version is expected to be most applicable for incident particle energies between 1 key and 50 MeV. General features of the code, the nuclear models that are utilized, input parameters needed to perform calculations, and the output quantities from typical problems are described in detail. The computational structure of the code and the subroutines and functions that are called are summarized as well. Two detailed examples are considered: 14-MeV neutrons incident on 93Nb and 12-MeV neutrons incident on 238U. The former example illustrates a typical calculation aimed at determining neutron, proton, and alpha emission spectra from 14-MeV reactions, and the latter example demonstrates use of the fission model in GNASH
VVER 1000 SBO calculations with pressuriser relief valve stuck open with ASTEC computer code
International Nuclear Information System (INIS)
Highlights: ► We modelled the ASTEC input file for accident scenario (SBO) and focused analyses on the behaviour of core degradation. ► We assumed opening and stuck-open of pressurizer relief valve during performance of SBO scenario. ► ASTEC v1.3.2 has been used as a reference code for the comparison study with the new version of ASTEC code. - Abstract: The objective of this paper is to present the results obtained from performing the calculations with ASTEC computer code for the Source Term evaluation for specific severe accident transient. The calculations have been performed with the new version of ASTEC. The ASTEC V2 code version is released by the French IRSN (Institut de Radioprotection at de surete nucleaire) and Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), Germany. This investigation has been performed in the framework of the SARNET2 project (under the Euratom 7th framework program) by Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Science (INRNE-BAS).
International Nuclear Information System (INIS)
Experiments were done on several aerosols in air atmospheres at varying temperatures and humidity conditions of interest in forming a data base for testing aerosol behavior models used as part of the process of evaluating the ''source term'' in light water reactor accidents. This paper deals with the problems of predicting the observed experimental data for suspended aerosol concentration with aerosol calculational codes. Comparisons of measured versus predicted data are provided
Efficient Calculations with Multisite Local Orbitals in a Large-Scale DFT Code CONQUEST
Nakata, A; Bowler, D. R.; Miyazaki, T.
2014-01-01
Multisite local orbitals, which are formed from linear combinations of pseudoatomic orbitals from a target atom and its neighbor atoms, have been introduced in the large-scale density functional theory calculation code CONQUEST. Multisite local orbitals correspond to local molecular orbitals so that the number of required local orbitals can be minimal. The multisite support functions are determined by using the localized filter diagonalization (LFD) method [ Phys. Rev. B 2009 , 80 , 205104 ]....
Opacity calculation for target physics using the ABAKO/RAPCAL code
Mínguez Torres, Emilio; Florido, Ricardo; Rodríguez, Rafael; Gil, J.M.; Garcia Rubiano, Jesus; Mendoza, M. A.; Suarez, D; Martel, Pablo
2010-01-01
Radiative properties of hot dense plasmas remain a subject of current interest since they play an important role in inertial confinement fusion (ICF) research, as well as in studies on stellar physics. In particular, the understanding of ICF plasmas requires emissivities and opacities for both hydro-simulations and diagnostics. Nevertheless, the accurate calculation of these properties is still an open question and continuous efforts are being made to develop new models and numerical codes th...
INTRACOIN level 1 benchmark calculations with EIR codes CONZRA, RANCH and RANCHN
International Nuclear Information System (INIS)
The authors present the results from calculations of INTRACOIN level 1, case 1 and 2 (one-dimensional advection-dispersion) benchmarks. The codes used are CONZRA and RANCH, corresponding to a semi-analytical solution of the transport equation, and RANCHN based on a fully numerical solution in the framework of the pseudo-spectral method. The influence of various boundary conditions is investigated. Excellent agreement between results from the different solution approaches is obtained. (Auth.)
Benchmark Calculation for the VHTR 2-D Core by Using the DeCART Code
Energy Technology Data Exchange (ETDEWEB)
Cho, Jin-Young; Kim, Kang-Seog; Lee, Chung-Chan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2006-07-01
Recently, a hexagonal module has been equipped to the DeCART (Deterministic Core Analysis based on Ray Tracing) whole core code for a hexagonal core analysis. The equipment includes a ray tracing module to solve the 2-D whole-core transport problem and a multi-group CMFD module to perform an efficient transport calculation. In this paper, the capability of the DeCART hexagonal module is examined by solving VHTR core problems.
Benchmark Calculation for the VHTR 2-D Core by Using the DeCART Code
International Nuclear Information System (INIS)
Recently, a hexagonal module has been equipped to the DeCART (Deterministic Core Analysis based on Ray Tracing) whole core code for a hexagonal core analysis. The equipment includes a ray tracing module to solve the 2-D whole-core transport problem and a multi-group CMFD module to perform an efficient transport calculation. In this paper, the capability of the DeCART hexagonal module is examined by solving VHTR core problems
Permanent boiling in rod bundles: calculations with the FLICA II B code
International Nuclear Information System (INIS)
Some calculations have been made with the FLICA II B code using a model representing rod-bundles by means of few interconnected channels. The result obtained give some idea of the similarities and differences of behavior to be expected between single channels and rod bundles during permanent forced convection sodium boiling regimes. A new phenomenon, the so-called internal flow excursion is described
Evaluation of Monte Carlo Codes Regarding the Calculated Detector Response Function in NDP Method
International Nuclear Information System (INIS)
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 (10B, 3He, 6Li, 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
MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA
International Nuclear Information System (INIS)
MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations. A CD-ROM is attached as an appendix. (J.P.N.)
Calculation of effective delayed neutron fraction with modified library of Monte Carlo code
International Nuclear Information System (INIS)
Highlights: ► We propose a new Monte Carlo method to calculate the effective delayed neutron fraction by changing the library. ► We study the stability of our method. When the particles and cycles are sufficiently great, the stability is very good. ► The final result is determined to make the deviation least. ► We verify our method on several benchmarks, and the results are very good. - Abstract: A new Monte Carlo method is proposed to calculate the effective delayed neutron fraction βeff. Based on perturbation theory, βeff is calculated with modified library of Monte Carlo code. To verify the proposed method, calculations are performed on several benchmarks. The error of the method is analyzed and the way to reduce error is proposed. The results are in good agreement with the reference data
Calculation of age-dependent effective doses for external exposure using the MCNP code
Energy Technology Data Exchange (ETDEWEB)
Hung, Tran Van [Research and Development Center for Radiation Technology, ThuDuc, HoChiMinh City (VT)
2013-07-15
Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV. (orig.)
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)
BETHSY 6.2TC test calculation with TRACE and RELAP5 computer code
International Nuclear Information System (INIS)
The TRACE code is still under development and it will have all capabilities of RELAP5. The purpose of the present study was therefore to assess the accuracy of the TRACE calculation of BETHSY 6.2TC test, which is 15.24 cm equivalent diameter horizontal cold leg break. For calculations the TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used. The overall results obtained with TRACE were similar to the results obtained by RELAP5/MOD3.3. The results show that the discrepancies were reasonable. (author)
International Nuclear Information System (INIS)
For safety evaluation of nuclear fuel facilities, a nuclear decay data library named JDDL and a computer code COMRAD have been developed to calculate isotopic composition of each nuclide, radiation source intensity, energy spectrum of γ-ray and neutron, and decay heat of spent fuel. JDDL has been produced mainly from the evaluated nuclear data file ENSDF to use new nuclear data. To supplement the data file for short life nuclides, the JNDC data set were also used which had been evaluated by Japan Nuclear Data Committee. Using these data, calculations became possible from short period to long period after irradiation. (author)
Radiation damage calculation by NPRIM computer code with JENDL3.3
International Nuclear Information System (INIS)
The Neutron Damage Evaluation Group of the Atomic Energy Society of Japan starts an identification of neutron-induced radiation damage in materials for typical neutron fields. For this study, a computer code, NPRIM, has been developed to be free from a tedious computational effort, which has been devoted to the calculation of derived quantities such as dpa and helium production rate. Neutron cross sections concerning to damage reactions based on JENDL3.3 are given with 640-group-structure. The impact of cross sections based on JENDL3.3 to damage calculation results has been described in this paper. (author)
Comparative calculations on selected two-phase flow phenomena using major PWR system codes
International Nuclear Information System (INIS)
In 1988 a comparative study on important features and models in six major best estimate thermal hydraulic codes for PWR systems was implemented (Comparison of thermal hydraulic safety codes for PWR Graham, Trotman, London, EUR 11522). It was a limitation of that study that the source codes themselves were not available but the comparison had to be based on the available documentation. In the present study, the source codes were available and the capability of four system codes to predict complex two-phase flow phenomena has been assessed. Two areas of investigation were selected: (a) pressurized spray phenomena; (b) boil-up phenomena in rod bundles. As regards the first area, experimental data obtained in 1972 on the Neptunus Facility (Delft University of Technology) were compared with the results of the calculations using Athlet, Cathare, Relap 5 and TRAC-PT1 and, concerning the second area, the results of two experimental facilities obtained in 1980 and 1985 on Thetis (UKEA) and Pericles (CEA-Grenoble) were considered
A group of neutronics calculations in the MNSR using the MCNP-4C code
International Nuclear Information System (INIS)
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 (UAl4-Al) to 19.75 % LEU fuel (UO2). 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 ( ιp l and βeff). (authors)
Accuracy evaluation of pin exposure calculations in current LWR core design codes
International Nuclear Information System (INIS)
The deregulated utility environment and better utilization of fuel assemblies in nuclear power plants has allowed designers to burn fuel assemblies to maximum allowable exposures. Any uncertainties, associated with the technical approach and numerical methods used to perform pin exposure calculations may cause either peak power exposure to exceed the Nuclear Regulatory Commission (NRC) exposure limit or lead to excessive conservatism and thus inefficient fuel utilization. In this work, a Monte Carlo based coupled depletion code (MCNP5/ORIGEN-S) is utilized to provide reference solutions in order to assess the accuracy of pin power and pin exposure reconstruction methods in the current commercial and licensed three-dimensional (3D) nodal Light Water Reactor (LWR) core design codes. The developed at Pennsylvania State University (PSU) MCNP5/ORIGEN-S coupled depletion code system was validated using measured data from the PSU TRIGA research reactor critical experiments. A number of test cases (modeling benchmarks) representative of LWRs were developed starting from the least complex model towards more complicated and more realistic models. The accuracy evaluation of the pin reconstruction methods was performed by using the CASMO-4 and SIMULATE-3 codes as the representative of current commercial LWR core design systems. Two-dimensional (2D) transport calculations with the TRITON module from the SCALE5 package were employed to produce the spectrum averaged cross-section libraries as a function of burnup for ORIGEN-S calculations. The burnup dependent cross-section libraries are specifically generated for each lattice configuration type. For the MCNP5 calculations continuous cross-section libraries for different isotopes at hot operating temperatures are generated and subsequently utilized. Realistic lattice configurations of the GE13 BWR fuel assemblies (unrodded and rodded) depleted under operating conditions were studied in this research because of their heterogeneous
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.
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Confidence level in the calculations of HCDA consequences using large codes
International Nuclear Information System (INIS)
The probabilistic approach to nuclear reactor safety is playing an increasingly significant role. For the liquid-metal fast breeder reactor (LMFBR) in particular, the ultimate application of this approach could be to determine the probability of achieving the goal of a specific line-of-assurance (LOA). Meanwhile a more pressing problem is one of quantifying the uncertainty in a calculated consequence for hypothetical core disruptive accident (HCDA) using large codes. Such uncertainty arises from imperfect modeling of phenomenology and/or from inaccuracy in input data. A method is presented to determine the confidence level in consequences calculated by a large computer code due to the known uncertainties in input invariables. A particular application was made to the initial time of pin failure in a transient overpower HCDA calculated by the code MELT-IIIA in order to demonstrate the method. A probability distribution function (pdf) for the time of failure was first constructed, then the confidence level for predicting this failure parameter within a desired range was determined
Recent progress with large-scale ab initio calculations: the CONQUEST code
Bowler, D. R.; Choudhury, R.; Gillan, M. J.; Miyazaki, T.
While the success of density functional theory (DFT) has led to its use in a wide variety of fields such as physics, chemistry, materials science and biochemistry, it has long been recognised that conventional methods are very inefficient for large complex systems, because the memory requirements scale as N 2 and the cpu requirements as N 3 (where N is the number of atoms). The principles necessary to develop methods with linear scaling of the cpu and memory requirements with system size (O(N ) methods) have been established for more than ten years, but only recently have practical codes showing this scaling for DFT started to appear. We report recent progress in the development of the Conquest code, which performs O(N ) DFT calculations on parallel computers, and has a demonstrated ability to handle systems of over 10000 atoms. The code can be run at different levels of precision, ranging from empirical tight-binding, through ab initio tight-binding, to full ab initio , and techniques for calculating ionic forces in a consistent way at all levels of precision will be presented. Illustrations are given of practical Conquest calculations in the strained Ge/Si(001) system.
A New, Efficient Stellar Evolution Code for Calculating Complete Evolutionary Tracks
Kovetz, Attay; Prialnik, Dina
2008-01-01
We present a new stellar evolution code and a set of results, demonstrating its capability at calculating full evolutionary tracks for a wide range of masses and metallicities. The code is fast and efficient, and is capable of following through all evolutionary phases, without interruption or human intervention. It is meant to be used also in the context of modeling the evolution of dense stellar systems, for performing live calculations for both normal star models and merger-products. The code is based on a fully implicit, adaptive-grid numerical scheme that solves simultaneously for structure, mesh and chemical composition. Full details are given for the treatment of convection, equation of state, opacity, nuclear reactions and mass loss. Results of evolutionary calculations are shown for a solar model that matches the characteristics of the present sun to an accuracy of better than 1%; a $1 \\Msun$ model for a wide range of metallicities; a series of models of stellar populations I and II, for the mass rang...
An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations
International Nuclear Information System (INIS)
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
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.
One and half dimensional particle in cell Euterpe code description
International Nuclear Information System (INIS)
A 1D 1/2 electromagnetic particle-in-cell named EUTERPE is described. Firstly, the main features of this code are reported. Secondly, stability tests are presented. As a conclusion, the present-day applications of this code in the electromagnetic field-plasma interaction is given
Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code
International Nuclear Information System (INIS)
A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were transported up to approximately 1 m before the region for benchmark calculation. Finally, the energy spectra of neutrons behind the very thick shield were calculated down to the thermal energy with good statistics, and typically agree well within a factor of two with the experimental data over a broad energy range. The 12C(n,2n)11C reaction rates behind the bulk shield were also calculated, which agree with the experimental data typically within 60%. These results are quite impressive in calculation accuracy for deep-penetration problem. In this report, the calculation conditions, geometry and the variance reduction techniques used in the deep-penetration calculation with the MARS14 code are clarified, and several subroutines of MARS14 which were used in our calculation are also given in the appendix. The numerical data of the calculated neutron energy spectra, reaction rates, dose rates and their C/E (Calculation/Experiment) values are also summarized. The
Numerical modeling of laser tunneling ionization in explicit particle-in-cell codes
International Nuclear Information System (INIS)
Methods for the calculation of laser tunneling ionization in explicit particle-in-cell codes used for modeling laser–plasma interactions are compared and validated against theoretical predictions. Improved accuracy is obtained by using the direct current form for the ionization rate. Multi level ionization in a single time step and energy conservation have been considered during the ionization process. The effects of grid resolution and number of macro-particles per cell are examined. Implementation of the ionization algorithm in two different particle-in-cell codes is compared for the case of ionization-based electron injection in a laser–plasma accelerator
FORTRAN Code for Glandular Dose Calculation in Mammography Using Sobol-Wu Parameters
Directory of Open Access Journals (Sweden)
Mowlavi A A
2007-07-01
Full Text Available Background: Accurate computation of the radiation dose to the breast is essential to mammography. Various the thicknesses of breast, the composition of the breast tissue and other variables affect the optimal breast dose. Furthermore, the glandular fraction, which refers to the composition of the breasts, as partitioned between radiation-sensitive glandular tissue and the adipose tissue, also has an effect on this calculation. Fatty or fibrous breasts would have a lower value for the glandular fraction than dense breasts. Breast tissue composed of half glandular and half adipose tissue would have a glandular fraction in between that of fatty and dense breasts. Therefore, the use of a computational code for average glandular dose calculation in mammography is a more effective means of estimating the dose of radiation, and is accurate and fast. Methods: In the present work, the Sobol-Wu beam quality parameters are used to write a FORTRAN code for glandular dose calculation in molybdenum anode-molybdenum filter (Mo-Mo, molybdenum anode-rhodium filter (Mo-Rh and rhodium anode-rhodium filter (Rh-Rh target-filter combinations in mammograms. The input parameters of code are: tube voltage in kV, half-value layer (HVL of the incident x-ray spectrum in mm, breast thickness in cm (d, and glandular tissue fraction (g. Results: The average glandular dose (AGD variation against the voltage of the mammogram X-ray tube for d = 4 cm, HVL = 0.34 mm Al and g=0.5 for the three filter-target combinations, as well as its variation against the glandular fraction of breast tissue for kV=25, HVL=0.34, and d=4 cm has been calculated. The results related to the average glandular absorbed dose variation against HVL for kV = 28, d=4 cm and g= 0.6 are also presented. The results of this code are in good agreement with those previously reported in the literature. Conclusion: The code developed in this study calculates the glandular dose quickly, and it is complete and
International Nuclear Information System (INIS)
Small-sample reactivity experiments are relevant to provide accurate information on the integral cross sections of materials. One of the specificities of these experiments is that the measured reactivity worth generally ranges between 1 and 10 pcm, which precludes the use of Monte Carlo for the analysis. As a consequence, several papers have been devoted to deterministic calculation routes, implying spatial and/or energetic discretization which could involve calculation bias. Within the Expert Group on Burn-Up Credit of the OECD/NEA, a benchmark was proposed to compare different calculation codes and methods for the analysis of these experiments. In four Sub-Phases with geometries ranging from a single cell to a full 3D core model, participants were asked to evaluate the reactivity worth due to the addition of small quantities of separated fission products and actinides into a UO2 fuel. Fourteen institutes using six different codes have participated in the Benchmark. For reactivity worth of more than a few tens of pcm, the Monte-Carlo approach based on the eigen-value difference method appears clearly as the reference method. However, in the case of reactivity worth as low as 1 pcm, it is concluded that the deterministic approach based on the exact perturbation formalism is more accurate and should be preferred. Promising results have also been reported using the newly available exact perturbation capability, developed in the Monte Carlo code TRIPOLI4, based on the calculation of a continuous energy adjoint flux in the reference situation, convoluted to the forward flux of the perturbed situation. (author)
Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment
International Nuclear Information System (INIS)
Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter und the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 times the hydraulic pipe diameter. (orig.)
Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment
International Nuclear Information System (INIS)
Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter and the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 tim es the hydraulic pipe diameter. (orig.)
A Comparative Study of the Code Calculations for Local Flow Blockages in the KALIMER-150 Core
Energy Technology Data Exchange (ETDEWEB)
Chang, Won Pyo; Ha, Ki Suk; Lee, Yong Bum
2009-09-15
A sub-channel blockage may be caused by ingression of damaged fuel debris or foreign obstacles into a core fuel subassembly for a liquid metal reactor(LMR) due to its geometrical compactness of the core design. Local coolant temperature could rise during the incident and it might eventually lead to the degradation of the fuel rods. An analysis computer code is obviously needed not only to assure the safe design of the core, but also to design an effective monitoring system to prevent it from propagating to a serious consequence. The code, therefore, must be capable of representing the thermal-hydraulic phenomena anticipated during the incident reasonably enough to be used to evaluate fuel rod intactness. Most of the technically leading countries for LMR have developed and are using the codes for sub-channel blockage analyses, Korea couldn't afford the resources sponsoring the develop of such a code past years. It was realized later that such an analysis code would ultimately be a prerequisite in the future licensing process of KALIMER as its conceptual design was being elaborated. Since those advanced countries had been reluctant to transfer such the codes, MATRA-LMR/FB had to be developed independently in Korea. It is a revised version of the existing MATRA-LMR code which was aimed for the core sub-channel analysis of LMRs. Some of its models have been improved so appropriately to be able to analyze the sub-channel blockages. Nevertheless, an experiment relevant to the sub-channel blockages had never been conducted for investigating the phenomenon in Korea. Further more, very few experimental data are available on published papers or reports world wide. Under this circumstance, a study has been made as an effort to evaluate the prediction capability of the MATRA-LMR/FB code by comparing the calculation results of the SABRE code, which had already been applied to EFR design. In result, a discrepancy has been observed in a case, but an overall agreement has
A Comparative Study of the Code Calculations for Local Flow Blockages in the KALIMER-150 Core
International Nuclear Information System (INIS)
A sub-channel blockage may be caused by ingression of damaged fuel debris or foreign obstacles into a core fuel subassembly for a liquid metal reactor(LMR) due to its geometrical compactness of the core design. Local coolant temperature could rise during the incident and it might eventually lead to the degradation of the fuel rods. An analysis computer code is obviously needed not only to assure the safe design of the core, but also to design an effective monitoring system to prevent it from propagating to a serious consequence. The code, therefore, must be capable of representing the thermal-hydraulic phenomena anticipated during the incident reasonably enough to be used to evaluate fuel rod intactness. Most of the technically leading countries for LMR have developed and are using the codes for sub-channel blockage analyses, Korea couldn't afford the resources sponsoring the develop of such a code past years. It was realized later that such an analysis code would ultimately be a prerequisite in the future licensing process of KALIMER as its conceptual design was being elaborated. Since those advanced countries had been reluctant to transfer such the codes, MATRA-LMR/FB had to be developed independently in Korea. It is a revised version of the existing MATRA-LMR code which was aimed for the core sub-channel analysis of LMRs. Some of its models have been improved so appropriately to be able to analyze the sub-channel blockages. Nevertheless, an experiment relevant to the sub-channel blockages had never been conducted for investigating the phenomenon in Korea. Further more, very few experimental data are available on published papers or reports world wide. Under this circumstance, a study has been made as an effort to evaluate the prediction capability of the MATRA-LMR/FB code by comparing the calculation results of the SABRE code, which had already been applied to EFR design. In result, a discrepancy has been observed in a case, but an overall agreement has been
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.)
MC2-2: a code to calculate fast neutron spectra and multigroup cross sections
International Nuclear Information System (INIS)
MC2-2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC2-2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC2-2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC2-2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC2-2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers
The modified calculation of the coolant temperature in the computer code TOODEE-2
International Nuclear Information System (INIS)
The programme is intended for the calculation of maximum cladding temperature of the hottest rod of a PWR and can be used to estimate the events after leakage of coolant. THe TOODEE-2 programme corresponds to the LOCTA-code of Westinghouse, which is gave a superior reproduction of reality. The new TOODEE-2 has been improved and a comparison of the calculation of a large fracture with Westinghouse gave good points of agreement for the first 16 seconds of the course of events. The rest describes a serious incident because of the conservation procedure according to 10 CFR 50 appendix K. Most of the calculations have used the form factor of 2.32. Also 2.12 which is the highest form factor for Ringhals 2 has been used. Furhter investigations are needed to clarify the difference in the results. (G.B.)
International Nuclear Information System (INIS)
AQUAMAN is an interactive computer code for calculating values of dose (50-year dose commitment) to man from aqueous releases of radionuclides from nuclear facilities. The data base contains values of internal and external dose conversion factors, and bioaccumulation (freshwater and marine) factors for 56 radionuclides. A maximum of 20 radionuclides may be selected for any one calculation. Dose and cumulative exposure index (CUEX) values are calculated for total body, GI tract, bone, thyroid, lungs, liver, kidneys, testes, and ovaries for each of three exposure pathways: water ingestion, fish ingestion, and submersion. The user is provided the option at the time of execution to change the default values of most of the variables, with the exception of the dose conversion factor values. AQUAMAN is written in FORTRAN for the PDP-10 computer
Program POD. A computer code to calculate cross sections for neutron-induced nuclear reactions
International Nuclear Information System (INIS)
A computer code, POD, was developed for neutron-induced nuclear data evaluations. This program is based on four theoretical models, (1) the optical model to calculate shape-elastic scattering and reaction cross sections, (2) the distorted wave Born approximation to calculate neutron inelastic scattering cross sections, (3) the preequilibrium model, and (4) the multi-step statistical model. With this program, cross sections can be calculated for reactions (n, γ), (n, n'), (n, p), (n, α), (n, d), (n, t), (n, 3He), (n, 2n), (n, np), (n, nα), (n, nd), and (n, 3n) in the neutron energy range above the resonance region to 20 MeV. The computational methods and input parameters are explained in this report, with sample inputs and outputs. (author)
Coding and traceability for cells, tissues and organs for transplantation.
Strong, D Michael; Shinozaki, Naoshi
2010-11-01
Modern transplantation of cells, tissues and organs has been practiced within the last century achieving both life saving and enhancing results. Associated risks have been recognized including infectious disease transmission, malignancy, immune mediated disease and graft failure. This has resulted in establishment of government regulation, professional standard setting and establishment of vigilance and surveillance systems for early detection and prevention and to improve patient safety. The increased transportation of grafts across national boundaries has made traceability difficult and sometimes impossible. Experience during the first Gulf War with mis-identification of blood units coming from multiple countries without standardized coding and labeling has led international organizations to develop standardized nomenclature and coding for blood. Following this example, cell therapy and tissue transplant practitioners have also moved to standardization of coding systems. Establishment of an international coding system has progressed rapidly and implementation for blood has demonstrated multiple advantages. WHO has held two global consultations on human cells and tissues for transplantation, which recognized the global circulation of cells and tissues and growing commercialization and the need for means of coding to identify tissues and cells used in transplantation, are essential for full traceability. There is currently a wide diversity in the identification and coding of tissue and cell products. For tissues, with a few exceptions, product terminology has not been standardized even at the national level. Progress has been made in blood and cell therapies with a slow and steady trend towards implementation of the international code ISBT 128. Across all fields, there are now 3,700 licensed facilities in 66 countries. Efforts are necessary to encourage the introduction of a standardized international coding system for donation identification numbers, such as ISBT
Recent R and D around the Monte-Carlo code Tripoli-4 for criticality calculation
Energy Technology Data Exchange (ETDEWEB)
Hugot, F.X.; Lee, Y.K.; Malvagi, F. [CEA - DEN/DANS/DM2S/SERMA/LTSD, Saclay (France)
2008-07-01
TRIPOLI-4 [1] is the fourth generation of the TRIPOLI family of Monte Carlo codes developed from the 60's by CEA. It simulates the 3D transport of neutrons, photons, electrons and positrons as well as coupled neutron-photon propagation and electron-photons cascade showers. The code addresses radiation protection and shielding problems, as well as criticality and reactor physics problems through both critical and subcritical neutronics calculations. It uses full pointwise as well as multigroup cross-sections. The code has been validated through several hundred benchmarks as well as measurement campaigns. It is used as a reference tool by CEA as well as its industrial and institutional partners, and in the NURESIM [2] European project. Section 2 reviews its main features, with emphasis on the latest developments. Section 3 presents some recent R and D for criticality calculations. Fission matrix, Eigen-values and eigenvectors computations will be exposed. Corrections on the standard deviation estimator in the case of correlations between generation steps will be detailed. Section 4 presents some preliminary results obtained by the new mesh tally feature. The last section presents the interest of using XML format output files. (authors)
YNOGK: A new public code for calculating null geodesics in the Kerr spacetime
Yang, Xiaolin
2013-01-01
Following \\cite{dexagol2009} we present a new public code for the fast calculation of null geodesics in the Kerr spacetime. Using Weierstrass' and Jacobi's elliptic functions, we express all coordinates and affine parameters as analytical and numerical functions of a parameter $p$, which is an integral value along the geodesic. This is a main difference of our code compares with previous similar ones. The advantage of this treatment is that the information about the turning points do not need to be specified in advance by the user, and many applications such as imaging, the calculation of line profiles or the observer-emitter problem, etc become root finding problems. All elliptic integrations are computed by Carlson's elliptic integral method as \\cite{dexagol2009} did, which guarantees the fast computational speed of our code. The formulae to compute the constants of motion given by \\cite{cunnbard1973} have been extended, which allow one readily to handle the situations, in which the emitter or the observer ...
CEQCSY: a new code for chemical equilibrium calculation in multiphased systems
International Nuclear Information System (INIS)
As part of the CEC Chemval/mirage project, a method is presented for calculating the thermodynamic equilibrium state of a multiphase system, by minimizing its Gibbs free energy constrained by mass balances. Compared to the other algorithms available in the literature, the method has three main characteristics: - the sets of equations corresponding to the conditions of homogeneous and heterogeneous equilibria are simultaneously solved, - a mathematical criterion for bringing a new multicomponent phase in the system is rigorously demontrated. - It enables a detailed representation of the multisite solid solutions with constraints called site closure relation. The code CEQCSY (Chemical Equilibrium in Complex SYstem) uses this formalism, and works with the thermodynamic data base from the EQ3/6 code. This choice makes easier several compared tests with EQ6: quartz dissolution in water, water-atmospheric air equilibrium, theoretical re-equilibrium of seawater, hydrothermal alteration of granite including solid solutions. The test results demonstrate the high efficiency and velocity of the code CEQCSY, when working on equilibrium state of multiphase systems. This high velocity was the aim of this work, in order to couple with thermic, hydrodynamic or mechanic codes
SPARC-90: A code for calculating fission product capture in suppression pools
International Nuclear Information System (INIS)
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
Benchmark calculations by the thermal reactor standard nuclear design code system SRAC
International Nuclear Information System (INIS)
This report summarizes the present status of the thermal reactor standard nuclear design code system SRAC developed by the nuclear design working group of the JAERI thermal reactor standard code committee which was started on July 1978. Descriptions are given at first on the brief introduction and the process of development of the code system SRAC, and then, the several benchmark tests performed to evaluate the performance of the code system. The results show the good predictions of the experimental keff values of the critical facilities; TCA for LWR, JMTRC for JAERI MTR, DCA for the Japanese Advanced Thermal Reactor and SHE for VHTR. A trial to the IAEA benchmark calculations on the Reduction of uranium Enrichment of Research and Test Reactors yields satisfactory agreements with the results of ANL. Another test to evaluate the fast group constants was also attempted by tracing the fast reactor benchmark problems which have been used to evaluate nuclear data file in the FBR reactor physics field. (author)
International Nuclear Information System (INIS)
The method to calculate the response function of spherical BF3 proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF3 proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within ±10%. (author)
The 'hot rod methodology' for intermediate break LOCA calculations with the CATHARE2 code
International Nuclear Information System (INIS)
The nuclear fuel that will be used for next decades in French PWR may be quite different from current ones, including technological evolutions of cladding alloys, new types of assemblies, etc. Beside these structural evolutions, calculation methodologies are also changing with the use of Best-estimate multi-physics multi-scale coupled calculations, and complex physical models development at 3D local scale. In the frame of its LOCA R and D program, the French 'Institut de Radioprotection et de Surete Nucleaire' is developing its own calculation methodology to figure out the maximum cladding temperature and the maximum oxidation rate reached by the hottest rod in the core. This HOT ROD methodology has been worked out for Intermediate Break LOCA calculations using the 'Best-Estimate' CATHARE2 system code. This calculation is based on a first modeling called 1D SYSTEM calculation which simulates the thermohydraulical behavior of the whole reactor. During this computation, key parameters are stored at core ends in order to be used as boundary conditions for a second calculation called 1D CHAINED calculation nodelizing in one dimension the hottest assembly of the core. Based on experimental evidences, the hot assembly has been considered to have no influence on the thermohydraulical behavior of the mean core. Hence its hydraulical environment has been modified at each calculation's timestep in order to fit with the SYSTEM mean one. This paper details the methodology's validation phase on PERICLES 2D BOIL-UP tests, which have quite the same swollen-level evolution's rate as in the boil up phase of an intermediate break LOCA. (author)
Development of burnup calculation function in reactor Monte Carlo code RMC
International Nuclear Information System (INIS)
This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua University of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency. (authors)
IRACM : A code system to calculate induced radioactivity produced by ions and neutrons
International Nuclear Information System (INIS)
It is essential to estimate of radioactivity induced in accelerator components and samples bombarded by energetic ion beams and the secondary neutrons of high-energy accelerator facilities in order to reduce the amount of radioactive wastes and to minimize radiation exposure to personnel. A computer code system IRACM has been developed to estimate product nuclides and induced radioactivity in various radiation environments of accelerator facilities. Nuclide transmutation with incident particles of neutron, proton, deuteron, alpha, 12C, 14N, 16O, 20Ne and 40Ar can be computed for arbitrary multi-layer target system in a one-dimensional geometry. The code system consists of calculation modules and libraries including activation cross sections, decay data and photon emission data. The system can be executed in both FACOM-M780 mainframe and DEC workstations. (author)
Wall surface temperature calculation in the SolEdge2D-EIRENE transport code
Denis, J.; Pégourié, B.; Bucalossi, J.; Bufferand, H.; Ciraolo, G.; Gardarein, J.-L.; Gaspar, J.; Grisolia, C.; Hodille, E.; Missirlian, M.; Serre, E.; Tamain, P.
2016-02-01
A thermal wall model is developed for the SolEdge2D-EIRENE edge transport code for calculating the surface temperature of the actively-cooled vessel components in interaction with the plasma. This is a first step towards a self-consistent evaluation of the recycling of particles, which depends on the wall surface temperature. The proposed thermal model is built to match both steady-state temperature and time constant of actively-cooled plasma facing components. A benchmark between this model and the Finite Element Modelling code CAST3M is performed in the case of an ITER-like monoblock. An example of application is presented for a SolEdge2D-EIRENE simulation of a medium-power discharge in the WEST tokamak, showing the steady-state wall temperature distribution and the temperature cycling due to an imposed Edge Localised Mode-like event.
DCHAIN-SP 2001: High energy particle induced radioactivity calculation code
Energy Technology Data Exchange (ETDEWEB)
Kai, Tetsuya; Maekawa, Fujio; Kasugai, Yoshimi; Takada, Hiroshi; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kosako, Kazuaki [Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan)
2001-03-01
For the purpose of contribution to safety design calculations for induced radioactivities in the JAERI/KEK high-intensity proton accelerator project facilities, the DCHAIN-SP which calculates the high energy particle induced radioactivity has been updated to DCHAIN-SP 2001. The following three items were improved: (1) Fission yield data are included to apply the code to experimental facility design for nuclear transmutation of long-lived radioactive waste where fissionable materials are treated. (2) Activation cross section data below 20 MeV are revised. In particular, attentions are paid to cross section data of materials which have close relation to the facilities, i.e., mercury, lead and bismuth, and to tritium production cross sections which are important in terms of safety of the facilities. (3) User-interface for input/output data is sophisticated to perform calculations more efficiently than that in the previous version. Information needed for use of the code is attached in Appendices; the DCHAIN-SP 2001 manual, the procedures of installation and execution of DCHAIN-SP, and sample problems. (author)
A code for the calculation of self-absorption fractions of photons
International Nuclear Information System (INIS)
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
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)
THEBES: a thermal hydraulic code for the calculation of transient two phase flow in bundle geometry
International Nuclear Information System (INIS)
The three dimensional thermal hydraulic code THEBES, capable to calculate transient boiling of sodium in rod bundles is described here. THEBES, derived from the transient single phase code SABRE-2A, was developed in CADARACHE by the SIES to analyse the SCARABEE N loss of flow experiments. This paper also presents the results of tests which were performed against various types of experiments: (1) transient boiling in a 7 pin bundle simulating a partial blockage at the bottom of a subassembly (rapid transient SCARABEE 7.2 experiment), (2) transient boiling in a 7 pin bundle simulating a coolant coast down (slow transient SCARABEE 7.3 experiment), (3) steady local and generalised boiling in a 19 pin bundle (GR 19 I experiment), (4) transient boiling in a 19 pin bundle simulating a coolant coast down (GR 19 I experiment), (5) steady local boiling in a 37 pin bundle with internal blockage (MOL 7C experiment). Excellent agreement was found between calculated and experimental results for these different situations. Our conclusion is that THEBES is able to calculate transient boiling of sodium in rod bundles in a quite satisfying way
International Nuclear Information System (INIS)
Three papers are brought together as a result of collaboration between the Nuclear Engineering Program of COPPE-UFRJ (Coordination of Engineering Post graduate Programs of the Federal University of Rio de Janeiro) and the Division of Nuclear Studies of the Department of Thermal generation of FURNAS S.A. aiming at the analysis of the neutronic behavior of PWR power reactors' core. Modifications were introduced in the methods of calculation utilized by the LEOPARD code. The results presented in the first two papers refer to the calculation of neutron flux in the homogenized reactor, only the dependence on energy being considered. Physical models and mathematical approximations are utilized as an alternative to those conventionally used in the code for the calculation of thermal and non-thermal flux. Some parameters, such as the average thermal cross sections of some elements showed to be sensible to the modifications introduced, and indicate that it is useful to carry on the study. In the third paper, comments are made on the MND (Mixed Number Density) method of effectuating the thermal average of the diffusion coefficient and of the absorption cross section, for application in the diffusion equation and consequent determination of flux in function of the spacial position. (I.C.R.)
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.
Calculation capability of NETFLOW++ code for natural circulation in sodium cooled fast reactor
International Nuclear Information System (INIS)
The present paper describes the simulation of the natural circulation in the secondary heat transport system (HTS) after an intentional plant trip of the experimental fast reactor 'Joyo' at 140MWt power using the plant dynamics analysis code NETFLOW++. This code is an integrated network code to calculate the nuclear steam supply system (NSSS) and the balance of the plant (BOP), i.e., turbine/feedwater system developed by the author. Up to now, the code has been validated using transient data of the experimental sodium facility PLANDTL, experimental fast reactor 'Joyo' and the prototype fast breeder reactor 'Monju'. These validations are steps to evaluate the natural circulation of a large-scale fast breeder reactor because it is important that the code can simulate not only a large plant but also a small plant or apparatus in order to have versatility. Therefore, the former validation results are introduced to show the degree of agreement. In order to consolidate the applicability of the code to the evaluation of the natural circulation, the present simulation was selected. Natural circulation tests were conducted at the experimental fast reactor 'Joyo' for the two reactor cores. The latest natural circulation test was conducted from the rated power of the Mark-II irradiation core at 100 MWt, and the NETFLOW++ code was validated using the measured data. At this occasion, the importance of the inter-subassembly heat transfer was recognized by the author in order to predict the core outlet temperature although the whole plant behavior was little effected. The 'Monju' reactor conducted some natural circulation tests in the primary and secondary heat transport systems utilizing the heat of primary pump operation. The NETFLOW++ code was validated using these data. Although these were not the complete natural circulation expected in the power plant,these results were important because basic factors relating the natural circulation such as the buoyancy force,the static
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.
DYN3D/M2 - a Code for Calculation of Reactivity Transients in Cores with Hexagonal Geometry
Rohde, Ulrich; Grundmann, Ulrich
2010-01-01
The code DYN3D/M2 consists of a the 3-dimensional neutron kinetic model of the code HEXDYN3D and the thermohydraulic model of the code FLOCAL. The neutron kinetics of DYN3D/M2 is calculated by using a nodal expansion method (NEM) for hexagonal geometry. The developed method solves the neutron diffusion equation for two energy groups. Stationary state and transient behaviour can be calculated. By help of the code PREPAR-EC parameterizid neutron physical constants of given burnup distribution c...
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.
International Nuclear Information System (INIS)
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
OPT13B and OPTIM4 - computer codes for optical model calculations
International Nuclear Information System (INIS)
OPT13B is a computer code in FORTRAN for optical model calculations with automatic search. A summary of different formulae used for computation is given. Numerical methods are discussed. The 'search' technique followed to obtain the set of optical model parameters which produce best fit to experimental data in a least-square sense is also discussed. Different subroutines of the program are briefly described. Input-output specifications are given in detail. A modified version of OPT13B specifications are given in detail. A modified version of OPT13B is OPTIM4. It can be used for optical model calculations where the form factors of different parts of the optical potential are known point by point. A brief description of the modifications is given. (author)
Power distribution and fuel depletion calculation for a PWR, using LEOPARD and CITATION codes
International Nuclear Information System (INIS)
By modifying LEOPARD a new program, LEOCIT, has been developed in which additional subroutines prepare cross-section libraries in 1, 2 or 4 energy groups and subsequently record these on disc or tape in a format appropriate for direct input to the CITATION code. Use of LEOCIT in conjunction with CITATION is demonstrated by simulating the first depletion cycle of Angra Unit 1. In these calculations two energy groups are used in 1/4, X - Y geometry to give the soluble boron curve, the fuel depletion and the point to point power distribution in Angra 1. Finally relevant results obtained here are compared with those published by Westinghouse, CNEN and Furnas and recommendations are made to improve the system of neutronic calculation developed in this work. (Author)
FOOD: an interactive code to calculate internal radiation doses from contaminated food products
International Nuclear Information System (INIS)
An interactive code, FOOD, has been written in BASIC for the UNIVAC 1108 to facilitate calculation of internal radiation doses to man from radionuclides in food products. In the dose model, vegetation may be contaminated by either air or irrigation water containing radionuclides. The model considers two mechanisms for radionuclide contamination of vegetation: direct deposition on leaves and uptake from soil through the root system. The user may select up to 14 food categories with corresponding consumption rates, growing periods and either irrigation rates or atmospheric deposition rates. These foods include various kinds of produce, grains and animal products. At present, doses may be calculated for the skin, total body and five internal organs from 190 radionuclides. Dose summaries can be displayed at the local terminal. Further details on percent contribution to dose by nuclide and by food type are available from an auxiliary high-speed printer. This output also includes estimated radionuclide concentrations in soil, plants and animal products
International Nuclear Information System (INIS)
In the fuel rods of the first DUELL experiment highly asymmetric fuel structures were found which had been caused by a steep transversal neutron flux gradient and eccentric pellet location. The TEXDIF-P computer code was developed to explain this phenomenon in quantitative terms. This computer code solves for an encapsulated fuel rod the equation of two-dimensional heat conduction using the finite differences method. Any distribution may be specified of the heat source density and of the gap between the fuel pellet and the cladding tube. By use of the modular structure the material relations are easily exchangeable. The TEXDIF-P code can be applied both to oxide and to carbide fuel rods. Coupling of the POUMEC subprogram of SATURN-1 allows the dynamic calculation of pore migration. Independent of this, the program includes an option for determination of the limit of the pore migration zone via a relation covering the minimum pore migration path according to Olander. TEXDIF-P has been used so far to verify the first start-up ramp experiment of DUELL. The agreement between the computation and the findings of post-examinations is quite satisfactory regarding the size and the location of the central void. Also the limit of the compacted zone is fairly well reproduced by the computation. The assumption on the size of the transversal neutron flux gradient has been essentially confirmed retroactively by transversal γ-scanning. (orig.)
Opacity calculation for target physics using the ABAKO/RAPCAL code
Mínguez, E.; Florido, R.; Rodríguez, R.; Gil, J. M.; Rubiano, J. G.; Mendoza, M. A.; Suárez, D.; Martel, P.
2010-01-01
Radiative properties of hot dense plasmas remain a subject of current interest since they play an important role in inertial confinement fusion (ICF) research, as well as in studies on stellar physics. In particular, the understanding of ICF plasmas requires emissivities and opacities for both hydro-simulations and diagnostics. Nevertheless, the accurate calculation of these properties is still an open question and continuous efforts are being made to develop new models and numerical codes that can facilitate the evaluation of such properties. In this work the set of atomic models ABAKO/RAPCAL is presented, as well as a series of results for carbon and aluminum to show its capability for modeling the population kinetics of plasmas in both LTE and NLTE regimes. Also, the spectroscopic diagnostics of a laser-produced aluminum plasma using ABAKO/RAPCAL is discussed. Additionally, as an interesting application of these codes, fitting analytical formulas for Rosseland and Planck mean opacities for carbon plasmas are reported. These formulas are useful as input data in hydrodynamic simulation of targets where the computation task is so hard that in line computation with sophisticated opacity codes is prohibitive.
Design, experiments and Relap5 code calculations for the perseo facility
International Nuclear Information System (INIS)
Research on innovative safety systems for light water reactors addressed to heat removal by in-pool immersed heat exchangers, led to design, build-up and test the PERSEO facility at SIET laboratories. The research started with the CEA-ENEA proposal of improving the GE-SBWR isolation condenser system, by moving the triggering valve from the high pressure primary side of the reactor to the low pressure pool side. A new configuration of the system was defined with the heat exchanger contained in a small pool, connected at bottom and top to a large water reservoir pool, the triggering valve being located on the pool bottom connecting pipe. ENEA funded the whole activity that included the definition and build-up of a new heat exchanger pool, on the basis of the already existing PANTHERS IC-PCC facility, at SIET laboratories, and the new plant requirements. The heat exchanger connections to the pressure vessel were maintained. An experimental campaign was executed at full scale and full thermal-hydraulic conditions for investigating the behaviour and performance of the plant in steady and unsteady conditions. The Relap5 code was utilised during all phases of the research: for the heat exchanger pool dimension definition and from pre-test and post-test analyses. The Cathare code was applied too from pre-test and post-test analyses. This paper deals with the experimental and calculated results limited to the Relap5 code
SMARTIES: User-friendly codes for fast and accurate calculations of light scattering by spheroids
Somerville, W. R. C.; Auguié, B.; Le Ru, E. C.
2016-05-01
We provide a detailed user guide for SMARTIES, a suite of MATLAB codes for the calculation of the optical properties of oblate and prolate spheroidal particles, with comparable capabilities and ease-of-use as Mie theory for spheres. SMARTIES is a MATLAB implementation of an improved T-matrix algorithm for the theoretical modelling of electromagnetic scattering by particles of spheroidal shape. The theory behind the improvements in numerical accuracy and convergence is briefly summarized, with reference to the original publications. Instructions of use, and a detailed description of the code structure, its range of applicability, as well as guidelines for further developments by advanced users are discussed in separate sections of this user guide. The code may be useful to researchers seeking a fast, accurate and reliable tool to simulate the near-field and far-field optical properties of elongated particles, but will also appeal to other developers of light-scattering software seeking a reliable benchmark for non-spherical particles with a challenging aspect ratio and/or refractive index contrast.
Calculation of the CAREM reactor with the HUEMUL-PUMA-THERMIT chain of codes
International Nuclear Information System (INIS)
The purpose of the work was the evaluation of the the CAREM 25 reactor core, using a chain of codes (HUEMUL-PUMA-THERMIT) different to the one used in the original design (CONDOR-CITVAP-THERMIT). First, we performed a partial validation of the our codes in lattices similar to CAREM and reproduced a benchmark for simulation of gadolinium burnup. The results were considered satisfactory for this stage of the project. Then, we calculated the core along the normal operating equilibrium cycle and in hot and cold shut-down conditions. The main outcome of our evaluation confirms the general behaviour of the reference calculations except in one important point referring to the cold shut down. In this condition, the failure of one single rod of bank number 13 of the shut down system, leaves the core in a supercritical state at the beginning of the cycle and this anomaly persists during almost a third of the overall cycle. A new design of the core is proposed with minor modifications of the reference one, without introducing new types of fuel elements and keeping the same fuel management scheme. This new core fulfills all the design requirements. (author)
Decay heat experiment and validation of calculation code systems for fusion reactor
International Nuclear Information System (INIS)
Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)
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)
FREEZE PROFILE AND HEAT BALANCE CALCULATION OF THE 160kA DRAINED CELL
Institute of Scientific and Technical Information of China (English)
X.P.Li; J.Li; Y.Q.Lai; H.Q.Zhao; Y.X.Liu
2004-01-01
A 2D full cell thermo-electric model of 160kA drained cell was set up using finite element code to calculate its freeze profile,then the drained cell model was modified according to the freeze profile computed and its heat balance was calculated.Compared with that of a 160kA conventional Hall-Heroult cell(H-H cell),though the melts volume of the drained cell reduced greatly,the whole heat loss from it didn't drop down apparently,and an analysis was presented in the paper.On the other hand,the anodecathode distance(ACD)of a drained cell was much less than that of a H-H cell,so the voltage drop on it and heat produced decreased too,steps should be taken to keep a workable heat balance on a drained cell.
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)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
International Nuclear Information System (INIS)
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
International Nuclear Information System (INIS)
The present report describes a computer code DEEP which calculates the organ dose equivalents and the effective dose equivalent for external photon exposure by the Monte Carlo method. MORSE-CG, Monte Carlo radiation transport code, is incorporated into the DEEP code to simulate photon transport phenomena in and around a human body. The code treats an anthropomorphic phantom represented by mathematical formulae and user has a choice for the phantom sex: male, female and unisex. The phantom can wear personal dosimeters on it and user can specify their location and dimension. This document includes instruction and sample problem for the code as well as the general description of dose calculation, human phantom and computer code. (author)
Calculation of equilibria at elevated temperatures using the MINTEQ geochemical code
Energy Technology Data Exchange (ETDEWEB)
Smith, R.W.
1988-12-01
Coefficients and equations for calculating mineral hydrolysis constants, solubility products and formation constants for 60 minerals and 57 aqueous species in the 13 component thermodynamic system K/sub 2/O-Na/sub 2/O-CaO-MgO-FeO-Al/sub 2/O/sub 3/-SiO/sub 2/-CO/sub 2/-H/sub 2/O-HF-HCl-H/sub 2/S-H/sub 2/SO/sub 4/ are presented in a format suitable for inclusion in the MINTEQ computer code. The temperature functions presented for minerals are based on the MINTEQ data base at 25/degree/C and the integration of analytical heat capacity power functions. This approach ensures that the temperature functions join smoothly with the low-temperature data base. A new subroutine, DEBYE, was added to MINTEQ that is used to calculate the theoretical Debye-Hueckel parameters A and B as a function of temperature. In addition, this subroutine also calculates a universal value of the extended Debye-Hueckel parameter, b/sub i/, as a function of temperature. The coefficients and equations provide the capability to use MINTEQ to more accurately calculate water/rock equilibrium for temperatures of up to 250/degree/C, and in dilute, low-sulfate, near neutral groundwaters to 300/degree/C. 52 refs., 1 fig., 6 tabs.
Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code
Nunomiya, T; Nakamura, T; Nakao, N
2002-01-01
A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were trans...
Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations
International Nuclear Information System (INIS)
This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO2-Gd2O3 poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)
Three-dimensional whole core transport calculation method and performance of the DeCART code
International Nuclear Information System (INIS)
The three-dimensional (3D) transport calculation method implemented in a whole core neutron transport code DeCART is presented and its performance is examined in terms of solution accuracy and execution speed. The 3D flux calculation in DeCART is based on a transverse-integration method in which the radial and axial dependencies are handled separately. The radial dependence is resolved by the elaborated two-dimensional method of characteristics (MOC) whereas the axial dependence is dealt with the simple one-dimensional diffusion model. The global balance of the 3D flux distribution is incorporated by the coarse mesh finite difference (CMFD) formulation. It is shown that the CMFD formulation enables the approximate three-dimensional transport calculation through the transverse-integration, and furthermore it is very effective in achieving rapid convergence. The accuracy of the approximate 3D whole-core transport calculation method is proved by analyzing rodded variations of the C5G7 MOX heterogeneous core benchmark problem for which Monte Carlo solutions are generated as the reference
International Nuclear Information System (INIS)
This report documents a modular data interpretation computer code. The MASFLO code is a Fortran code used in the Oak Ridge National Laboratory Blowdown Heat Transfer Program to convert measured quantities of density, volumetric flow, and momentum flux into a calculated quantity: mass flow rate. The code performs both homogeneous and two-velocity calculations. The homogeneous models incorporate various combinations of the Thermal-Hydraulic Test Facility instrumented spool piece turbine flow meter, gamma densitometer, and drag disk readings. The two-velocity calculations also incorporate these instruments, but in models developed by Aya, Rouhani, and Popper. Each subroutine is described briefly, and input instructions are provided in the appendix along with a sample of the code output
Input model of a VVER 440/213 fuel assembly and CFD calculations by the FLUENT code
International Nuclear Information System (INIS)
The preparation of the final version of the computation network of a VVER 440/213 fuel assembly by the GAMBIT code is described. Qualified estimates of the size of the networks (numbers of cells) are made especially in dependence on the network density in the transverse section. Some sensitivity analyses performed on smaller geometries, devoted to the effect of computation network density on turbulence modelling and to the effect of the limiting layer thickness on the temperature and flow rate fields are described. Of importance was the analysis of spacer grid replacement by means of the porous media function with regard to savings in the number of computational cells. The CFD calculations in the FLUENT code, performed first for smaller test problems and subsequently for large problems describing 1/6 to 1/2 of the fuel assembly, are described. Analyzed are the coolant distribution within the assembly after passing the bottom supporting plate and the effect of the spacer grids on coolant distribution and on heat transfer
SFR whole core burnup calculations with TRIPOLI-4 Monte Carlo code
International Nuclear Information System (INIS)
Under the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD/NEA, an international collaboration benchmark was recently established on the neutronic analysis of four Sodium-cooled Fast Reactor (SFR) concepts of the Generation- IV nuclear energy systems. As the whole core Monte Carlo depletion calculation is one of the essential challenges of current reactor physics studies, the continuous-energy TRIPOLI-4 Monte Carlo transport code was firstly used in this study to perform whole core 3D neutronic calculations for these four SFR cores. Two medium size (1000 MWt) and two large size (3600 MWt) SFR of GEN-IV systems were analyzed. The medium size SFR concepts are from the Advanced Burner Reactors (ABR). The large size SFR concepts are from the self-breeding reactors. The TRIPOLI-4 depletion calculations were made with MOX and metallic U-Pu-Zr fuels for the ABR cores and with MOX and Carbide (U,Pu)C fuels for the self-breeding cores. The whole core reactor physics parameters calculations were performed for the BOEC and EOEC (Beginning and End of Equilibrium Cycle) conditions. This paper summarizes the TRIPOLI-4 calculation results of Keff, βeff, sodium void worth, Doppler constant, control rod worth, and core power distributions for the BOEC and EOEC conditions. The one-cycle depletion calculation results of the core inventory of U and TRU (Pu, Am, Cm, and Np) are also analyzed, after 328.5 days depletion irradiation for the ABR cores, 410 days for the large MOX core, and 500 days for the large carbide core. (author)
International Nuclear Information System (INIS)
A computer code system for fast calculation of activation and transmutation has been developed. The system consists of a driver code, cross-section libraries, flux libraries, a material library, and a decay library. The code is used to predict transmutations in a Ti-modified 316 stainless steel, a commercial ferritic alloy (HT9), and a V-15%Cr-5%Ti alloy in various magnetic fusion energy (MFE) test facilities and conceptual reactors
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For the analysis of transient and emergency processes during reactor operation it is necessary to have a set of codes, which calculate physical processes with a various degree of accuracy. Codes CORT and BUMT for three-dimensional thermohydraulic calculation of fast reactor core in steady state, transient and accident conditions are described in this paper. The code CORT calculates thermohydraulics of the whole fast reactor core or group of subassemblies in simplified approximation. The core is described as a set of coupled one-dimensional channels or is divided into a set of ring zones, each of those is also represented by one subassembly (S/A). The detailed three-dimensional calculation of particular S/A is carried out by code BUMT. For description of S/A thermohydraulics the authors have chosen so called 'subchannel model. In this model the S/A is split into number of channels exchanging one by one with mass, momentum and energy. The coefficients of inter channel exchange are calculated on the basis of empirical correlations. The subchannel model is supplemented by detailed (two-dimensional in each axial cross-section) calculation of fuel pin and S/A wrapper temperatures. For solution of hydrodynamic equations the full-implicit scheme is used. Code BUMT was verified using experimental data for S/A-simulators and results of calculations obtained by other codes. These codes when used in complex with neutronic code and first circuit thermohydraulic code could describe in detail the thermal state of coolant and performance of fuel pins and construction elements of reactor during steady and transient states of its operation. (author)
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Analysis of known approaches in area of verification and validation of calculation means (codes) modelling the accident / transition processes in nuclear power plant (NPP) equipment is represented in this review article. Needs to develop and realise the generalised calculation means verification / validation methodology taking into account, together with traditional procedures, the codes applicability assessment criteria for decision of specific tasks and for specific equipment, mathematical models and experimental stands adequacy to full-scale conditions are shown
International Nuclear Information System (INIS)
In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B4C oxidation do not affect significantly the total calculated hydrogen release rates. (authors)
Energy Technology Data Exchange (ETDEWEB)
Austregesilo, Henrique [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching (Germany)]. E-mail: Henrique.Austregesilo@grs.de; Bals, Christine [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching (Germany); Trambauer, Klaus [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching (Germany)
2007-09-15
In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B{sub 4}C oxidation do not affect significantly the total calculated hydrogen release rates.
International Nuclear Information System (INIS)
In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B4C oxidation do not affect significantly the total calculated hydrogen release rates
Solution of the BEAVRS benchmark using the nTRACER direct whole core calculation code
International Nuclear Information System (INIS)
The BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulation) benchmark is solved by the nTRACER direct whole core calculation code to assess its accuracy and to examine the solution dependence on modeling parameters. A sophisticated nTRACER core model representing the BEAVRS core is prepared after a series of sensitivity study to ensure solution accuracy. The resulting solutions for several hot-zero-power (HZP) states are compared first with the corresponding Monte Carlo solutions, which consist of the McCARD solutions for the assembly problems and the OpenMC solutions for the core problems, and then with the measured data which include the control rod worths (CRWs) and incore detector signals as well as the critical boron concentrations (CBC). The core depletion calculation is performed for the initial and second cycles with a set of approximated power histories and the calculated CBCs are compared with the measured data. The comparison results show that the criticality, control rod bank worths at HZP and the boron let-down curves of two cycles agree well with the measurements within 180 pcm and 25 ppm, respectively. (author)
BARS - a heterogeneous code for 3D pin-by-pin LWR steady-state and transient calculation
International Nuclear Information System (INIS)
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)
Computing Moment-Based Probability Tables for Self-Shielding Calculations in Lattice Codes
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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
International Nuclear Information System (INIS)
Criticality calculation codes/code systems MCNP, MVP, SCALE and JACS, which are currently typically used in Japan for nuclear criticality safety evaluation, were benchmarked for so called dissolver-typed systems, i.e., fuel rod arrays immersed in fuel solution. The benchmark analyses were made for the evaluated critical experiments published in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook: one evaluation representing five critical configurations from heterogeneous core of low-enriched uranium dioxides at the Japan Atomic Energy Research Institute and two evaluations representing 16 critical configurations from heterogeneous core of mixed uranium and plutonium dioxides (MOXs) at the Battelle Pacific Northwest Laboratories of the U.S.A. The results of the analyses showed that the minimum values of the neutron multiplication factor obtained with MCNP, MVP, SCALE and JACS were 0.993, 0.990, 0.993, 0.972, respectively, which values are from 2% to 4% larger than the maximum permissible multiplication factor of 0.95. (author)
International Nuclear Information System (INIS)
A computer code HEXNEM based on a Nodal Expansion method has been developed for solving of the neutron diffusion equation for two-dimensional hexagonal geometries. The nodal equations are derived using higher order polynomial approximations to the spatial dependence of the flux within the hexagonal node. The final equations which are cast in the form of inhomogeneous response matrix equations for each group, involve spatial moments of the node interior flux distribution and surface averaged partial currents across the faces of the node. These equations are solved using a conventional fission source iteration accelerated by coarse mesh rebalance and asymptotic source extrapolation. Numerical calculations for models of reactor cores like PFBR or VVER designs have shown accuracy of the nodal schemes to be superior to that the coarse mesh finite difference method. The higher order axial approximations in the nodal method permit the use of an axial mesh which is at least four times coarser than a typical finite difference mesh. Two benchmark problems have been solved and the results are presented. This report describes the mathematical development and numerical solution of the nodal equations, as well as the input options used in the computer code. (author)
NXDC-neutron and x-ray diffraction code for crystal structures calculations
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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
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The authors investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. They consider the 2.0347 to 3.3546 keV energy region for /sup 238/U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
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.)
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This report presents a detailed analysis of the results obtained by five estimate calculations concerning the safety of the first test of the Sodium-water reaction experimental programme, FLASH. This programme is carried out the frame of the LMFBR R and D European Agreement. The calculations were executed by the PLEXUS code of the CASTEM system
International Nuclear Information System (INIS)
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. Each carrier can be loaded with 4 cardboard boxes (0.4x0.4x 0.4 m3 ). Each box was divided in eight equal cubes. Absorbed dose rate, gamma flux per energy interval and average gamma energy were calculated inside the eight cubes. For comparison purposes, boxes filled with air and 'dummy' boxes loaded with layers of folded and crumpled newspapers to reach the desired density were used. The contributions from source, irradiator structures, sample material and other origins (ceiling, floor and walls) for the total photon spectra were also calculated. The dose distribution in the irradiator depends on the material and its density. These results show that the MCNP is an important tool to perform a dose mapping of the irradiator for each material to be irradiated. The economic benefits of the knowledge of the dose mapping for each material are important because allow to save time utilisation of UTR, dosimeters and man power. The previous knowledge of the dose mapping permits to establish an appropriate irradiation planning, which results in good dose uniformity in the material and reducing previous experimental work. (author)
International Nuclear Information System (INIS)
The calculations performed for the Almaraz Unit 2 PWR using the code packages of the Atomic Energy Corporation of South Africa Ltd. are summarized. These calculations were done as part of the IAEA Coordinated Research Programme on In-Core Fuel Management Code Package Validation for LWRs. A brief description of the one-dimensional cross section generation package as well as of the Level II (scoping type) global core calculational package which was used is given. Detailed results are presented in several appendices. 29 figs., 20 tabs., 10 refs
International Nuclear Information System (INIS)
The results of thermal hydraulic analyses of anticipated transients without scram (ATWS) served as the basis for the new Emergency Operating Procedures for WWER-440/V-213 reactors. Because of the differences in the behavior of parameters in the calculations by the ATHLET code (for the Dukovany NPP) and by the RELAP code (for the Bohunice V2 plant), the major parameters in selected calculations were compared and the differences were explained on graphs. The starting calculations, in which no operator intervention was taken into account, were used for the comparison. (P.A.)
International Nuclear Information System (INIS)
MUDE is a nuclear code written in FORTRAN II for IBM 7090-7094. It resolves a system of difference equations approximating to the one-dimensional multigroup neutron scattering problem. More precisely, this code makes it possible to: 1. Calculate the critical condition of a reactor (keff, critical radius, critical composition) and the corresponding fluxes; 2. Calculate the associated fluxes and various subsidiary results; 3. Carry out perturbation calculations; 4. Study the propagation of fluxes at a distance; 5. Estimate the relative contributions of the cross sections (macroscopic or microscopic); 6. Study the changes with time of the composition of the reactor. (authors)
Civil engineering: calculations of pre-stressed concrete structures using CodeAster
International Nuclear Information System (INIS)
This document presents an analysis of the different calculation methods for pre-stressed concrete structure which can be performed by using finite element methods. Two methods of calculating the pre-stressing of concrete structures with finite elements have been determined. The equivalent method which consists of replacing the action of pre-stressing the concrete by equivalent forces. These method is well suited to dimensioning and studying the overall stability of a structure. It is not an easy matter to take into account the coupled or time-varying phenomena. This approach ignores the evolution of the interaction between the pre-stressing and the concrete. The explicit method which consists of including the mechanical resolution of the pre-stressed cables in that of a concrete structure. Not only does this allow a local study of the pre-stressed to be made, it also allows the coupling which developed over time to be determined, e.g. slip, deferred deformation and coupling between the steel and concrete behaviours. This method enables non-linear phenomena with varying degrees of complexity, such as fracture or yielding of the steels, drying out of the concrete, creep, etc to be described. The two methods are complementary. This document presents the mathematical and computer developments relating to each of this method. In the case of the explicit method, certain of the Code-Aster functions already make it possible to meet several EDF application requirements. Several couplings can be taken into account, such as thermomechanical, shrinkage in drying, creep, relaxation and injection of the cables. Three immediate developments of Code-Aster are proposed for the following applications: - a procedure for calculating the pre-stress losses along the pre-stressing cables; - a command to allocate these forces in the form of an initial force field in the bar elements associated with the cables; - a procedure for linking elements whose nodes do not coincide with each other
Development of a multi-group SN transport calculation code with unstructured tetrahedral meshes
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This paper reviews the computational methods used in the MUST (Multi-group Unstructured geometry SN Transport) code for solving the multi-group Sn transport equation in general geometries and describes the status of development of MUST. MUST solves the multi-group transport equation with unstructured tetrahedral meshes for modeling complicated geometrical problems. For tetrahedral mesh generation, input generation, and output visualization, we developed a management program where the mesh generation is based on Gmsh and TetGen that are open softwares. The geometrical modeling is done with the commercial CAD softwares such as CATIA. MUST uses the discontinuous finite element method (DFEM) and two-sub cell balance methods with linear discontinuous expansion (LDEM-SCB) to spatially discretize the transport equation. We applied MUST to three neutron and gamma coupled test problems for testing MUST. (author)
Energy Technology Data Exchange (ETDEWEB)
Holly R. Trellue
1998-12-01
Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.
The calculation of resonance parameters for the DeCART MOC code
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Accurate resonance parameters can be as important as the multi-group neutron cross sections themselves in the overall accuracy of a multigroup library. The work here describes the generation of resonance parameters for the MOC DeCART which utilizes the subgroup method for its resonance treatment. In this paper, we first introduce a procedure for determining the intermediate resonance parameters for all scattering isotopes, also know as lambda parameters or Goldstein-Cohen parameters which are used in the subgroup method. The lambda factors of scattering isotopes are determined as hydrogen equivalence factors by comparing group average cross sections in the mixture of resonance isotope and hydrogen with the average cross sections in mixtures with the hydrogen which is partly replaced by other isotopes. The NJOY code is used for the calculation of spectra in these mixtures. In addition to U-238, which was used as resonance isotope in previous work on lambda factors, U-235 is also treated as a resonance isotope in the lambda calculation developed here which thus provides lambdas for the groups in which U-238 has no significant resonance. After developing a procedure for generating lambda factors for scattering isotopes, a method is then described for generating subgroup parameters. Again NJOY is used for resonance calculations of a set of mixtures for each resonance isotope at each selected temperature. The group average cross sections instead of the resonance integrals of these mixtures are used to generate subgroup parameters using an optimization algorithm. The generated library is then verified by comparing the solution from DeCART with the solution from MCNP. (authors)
International Nuclear Information System (INIS)
In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.)
Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes
International Nuclear Information System (INIS)
The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR-R1 reactor are calculated. The idea is to obtain the systematic error for k∞ for this methodology comparing the calculated and the experimental results
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PUCHOK BM-DF code is described which is designated for local thermal-hudraulic parameters calculations in rod clusters with large rod numbers (300 and more) using the cell method. In this code iterations are applied for the convective crossflow mixing calculation. For the turbulent heat transfer description several empirical correlations are employed the choice among which can be done by the user. On the basis of the calculated cell thermohydraulic parameters numerous experimental data on the heat transfer crisis in various test facilities have been analyzed including those with high length. A good agreement between the calculated and experimental data allows to recommend with a high degree of reliability the cell method realized in RUCHOK BM-DF code for the fuer rod performance analysis in clusters of high length
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)
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)
The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium-plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 (n,γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX.
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,
A computer code to calculate the fast induced signals by electron swarms in gases
Energy Technology Data Exchange (ETDEWEB)
Tobias, Carmen C.B. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Mangiarotti, Alessio [Universidade de Coimbra (Portugal). Dept. de Fisica. Lab. de Instrumentacao e Fisica Experimental de Particulas
2010-07-01
Full text: The study of electron transport parameters (i.e. drift velocity, diffusion coefficients and first Townsend coefficient) in gases is very important in several areas of applied nuclear science. For example, they are a relevant input to the design of particle detector employing micro-structures (MSGC's, micromegas, GEM's) and RPC's (resistive plate chambers). Moreover, if the data are accurate and complete enough, they can be used to derive a set of electron impact cross-sections with their energy dependence, that are a key ingredient in micro-dosimetry calculations. Despite the fundamental need of such data and the long age of the field, the gases of possible interest are so many and the effort of obtaining good quality data so time demanding, that an important contribution can still be made. As an example, electrons drift velocity at moderate field strengths (up to 50 Td) in pure Isobutane (a tissue equivalent gas) has been measured only recently by the IPEN-LIP collaboration using a dedicated setup. The transport parameters are derived from the recorded electric pulse induced by a swarm started with a pulsed laser shining on the cathode. To aid the data analysis, a special code has been developed to calculate the induced pulse by solving the electrons continuity equation including growth, drift and diffusion. A realistic profile of the initial laser beam is taken into account as well as the boundary conditions at the cathode and anode. The approach is either semi-analytic, based on the expression derived by P. H. Purdie and J. Fletcher, or fully numerical, using a finite difference scheme improved over the one introduced by J. de Urquijo et al. The agreement between the two will be demonstrated under typical conditions for the mentioned experimental setup. A brief discussion on the stability of the finite difference scheme will be given. The new finite difference scheme allows a detailed investigation of the importance of back diffusion to
Validation of the ROVER-F code for ROP trip probability calculations
International Nuclear Information System (INIS)
An important task in the operation of CANDU reactors is the prevention of fuel damage as a result of fuel dryout that can occur when the fuel sheath temperature exceeds the temperature at which the coolant can efficiently remove heat. The power at which fuel dryout is expected to occur is called the critical channel power and is a function of flux shape and the fuel channel thermalhydraulics. In CANDU reactors, protection against overpowers large enough to cause dryout is provided by two regional overpower protection (ROP) systems of in-core flux detectors, arrayed through the core, each organized into three safety (or logic) channels. Each of the two independent ROP systems is associated with one of the two independent shutdown systems (SDS-1 and SDS-2). The detectors in one ROP system (associated with SDS-1) are placed in vertical penetrations, whereas the other system (associated with SDS-2) uses detectors in horizontal penetrations in the core. Each ROP system is capable of initiating the shutdown of the reactor by actuating the corresponding shutdown system. Each ROP system must be so designed that in each safety channel at l east one detector will reach its setpoint before there is damaging overpower in any fuel channel. The trip of a single detector in a safety channel will trip that channel, and the trip of two of the three safety channels in an ROP system will trip that ROP system. ROVER-F is a FORTRAN program which calculates the trip probability and the setpoint adjustment required to attain the target trip probability, for a given set of flux shapes. This calculation is performed with the assumption that the most effective safety channel is unavailable and therefore the remaining two safety channels must both trip. The calculation of trip probability itself is non-iterative, but once the trip probability of the specified system has been calculated, a convergence iteration using a binomial search is used to determine the adjustment to the trip setpoints
Truchet, G.; Leconte, P.; Peneliau, Y.; Santamarina, A.; Malvagi, F.
2014-06-01
Pile-oscillation experiments are performed in the MINERVE reactor at the CEA Cadarache to improve nuclear data accuracy. In order to precisely calculate small reactivity variations (TRIPOLI-4® by using the eigenvalue difference method. This "direct" method has shown limitations in the evaluation of very small reactivity effects because it needs to reach a very small variance associated to the reactivity in both states. To answer this problem, 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 effciency 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. Other applications of this perturbation method are presented and tested like the calculation of exact kinetic parameters (βeff, Λeff) or sensitivity parameters.
Khedr Ahmed; Adorni Martina; d’Auria Francesco
2005-01-01
The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the si...
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his report presents the description of the computer code REACT/THERMIX, which was developed for calculations of the graphite corrosion phenomena and accident transients in gas cooled High Temperature Reactors (HTR) under air and/or water ingress accident conditions. The two-dimensional code is characterized by direct coupling of thermodynamic, fluiddynamic and chemical processes with a separate handling of heterogeneous chemical reactions. (orig.)
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RAPTA-5 code used for licensing calculations to validate the compliance with the requirements for WWER fuel safety in design basis accidents. The characteristic results are given of design modelling experiments simulating thermomechanical and corrosion behaviour of WWER and PWR fuel rods in LOCA. The results corroborate the adequate predictability of both individual design models and the code as a whole. (author). 14 refs, 12 figs
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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)
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A computer code OSCAR, operated on a main frame computer was developed for the calculation of the yield of radioisotopes produced by charged-particle induced nuclear reactions. The excitation functions required for calculating the yield were evaluated by means of an empirical rule which we developed on the basis of a systematics derived from a number of experimental data reported in the literature. The rule is valid for light ion (Z ≤ 2)-induced reactions followed by neutron emission processes. Other excitation functions are also obtainable from the data file in OSCAR. In addition, the code possesses functions useful for the calculation of the stopping power and range. The energy loss and the distribution of recoil products in stacked targets are also provided as options. The formalism, structure, and direction for the usage of the code are described together with the explanation of the functions of some routines. (author)
Owen, Albert K.
1987-01-01
A computer code was written which utilizes ray tracing techniques to predict the changes in position and geometry of a laser Doppler velocimeter probe volume resulting from refraction effects. The code predicts the position change, changes in beam crossing angle, and the amount of uncrossing that occur when the beams traverse a region with a changed index of refraction, such as a glass window. The code calculates the changes for flat plate, cylinder, general axisymmetric and general surface windows and is currently operational on a VAX 8600 computer system.
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Full text: We have briefly reported the activities in the study of atomic structure and dynamics calculations using the GRASP (General purpose Atomic Structure Program) family codes and also of some activities in NIFS. We have introduced the following items: 1. Analysis of Visible M1 Lines in Tungsten Ions, 2. Collisional-radiative model for W ions, 3. Code development for single electron capture by H nucleus from metal surface, 3. Kα radiation from low charge chlorine heated by an ion beam for plasma diagnostics, 4. Code Availability. And we have summarized the talk. The visible lights emitted from highly charged tungsten ions are of special interest; they are mainly realized by magnetic dipole transitions between the fine structure levels of ions and therefore they suffer less self-absorption by surrounding plasmas providing us with a great advantage for the diagnostics of core plasmas. We have studied the source of these visible line emissions in terms of accurate non-empirical atomic structure calculations and of population kinematics analyses. The GRASP package for multi-configuration Dirac-Fock atomic structure calculation have provided us with the transition energies that are accurate enough to discriminate the states and the charges of tungsten ions. We tried, further, to reproduce the emission spectra by modifying a population kinetics code that has been used for the analysis of boron like ion plasmas. We have assigned new visible emission lines. We have shown that the use of the GRASP family of codes is quite effective for the spectroscopy of the visible line emissions of tungsten ions and for the diagnostics of the MCF plasmas. We have extended a CR model calculation code to include the case of tungsten ions plasmas; the atomic data have been calculated using HULLAC code. We have applied the code to the analysis of W35+ to W37+ plasmas. The code is now still under development. We have now under the development of a code for single electron capture by H
SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR
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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)
Wall touching kink mode calculations with the M3D code
Breslau, J. A.
2014-10-01
In recent years there have been a number of results published concerning the transient vessel currents and forces occurring during a tokamak VDE, as predicted by simulations with the nonlinear MHD code M3D. The nature of the simulations is such that these currents and forces occur at the boundary of the computational domain, making the proper choice of boundary conditions critical to the reliability of the results. The M3D boundary condition includes the prescription that the normal component of the velocity vanish at the wall. It has been argued that this prescription invalidates the calculations because it would seem to rule out the possibility of advection of plasma surface currents into the wall. This claim has been tested by applying M3D to an idealized case - a kink-unstable plasma column - in order to abstract the essential physics from the complications involved in the attempt to model real devices. While comparison of the results is complicated by effects arising from the higher dimensionality and complexity of M3D, we have verified that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the ``Hiro'' currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.
Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code
Energy Technology Data Exchange (ETDEWEB)
Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)
2013-09-15
Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)
Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code
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A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (keff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of keff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of keff from MCNP5 were obtained when the control fuel rods were fully inserted, the keff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370±0.00054. When the control fuel rods were fully withdrawn the value of keff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)
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A flexible and interactive code, NEPTUN, has been written in FORTRAN IV for the PDP-10 computer to assess the impact on man of radionuclides in aquatic food chains. NEPTUN is based on an equilibrium model of the linear-chain type, and calculates aquatic food concentrations and doses to man. A decay term is included for the holdup time of the various food types. A total of seven food types can be selected, which include drinking water, freshwater and salt-water plants, inverebrates and fish. Thirty different diets can be implemented and five different dose factor files can be chosen. These include dose conversion factors for infants and adults based on ICRP 2 and ICRP 26 methodologies. All dose factors involve a dose commitment of 50 years, or equivalently, 50 years of chronic exposure. To date, only stochastic ICRP 26 dose caluclations have been implemented. The basic concentration factor file contains data for 211 different radionuclides; the dose factor files are less comprehensive. However, all files can be readily expanded. The output includes tables of concentrations and doses for individual radionuclides, as well as summaries for groups of radionuclides. Existing aquatic food chain models and the sources of currently-used generic concentration factors are briefly reviewed, and dose factors based on ICRP 2 and ICRP 26 methodologies are contrasted. (auth)
Initial core calculation of 1 MW reactor TRIGA PUSPATI (RTP) using SRAC code system
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The 1 MWatt TRIGA PUSPATI Reactor (RTP) was located in Malaysian Institute for Nuclear Technology Research (MINT). This research reactor was from TRIGA MARK II type and was put into operation on 1983 and has reached its first criticality on 28 June 1982. Since then, this reactor has been used for various beam experiments, irradiation facilities, radioisotope production and education and training. The RTP uses three types of fuel elements, namely, 8.5wt%, 12wt% and 20wt% which enriched to about 20% of U-235 for all types. The RTP has four control rods which made up of boron carbide. It has cylindrical core but not in periodically in its lattice structure, which possibly locates 127 of fuel elements. Both of the coolant and moderator uses light water system and the reflector was made from high purity graphite. Because of this research reactor's power is relatively small compared to the power reactor; it uses natural convection for its cooling system. To ensure the integrity of the core, fuel shuffling have been made for several times. Until now, there are 11 configurations of the core and recently has achieved the 12th configuration. This paper will described the first core configuration calculation using SRAC code system which was first introduced in 2005 during the FNCA workshop. (author)
The Plasma Simulation Code: A modern particle-in-cell code with patch-based load-balancing
Germaschewski, Kai; Fox, William; Abbott, Stephen; Ahmadi, Narges; Maynard, Kristofor; Wang, Liang; Ruhl, Hartmut; Bhattacharjee, Amitava
2016-08-01
This work describes the Plasma Simulation Code (PSC), an explicit, electromagnetic particle-in-cell code with support for different order particle shape functions. We review the basic components of the particle-in-cell method as well as the computational architecture of the PSC code that allows support for modular algorithms and data structure in the code. We then describe and analyze in detail a distinguishing feature of PSC: patch-based load balancing using space-filling curves which is shown to lead to major efficiency gains over unbalanced methods and a previously used simpler balancing method.
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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.)
New beam-tracking simulation code using bulk-to-point calculation technique for space charge fields
Mizuno, A.
2016-02-01
A new two-dimensional beam-tracking simulation code for electron injectors using a bulk-to-point calculation technique for space charge fields was developed. The calculated space charge fields are produced not by a point charge but by a hollow cylinder that has a volume. Each tracked electron is a point charge. This bulk-to-point calculation technique for space charge fields is based on that used in the multiple beam envelope equations, which were developed by the author. The multiple beam envelope equations are a set of differential equations for investigating the beam dynamics of electron injectors and can be used to calculate bunched beam dynamics with high accuracy. However, there is one limitation. The bunched beam is assumed to be an ensemble of several segmentation pieces in both the transverse and longitudinal directions. In this bunch model, each longitudinal segmentation slice in a bunch must not warp; consequently, the accuracy of the calculated emittance is reduced in the case of a highly charged beam for calculations of a typical rf gun injector system. This limitation is related to the calculation model of longitudinal space charge fields. In the newly developed beam-tracking simulation code, the space charge field calculation scheme is upgraded and the limitation has been overcome. Therefore, the applicable range is extended while maintaining the high accuracy of emittance calculations. Simultaneously, the calculation time is markedly shortened because the emittance dependence on the segmentation number is extremely weak. In this paper, several examples of beam dynamics that cannot be calculated accurately using the multiple beam envelope equations are demonstrated using the new beam-tracking simulation code. The accuracy of the calculated emittance is also discussed.
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One of the most important unresolved issues governing risk in many nuclear power plants involves the phenomenon called direct containment heating (DCH), in which it is postulated that molten corium ejected under high pressure from the reactor vessel is dispersed into the containment atmosphere, thereby causing sufficient heating and pressurization to threaten containment integrity. Models for the calculation of potential DCH loads have been developed and incorporated into the CONTAIN code for severe accident analysis. Using CONTAIN, DCH scenarios in PWR plants having three different representative containment types have been analyzed: Surry (subatmospheric large dry containment), Sequoyah (ice condenser containment), and Bellefonte (atmospheric large dry containment). A large number of parameter variation and phenomenological uncertainty studies were performed. Response of DCH loads to these variations was found to be quite complex; often the results differ substantially from what has been previously assumed concerning DCH. Containment compartmentalization offers the potential of greatly mitigating DCH loads relative to what might be calculated using single-cell representations of containments, but the actual degree of mitigation to be expected is sensitive to many uncertainties. Dominant uncertainties include hydrogen combustion phenomena in the extreme environments produced by DCH scenarios, and factors which affect the rate of transport of DCH energy to the upper containment. In addition, DCH loads can be aggravated by rapid blowdown of the primary system, co-dispersal of moderate quantities of water with the debris, and quenching of de-entrained debris in water; these factors act by increasing steam flows which, in turn, accelerates energy transport. It may be noted that containment-threatening loads were calculated for a substantial portion of the scenarios treated for some of the plants considered
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In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor keff (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.
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The report describes a revision of the SFACTOR computer code, which has been developed to estimate the average dose equivalent to each of a specified list of target organs per microcurie-day residence of a radionuclide in source organs in man. Source and target organs of interest are specified in the input data stream, along with nuclear decay information. The SFACTOR code computes components of dose equivalent rate from each type of decay present for a particular radionuclide, including alpha, electron, gamma radiation, and spontaneous fission. The principal refinement to the program is the addition of a method for calculating components of the dose equivalent rate from alpha particles to endosteal cells and red bone marrow from a source in mineral bone. Other details of the calculations remain unchanged. Corrected tabulations of all components of S are provided for an array of 22 source organs and 24 target organs for 19 radionuclides in an adult
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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
TRANS4: a computer code calculation of solid fuel penetration of a concrete barrier
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The computer code, TRANS4, models the melting and penetration of a solid barrier by a solid disc of fuel following a core disruptive accident. This computer code has been used to model fuel debris penetration of basalt, limestone concrete, basaltic concrete, and magnetite concrete. Sensitivity studies were performed to assess the importance of various properties on the rate of penetration. Comparisons were made with results from the GROWS II code
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In this paper the uncertainties of the neutronic calculations at core level - originating from the uncertainties of the basic nuclear data - are presented. The investigations have been made for a VVER-1000 core (Kozloduy-6) defined in the frame of the OECD NEA UAM benchmark. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution and the rod worth are shown. After that the preliminary evaluation of the uncertainties of the neutron kinetic calculations are presented for a rod movement transient at HZP (Hot Zero Power) state, where the uncertainties of the time dependent core and assembly powers and the dynamic reactivity were evaluated. In both cases, we will see that the most important quantities - at core level and at HZP state - have a considerable uncertainty which is originating from the uncertainties of the basic cross section library in these investigations. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Panka, Istvan; Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.
2013-09-15
In this paper the uncertainties of the neutronic calculations at core level - originating from the uncertainties of the basic nuclear data - are presented. The investigations have been made for a VVER-1000 core (Kozloduy-6) defined in the frame of the OECD NEA UAM benchmark. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution and the rod worth are shown. After that the preliminary evaluation of the uncertainties of the neutron kinetic calculations are presented for a rod movement transient at HZP (Hot Zero Power) state, where the uncertainties of the time dependent core and assembly powers and the dynamic reactivity were evaluated. In both cases, we will see that the most important quantities - at core level and at HZP state - have a considerable uncertainty which is originating from the uncertainties of the basic cross section library in these investigations. (orig.)
International Nuclear Information System (INIS)
Full text of publication follows: The QUENCH fuel bundle experiments, performed at the Forschungszentrum Karlsruhe in Germany, aim to investigate the hydrogen source term and the bundle degradation during reflood of an overheated reactor core. The test QUENCH-07, in which the bundle was cooled from high temperatures by steam injected from the bottom, was the first experiment in this test series with a boron carbide absorber rod in the bundle. One major objective of this test was to provide information on the B4C/SS/Zry interactions, on the formation of gaseous reaction products during B4C oxidation and control rod degradation, and on the impact of control rod degradation on surrounding rods. In the general frame of developmental assessment and code validation, a post-test calculation of test QUENCH-07 complemented by an uncertainty analysis was performed with the code ATHLET-CD. The system code ATHLET-CD is being developed for realistic simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed and validated models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation, including mechanical rod behaviour, zirconium and B4C oxidation, melting and relocation of metallic and ceramic components, and for the release and transport of fission products and aerosols. The first step of the work was the simulation of the QUENCH-07 experiment, applying the modeling options recommended in the code User's Manual (reference calculation). The global results of this calculation, mainly with respect to the hydrogen release rate and to the time evolution of bundle temperatures in different elevations, showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed in the experiment. For this
Load-balancing techniques for a parallel electromagnetic particle-in-cell code
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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.
International Nuclear Information System (INIS)
The Institute for Reactor Safety (IRS) of the Research Centre of Karlsruhe (FZK) is involved in the qualification and further development of design and safety program systems for current LWR and innovative reactors. These investigations are focused not only on the study of the behavior of the whole nuclear power plant by means of system codes but also on the core behavior by means of detailed coupled neutronic and thermal hydraulic solutions. To improve the prediction of the safety margins based on local parameters the coupling of detailed neutronic (transport or Monte Carlo) and thermal hydraulic (subchannel or CFD codes) at the fuel pin level is required. At IRS/FZK a coupling on the sub channel code COBRA-TF with the neutron transport code THREEDANT has been realized within the Karlsruhe modular reactor calculation system KAPROS-E. This coupling scheme has been applied to determine the pin power of a German PWR fuel assembly obtaining promising results. In addition, this coupling scheme is being qualified by performing Monte Carlo calculations for selected fuel assembly conditions. In this contribution, the approach followed to couple 3D transport code THREEDANT with the thermal hydraulic subchannel code COBRA -TF as well as selected results for a PWR fuel assembly will be presented. In addition the qualification work using Monte Carlo solutions for the pin-by-pin simulation of the PWR fuel assembly will be given and discussed. (authors)
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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)
Verification of the CFD code FLUENT by post test calculation of ROCOM experiments
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Full text of publication follows: The TUV NORD e.V. is an independent Technical Support Organisation (TSO) performing safety assessments in almost every field of technology. In nuclear safety the TUV can look back on more than 40 years of experience. In the last years in Germany PWR safety analyses were focussed on boron dilution events with the potential of reactivity transients. The possibility of coolant with a low boron concentration collected in localized areas of the reactor coolant system (RCS) can be caused by injection of coolant with less boron content from interfacing systems (external dilution) as well as separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution). Inherent dilution can e.g. occur after reflux-condenser heat transfer after a small break loss of coolant accident (SBLOCA) with a limited operability of the emergency core cooling (ECC) systems. The TUV Nord e.V. was charged by German supervisory authorities with the assessment of the safety analyses presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The simulation of boron dilution and transport processes in PWR reactor coolant systems (RCS) and especially reactor pressure vessels (RPV) requires the application of computational fluid dynamic (CFD) codes. At present the validation of these codes is performed by post test calculations of boron dilution experiments e.g. Rossendorf Coolant Mixing Model (ROCOM). They were chosen by TUV Nord e.V. for validation of FLUENT, because of the excellent experimental data base, especially the high spatial and temporal resolution measurements of boron concentration distribution with wire mesh sensors. The ROCOM facility was built at the Forschungszentrum Rossendorf e.V. near Dresden in linear scale of 1:5 for the investigation of coolant mixing in a wide range of PWR flow conditions. ROCOM is a
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These scientific days were organised by the 'technical protection' Section of the French Society of Radiation Protection (SFRP) in cooperation with the French society of medical physicists (SFPM), the Swiss Romandie association of radioprotection (ARRAD) and the associated laboratories of radio-physics and dosimetry (LARD). The objective of these days was to review the existing calculation codes used in radiation transport, source estimation and dose management, and to identify some future prospects. This document brings together the available presentations (slides) together with their corresponding abstracts (in French) and dealing with: 1 - Presentation of the conference days (L. De Carlan); 2 - Simulating radionuclide transfers in the environment: what calculation codes and for what? (C. Mourlon); 3 - Contribution of Monte-Carlo calculation to the theoretical foundation analysis of calibration procedures and dosemeters design for radioprotection photon dosimetry (J.M. Bordy); 4 - Use of calculation codes in R and D for the development of a new passive dosemeter for photons and beta radiations (B. Moreno); 5 - Development of a new virtual sources model for the Monte-Carlo prediction of EPID (Electronic Portal Imaging Device) images and implementation in PENELOPE (I. Chabert); 6 - Prediction of high-resolution EPID images for in-vivo dosimetry (D. Patin); 7 - 4D thorax modeling by artificial neural networks (P.E. Leni); 8 - Presentation of the calculation utilities of the book 'Calculation of ionizing radiations generated doses' (Vivier, Lopez, EDP Sciences 2012) (A. Vivier); 9 - RayXpertC: a 3D modeling and Monte-Carlo dose rate calculation software (C. Dossat); 10 - TRIPOLI-4R Version 9 S Monte-Carlo code for radioprotection (F. Damian); 11 - Realistic radioprotection training with the digital school workshop (E. Courageot); 12 - Use of BEAMNRC code for dental prostheses influence evaluation in ENT cancers treatment by external radiotherapy (C. De Conto); 13
DCHAIN 2: a computer code for calculation of transmutation of nuclides
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DCHAIN2 is a one-point depletion code which solves the coupled equation of radioactive growth and decay for a large number of nuclides by the Bateman method. A library of nuclear data for 1170 fission products has been prepared for providing input data to this code. The Bateman method surpasses the matrix exponential method in computational accuracies and in saving computer storage for the code. However, most existing computer codes based on the Bateman method have shown serious drawbacks in treating cyclic chains and more than a few specific types of decay chains. The present code has surmounted the above drawbacks by improving the code FP-S, and has the following characteristics: (1) The code can treat any type of transmutation through decays or neutron induced reactions. Multiple decays and reactions are allowed for a nuclide. (2) Unknown decay energy in the nuclear data library can be estimated. (3) The code constructs the decay scheme of each nuclide in the code and breaks it up into linear chains. Nuclide names, decay types and branching ratios of mother nuclides are necessary as the input data for each nuclide. Order of nuclides in the library is arbitrary because each nuclide is destinguished by its nuclide name. (4) The code can treat cyclic chains by an approximation. A library of the nuclear data has been prepared for 1170 fission products, including the data for half-lives, decay schemes, neutron absorption cross sections, fission yields, and disintegration energies. While DCHAIN2 is used to compute the compositions, radioactivity and decay heat of fission products, the gamma-ray spectrum of fission products can be computed also by a separate code FPGAM using the composition obtained from DCHAIN2. (J.P.N.)
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The objective of the study is to compare the thermal neutron fluxes at specimen positions of neutron radiography facility calculated by MCNP4C code with the measurement. A model for calculation was developed using details of the reactor core configuration no. 14 and neutron radiography facility installed at the existing research reactor, TRR-1/M1 reactor. Assuming all fresh fuel elements and all control rod out condition, the thermal neutron fluxes at various specimen positions were calculated using MCNP4C code. The calculation are verified by the measurement using foil activation method. Generally, the calculated neutron fluxes are overestimated by 16-20% which is reasonably good agreement and acceptable for the complex system. The discrepancy is expected to the assumption of using fresh fuel elements, all control rod out condition, and also lacks of information in develop a more accurate model for calculation. This study shows the possibility of using the MCNP4C code to verify the thermal neutron fluxes at specimen position and shielding design of the new neutron radiography facility at the new Thai research reactor
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At present at the Ignalina NPP the process of extensive use of the new uranium - erbium fuel is going on. The loading process of the new uranium - erbium fuel assemblies into the reactor cores is accompanied by experiments and analytical investigation of the behaviour of the main neutron - physical characteristics of the reactor. This article presents the results of independent calculations of different core states obtained using the German code QUABOX/CUBBOX which are compared with the results of similar calculations performed using codes SADCO and STEPAN as well as with experimental data. This is one of the steps in the process of validation of the code QUABOX/CUBBOX for the modeling of the process taking place in RBMK-1500 reactors. (author)
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A numerical analysis of some neutronic parameters calculated by LEOPARD computer code compared with the literature data are presented. A computer code (LEOCIT) that is a modified version of LEOPARD, was developed, with subroutines that prepare cross sections libraries for 1,2 or 4 energy groups, writing them on tape or on disk, in special format aiming to be diretly used by citation computer codes. Finally, a simulation of the first cycle of Angra I burnup, is done, by CITATION, modelling 1/4 of the core in XY geometry, calculation, the soluble boron curve and the pin to pin power distribution, for two energy group. The more relevant results are compared with those supplied by Westinghouse, CNEN and FURNAS, and some recommendations aiming to perfect the developed system, are done. (E.G)
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Comparisons of the calculation of the differential pressure through four MOV's between Flowmaster code and SFM model has been presented. The Flowmaster and SFM model basically use 1-D steady-state equation, but for the transient analysis, the Flowmaster uses Joukowsky equation considering the effect of fluid velocity variation and wave speed, while, the SFM model uses quasi-steady equation including fluid inertia effect due to pipe inertia. The maximum differential pressures in opening stroke are almost the same between Flowmaster and SFM model, because the two code have the same steady-state equation. For closing stroke, however, the maximum differential pressure is somewhat different, the Flowmaster code shows higher large estimation than SFM code
Energy Technology Data Exchange (ETDEWEB)
Rey, H. K.; Park, S. K.; Kim, D. W.; Kang, S. C.; Jung, H. K.; Park, S. K. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
2000-10-01
Comparisons of the calculation of the differential pressure through four MOV's between Flowmaster code and SFM model has been presented. The Flowmaster and SFM model basically use 1-D steady-state equation, but for the transient analysis, the Flowmaster uses Joukowsky equation considering the effect of fluid velocity variation and wave speed, while, the SFM model uses quasi-steady equation including fluid inertia effect due to pipe inertia. The maximum differential pressures in opening stroke are almost the same between Flowmaster and SFM model, because the two code have the same steady-state equation. For closing stroke, however, the maximum differential pressure is somewhat different, the Flowmaster code shows higher large estimation than SFM code.
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The aim of this contribution is development of methodology for verification of selected input calculation data (performance unit parameters, work group structure, and duration of time-dependent activities) of the OMEGA Code in the individual PSL (Proposed Standardised List) structure parts. (author)
International Nuclear Information System (INIS)
In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting γ ray and β ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted γ ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library 'JNDC Nuclear Data Library of Fission Products - second version -', which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author)
Calculation of the Fast Flux Test Facility fuel pin tests with the WIMS-E and MCNP codes
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The Fuel Assembly Area (FAA) at the Fast Flux Test Facility site on the Hanford Site at Richland, Washington currently is being prepared to fabricate mixed oxide fuel (U, Pu) for the FFTF. Calculational tools are required to perform criticality safety analyses for various process locations and to establish safe limits for fissile material handling at the FAA. These codes require validation against experimental data appropriate for the compositions that will be handled. Critical array experiments performed by Bierman provide such data for mixed oxide fuel in the range Pu/(U+Pu) = 22 wt %, and with Pu-240 contents equal to 12 wt %. Both the Monte Carlo Neutron Photon (MCNP) and the Winfrith Improved Multigroup Scheme (WIMS-E) computer codes were used to calculate the neutron multiplication factor for explicit models of the various critical arrays. The W-CACTUS modules within the WIMS-E code system was used to calculate k∞ for the explicit array configuration, as well as few-group cross sections that were then used in a three-dimensional diffusion theory code for the calculation of keff for the finite array. 10 refs., 15 figs., 7 tabs
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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
LWR-WIMS, a computer code for light water reactor lattice calculations
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LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)
Electron and ion cyclotron heating calculations in the tandem-mirror modeling code MERTH
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To better understand and predict tandem-mirror experiments, we are building a comprehensive Mirror Equilibrium Radial Transport and Heating (MERTH) code. In this paper we first describe our method for developing the code. Then we report our plans for the installation of physics packages for electron- and ion-cyclotron heating of the plasma
Investigation of methods used in calculations of solar cell parameters
Shvets, E. Ya.; Khrypko, S. L.; Zubko, E. I.
2009-01-01
Analytical expressions have been obtained for extracting the electrical parameters and characteristics of solar cells, including series and shunt resistances, and the saturation current. The method of Lagrange multipliers was used for computing the shape factor of the current–voltage characteristic (CVC) of solar cell. The calculation results demonstrated a satisfactory agreement with experimental data.
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The RTP is a light-water moderated and pool-type TRIGA MARK II reactor with power capacity of 1MWt. It was built in 1979 and attained the first criticality on 28 June 1982. The RTP was designed mainly for neutron activation analysis, small angle neutron scattering, neutron radiography, radioisotope production, education and training purposes. It uses standard TRIGA fuel developed by General Atomic in which the zirconium hydride moderator is homogeneously combined with enriched uranium. It has a cylindrical core with which possibility of locating 127 of fuel elements. Both of the coolant and moderator uses light water system and the reflector is made of high purity graphite. Because of its relatively small power, it uses natural convection for its cooling system. To ensure the integrity of the core, fuel shuffling have been carried out several times. Until now, there were 12 configurations of the core, the most recent change being in July 2006. This paper will describe the RTP core calculation using the Monte Carlo MVP code system. VP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation in order to have an accurate and fast Monte Carlo simulation of neutron and photon transport problems. The MVP Monte Carlo code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique. When compared to the conventional scalar method, this code could achieve higher computation speed by up to a factor of 10 on the vector super-computer. The RTP core has been modelled using cylinder geometry along the z-coordinate geometry with the MVP code system while its material cross section data is calculated beforehand. The JENDL3.3 data library was used in the whole calculation. The objectives of the calculation are to calculate the multiplication factor values (keff), fission density and flux distribution from the tally data. The calculation also
Comparison of calculations made with three two-dependent neutron codes TDA, MORSE and POW
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Three computer codes were compared to determine their usefulness in analysing a pulsed neutron experiment. The codes were a Monte Carlo code (MORSE) a diffusion kinetics code (POW) and a time dependent SN code (TDA). A series of test problems were devised to progressively model the experiment. All problems assumed a spherical system with an isotropic source. The first problem had its source in the first energy group for the first nano-second. The second problem had its source distributed in time but not distributed in energy. The third problem had a source distributed in energy and time. POW and MORSE were shown to be in good agreement, with significant differences occurring only at times when the system did not correspond with the approximations made in POW. The AAEC version of the TDA code did not handle a time-dependent source. There was also a tendency for the results beyond 50 ns to be higher than those for the other two codes for the problems having a source constant over one time interval. (author)
Computer code TOBUNRAD for PWR fuel bundle heat-up calculations
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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)
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Calculation model of KLT-40S reactor concerning stationary and transition regimes of the reactor with the use of heat-hydraulic code SERPENT has been developed. The model validation was carried out by comparison of calculation results of stationary regime at the rated power level obtained with the reactor developers' data. Calculation analysis of transition regimes was performed. These transition regimes were: shutdown 2 of 4 steam generator's sections; shutdown 2 of 4 circulation pumps of first circuit. The fact that position of control rods was invariable must be in parenthesis. Alterations of power, heat carrier temperatures, fuel and dispersive fuel composition cans were specified
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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
Directory of Open Access Journals (Sweden)
Skrzypek Maciej
2015-09-01
Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.
Energy Technology Data Exchange (ETDEWEB)
Endo, Akira; Kim, Eunjoo; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-10-01
A Monte Carlo code SCINFUL has been utilized for calculating response functions of organic scintillators for high-energy neutron spectroscopy. However, the applicability of SCINFUL is limited to the calculations for cylindrical NE213 and NE110 scintillators. In the present study, SCINFUL-CG was developed by introducing a geometry specifying function and high-energy neutron cross section data into SCINFUL. The geometry package MARS-CG, the extended version of the CG (Combinatorial Geometry), was programmed into SCINFUL-CG to express various geometries of detectors. Neutron spectra in the regions specified by the CG can be evaluated by the track length estimator. The cross section data of silicon, oxygen and aluminum for neutron transport calculation were incorporated up to 100 MeV using the data of LA150 library. Validity of SCINFUL-CG was examined by comparing calculated results with those by SCINFUL and MCNP and experimental data measured using high-energy neutron fields. SCINFUL-CG can be used for the calculations of the response functions and neutron spectra in the organic scintillators in various shapes. The computer code will be applicable to the designs of high-energy neutron spectrometers and neutron monitors using the organic scintillators. The present report describes the new features of SCINFUL-CG and explains how to use the code. (author)
Development of a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport
Jia, Xun; Sempau, Josep; Choi, Dongju; Majumdar, Amitava; Jiang, Steve B
2009-01-01
Monte Carlo simulation is the most accurate method for absorbed dose calculations in radiotherapy. Its efficiency still requires improvement for routine clinical applications, especially for online adaptive radiotherapy. In this paper, we report our recent development on a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport. We have implemented the Dose Planning Method (DPM) Monte Carlo dose calculation package (Sempau et al, Phys. Med. Biol., 45(2000)2263-2291) on GPU architecture under CUDA platform. The implementation has been tested with respect to the original sequential DPM code on CPU in two cases. Our results demonstrate the adequate accuracy of the GPU implementation for both electron and photon beams in radiotherapy energy range. A speed up factor of 4.5 and 5.5 times have been observed for electron and photon testing cases, respectively, using an NVIDIA Tesla C1060 GPU card against a 2.27GHz Intel Xeon CPU processor .
International Nuclear Information System (INIS)
The code PERTURB.D computes the thermal neutron flux perturbation factor, K, due to circular foils located in an isotropic neutron field. The calculation is based on the expression K = G.E.F, where G denotes the neutron self-shielding in the foil, E the edge correction factor and F the flux depression in the diffusing medium surrounding the foil. By comparison with published experimental results is was found that, for 66% (85%) of the sample cases, calculated K-values agree to better than 1% (2%) with the experimental ones. As an application of the code PERTURB.D, tables of K-values for different materials, including Mn, Co, In and Au, for diameters in the range 0.5 to 2 cm, and various practical thicknesses, both in water and graphite, are calculated and presented. (Auth.)
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. The neutronic calculation has been carried out for TRIGA 2 MW reactor. These included criticality flux and power distributions. Computer code Citation which solves 7-groups, 3-dimensional hexagonal geometry has been used. The multi groups-cross-section is generated by the WIMS-D/4 code.This 7-group-39x39x38-mesh-points problem takes about 90 minutes on the Pentium-133 MHz PC. The calculation of the initial core of TRIGA 2 MW reactor shows that the excess reactivity of the core is 7,8% and the thermal fluxes in the irradiation positions are between 1.0-2.9*1013n cm-2s-1. The results are about 10% deviate from those calculated by General Atomics. In the initial core, the highest power is produced in the C-9 position. The fuel element in this position produces 30.7 k W thermal power
DRAGON 3.05D, Reactor Cell Calculation System with Burnup
International Nuclear Information System (INIS)
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
The MARS15-based FermiCORD code system for calculation of the accelerator-induced residual dose
Grebe, A; Lu, T; Mokhov, N; Pronskikh, V
2016-01-01
The FermiCORD code system, a set of codes based on MARS15 that calculates the accelerator-induced residual doses at experimental facilities of arbitrary configurations, has been developed. FermiCORD is written in C++ as an add-on to Fortran-based MARS15. The FermiCORD algorithm consists of two stages: 1) simulation of residual doses on contact with the surfaces surrounding the studied location and of radionuclide inventories in the structures surrounding those locations using MARS15, and 2) simulation of the emission of the nuclear decay gamma-quanta by the residuals in the activated structures and scoring the prompt doses of these gamma-quanta at arbitrary distances from those structures. The FermiCORD code system has been benchmarked against similar algorithms based on other code systems and showed a good agreement. The code system has been applied for calculation of the residual dose of the target station for the Mu2e experiment and the results have been compared to approximate dosimetric approaches.
A thermo mechanical benchmark calculation of a hexagonal can in the BTI accident with INCA code
International Nuclear Information System (INIS)
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
Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes
International Nuclear Information System (INIS)
The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)
International Nuclear Information System (INIS)
A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for 3He, H2, or BF3 gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of 3He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the 3He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF3 counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations
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The aim of this study was to investigate the Monte Carlo (MC) code FLUKA, regarding its ability to accurately simulate electron transport at density inhomogeneities and in ionization chamber geometries. In order to evaluate the accuracy of FLUKA's electron transport algorithm and the implementation of the condensed history technique, a Fano test was used. This test allows the comparison of calculated and theoretically expected results. The ratio of the two results is ideally equal to unity, and a deviation usually indicates artifacts in the treatment of density interfaces. As a more practical problem, wall perturbation factors pwall of a plane parallel chamber in electron beams were calculated and compared with results based on the EGSnrc MC code. Additionally, the impact of wall material and thickness on calculated cavity dose was investigated for two different thimble chambers irradiated by 60Co. The correct choice of parameters within FLUKA's electron transport algorithm ensured passing the Fano test within ∼0.7% and a good agreement for practical examples within 0.4% compared to results of the EGSnrc MC code. The latter is known to allow an artifact free simulation of ionization chamber response in photon and electron beams. Based on these results, the electron transport accuracy within the FLUKA code can generally be regarded as much better than 1% for typical ionization chamber dosimetry problems. (author)
International Nuclear Information System (INIS)
At present, at the Ignalina NPP the process of a wider use of the new uranium-erbium fuel is going on. The loading process of the new uranium-erbium fuel assemblies into the reactor cores is accompanied by experiments and analytical investigation of the behaviour of the main neutron-physical characteristics of the reactors. The article presents the results of independent calculations of different core states of the reactor of the 1 unit of the Ignalina NPP, obtained using the German code QUABOX/CUBBOX. Ale data are compared with the results of similar calculations performed using the codes SADCO and STEPAN, as well as with experimental data. This is one of the further steps in the process of validation of the code QUABOX/CUBBOX. This article is a continuation of the discussion of the results of calculations performed in the process of verification of the German code QUABOX/CUBBOX for the modeling of the processes taking place in RBMK-1500 reactors. (author)
Energy Technology Data Exchange (ETDEWEB)
Takeda, N. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Kudo, K. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Toyokawa, H. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Torii, T. [Japan Power Reactor and Nuclear Fuel Development Corporation, Tsuruga Office, Fukui 919-12 (Japan); Hashimoto, M. [Japan Power Reactor and Nuclear Fuel Development Corporation, O-arai Engineering Center, Ibaraki 311-13 (Japan); Sugita, T. [Science System Laboratory, Ibaraki 309-17 (Japan); Dietze, G. [Physikalisch-Technische Bundesanstalt, 38023 Braunschweig (Germany); Yang, X. [China Institute of Atomic Energy (China)
1999-02-11
A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for {sup 3}He, H{sub 2}, or BF{sub 3} gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of {sup 3}He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the {sup 3}He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF{sub 3} counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations.
International Nuclear Information System (INIS)
The use of universal neutron transport codes in order to calculate the parameters of well-logging probes presents a new approach first tried in U.S.A. and UK in the eighties. This paper deals with first such an attempt in Poland. The work is based on the use of MORSE code developed in Oak Ridge National Laboratory in U.S.A.. Using CG MORSE code we calculated neutron detector response when surrounded with sandstone of porosities 19% and 38%. During the work it come out that it was necessary to investigate different methods of estimation of the neutron flux. The stochastic estimation method as used currently in the original MORSE code (next collision approximation) can not be used because of slow convergence of its variance. Using the analog type of estimation (calculation of the sum of track lengths inside detector) we obtained results of acceptable variance (∼ 20%) for source-detector spacing smaller than 40 cm. The influence of porosity on detector response is correctly described for detector positioned at 27 cm from the source. At the moment the variances are quite large. (author). 33 refs, 8 figs, 8 tabs
Energy Technology Data Exchange (ETDEWEB)
Habib, B.; Poumarede, B.; Tola, F.; Barthe, J. [CEA, LIST, Dept Technol Capteur et Signal, F-91191 Gif Sur Yvette, (France)
2010-07-01
The aim of the present study is to demonstrate the potential of accelerated dose calculations, using the fast Monte Carlo (MC) code referred to as PENFAST, rather than the conventional MC code PENELOPE, without losing accuracy in the computed dose. For this purpose, experimental measurements of dose distributions in homogeneous and inhomogeneous phantoms were compared with simulated results using both PENELOPE and PENFAST. The simulations and experiments were performed using a Saturne 43 linac operated at 12 MV (photons), and at 18 MeV (electrons). Pre-calculated phase space files (PSFs) were used as input data to both the PENELOPE and PENFAST dose simulations. Since depth-dose and dose profile comparisons between simulations and measurements in water were found to be in good agreement (within {+-} 1% to 1 mm), the PSF calculation is considered to have been validated. In addition, measured dose distributions were compared to simulated results in a set of clinically relevant, inhomogeneous phantoms, consisting of lung and bone heterogeneities in a water tank. In general, the PENFAST results agree to within a 1% to 1 mm difference with those produced by PENELOPE, and to within a 2% to 2 mm difference with measured values. Our study thus provides a pre-clinical validation of the PENFAST code. It also demonstrates that PENFAST provides accurate results for both photon and electron beams, equivalent to those obtained with PENELOPE. CPU time comparisons between both MC codes show that PENFAST is generally about 9-21 times faster than PENELOPE. (authors)
Calculation of behaviour of the Juragua NPP containment with code TRACOV/MOD1
International Nuclear Information System (INIS)
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
Performance of large LWR system codes in calculating the steam-generator heat-transfer behavior
International Nuclear Information System (INIS)
This paper presents a series of modeling experiences and problems in simulating the thermal-hydraulic behavior of large PWR steam generators using the RELAP4 and RELAP5 computer codes. Sensitivity studies investigating the heat transfer characteristics of both once-through and U-tube steam generators are discussed. Suggestions and recommendations are given for effective use and potential future improvements of these codes
Improvement of fitting method of multiband parameters for cell calculations
International Nuclear Information System (INIS)
To accurately perform cell calculations of nuclear reactors, a new fitting procedure has been developed for calculating multiband parameters, which are necessary for effective cross section calculations. By using the new fitting procedure, the error of multiband parameters becomes always zero. Reactor cell calculations have been performed to compare the effective cross sections and the infinite multiplication factors etc. calculated using the multiband parameters obtained by the new and the conventional fitting procedures by using the cross section set based on the JENDL-3.1 library with 107 energy groups. It is found that there is a small difference of the calculational results between the two fitting procedures and it is found from burnup calculations that the difference of the infinite multiplication factors is not dependent on the burnup period up to about 30 GWd/t. The onion skin effect can be exactly treated by dividing a fuel pellet to multiple regions and by using the multiband method. Thus the difference of burnup properties between two fitting procedures are investigated for the divided and the undivided fueled cells. The total inventory of Pu, Am etc. at the divided case is almost the same to the undivided case at the end of the burnup period. However it is found that the radial distribution of atomic density is slightly different between the two fitting procedures. (author)
International Nuclear Information System (INIS)
Shutdown dose rate (SDDR) inside and around the diagnostics ports of ITER is performed at PPPL/UCLA using the 3-D, FEM, Discrete Ordinates code, ATTILA, along with its updated FORNAX transmutation/decay gamma library. Other ITER partners assess SDDR using codes based on the Monte Carlo (MC) approach (e.g. MCNP code) for transport calculation and the radioactivity inventory code FISPACT or other equivalent decay data libraries for dose rate assessment. To reveal the range of discrepancies in the results obtained by various analysts, an extensive experimental and calculation benchmarking effort has been undertaken to validate the capability of ATTILA for dose rate assessment. On the experimental validation front, the comparison was performed using the measured data from two SDDR experiments performed at the FNG facility, Italy. Comparison was made to the experimental data and to MC results obtained by other analysts. On the calculation validation front, the ATTILA's predictions were compared to other results at key locations inside a calculation benchmark whose configuration duplicates an upper diagnostics port plug (UPP) in ITER. Both serial and parallel version of ATTILA-7.1.0 are used in the PPPL/UCLA analysis performed with FENDL-2.1/FORNAX databases. In the FNG 1st experimental, it was shown that ATTILA's dose rates are largely over estimated (by ∼30–60%) with the ANSI/ANS-6.1.1 flux-to-dose factors whereas the ICRP-74 factors give better agreement (10–20%) with the experimental data and with the MC results at all cooling times. In the 2nd experiment, there is an under estimation in SDDR calculated by both MCNP and ATTILA based on ANSI/ANS-6.1.1 for cooling times up to ∼4 days after irradiation. Thereafter, an over estimation is observed (∼5–10% with MCNP and ∼10–15% with ATTILA). As for the calculation benchmark, the agreement is much better based on ICRP-74 1996 data. The divergence among all dose rate results at ∼11 days cooling time is no
ANISN-JR, a one-dimensional discrete ordinates code for neutron and gamma-ray transport calculations
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The ANISN code available from RSIC is designed to solve the one-dimensional Boltzmann equation for deep penetration problems taking into consideration the anisotropic scattering by Legendre expansion of the scattering cross sections. To extend its applicability for shielding analyses, the code has been modified by adding options of calculating the reaction rates distributions from detector response, generating the volume-flux weighted cross sections in arbitrary regions or zones and plotting the neutron or gamma-ray spectra and the reaction rates distributions. The formats of input data necessary in the options are described in detail. (auth.)
International Nuclear Information System (INIS)
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
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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 P1-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)
A reliable Monte Carlo based investigation of ion chambers in medical physics problems depends on the accuracy of the charged particle transport and implementations of the condensed history technique. Improper handling of media interfaces can lead to anomalous results or 'interface artefacts'. This work presents a rigorous investigation of the electron transport algorithm in the general purpose Monte Carlo (MC) code FLUKA (2008.3b.1). A 'Fano test' was implemented in order to benchmark the accuracy of the algorithm. Furthermore, the calculation of wall perturbation factors pwall of a Roos type chamber irradiated by electrons were performed and compared with values based on the EGSnrc MC code
International Nuclear Information System (INIS)
A general purpose user's version of the EGS4 code system has been developed to make EGS4 easily applicable to the safety analysis of nuclear fuel cycle facilities. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with Kansas State University (KSU) photon skyshine experiment of 1977. The results of the simulation showed that this version of EGS4 would be appicable to the skyshine calculation. (author)
International Nuclear Information System (INIS)
The Intergovernmental Maritime Consultative Organization IMCO) is currently preparing guidelines concerning the safety of nuclear-powered merchant ships. An important aspect of these guidelines is the determination of the releases of radioactive material in effluents from these ships and the control exercised by the ships over these releases. To provide a method for the determination of these releases, the NRC staff has developed a computerized model, the NMS-GEFF Code, which is described in the following chapters. The NMS-GEFF Code calculates releases of radioactive material in gaseous effluents for nuclear-powered merchant ships using pressurized water reactors
International Nuclear Information System (INIS)
The linear integral-equation-based computer code 'Roger Oleg Nikolai' (RON), which was recently developed at Argonne National Laboratory, was used to calculate the self-amplified spontaneous emission (SASE) performance of the free-electron laser (FEL) being built at Argonne. Signal growth calculations under different conditions were used to estimate tolerances of actual design parameters and to estimate optimal length of the break sections between undulator segments. Explicit calculation of the radiation field was added recently. The measured magnetic fields of five undulators were used to calculate the gain for the Argonne FEL. The result indicates that the real undulators for the Argonne FEL (the effect of magnetic field errors alone) will not significantly degrade the FEL performance. The capability to calculate the small-signal gain for an FEL-oscillator is also demonstrated
International Nuclear Information System (INIS)
The fluence to effective-dose and organ-absorbed-dose conversion coefficients for charged pions and muons were calculated based on the instructions given in ICRP Publication 103. For the calculation, the particle motions in the ICRP/ICRU adult reference computational phantoms were simulated using the PHITS code for four idealized irradiation geometries as well as those closely representing the geometrical simulations of cosmic-ray muon exposure. Cosmic-ray pion and muon dose rates over a wide altitude range were estimated using the calculated dose conversion coefficients. The results of the calculation indicate that the assumption of the isotropic irradiation geometry is suitable to be utilized in the dose estimations for cosmic-ray pions and muons. It is also found from the calculation that the introduction of ICRP103 gives little impact on the pion and muon dosimetries, since the radiation weighting factors assigned to those particles are maintained in the issue. (author)
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.
Desch, Steven; Lorenzo, Alejandro; Ko, Byeongkwan
2016-06-01
We present a computer code we have written for general release that calculates the interior structure and mass-radius relationships of solid exoplanets up to a few Earth masses. The basic algorithm is that of Seager et al. (2007), Zeng & Sasselov (2013) and Dorn et al. (2015): the code integrates the 1-D (spherical) equation of hydrostatic equilibrium to find pressure in shells of various depths assuming a gravitational acceleration, uses the bulk modulus of the materials as inputs to an equation of state to convert pressures into density and volume in each shell, recomputes the shell thicknesses and gravitational acceleration, and iterates the solution to convergence. Unlike most existing codes, we do not impose a particular mineralogy in each shell. Instead we adopt the approach of Dorn et al. (2015), in which we impose a stoichiometry in each shell; for rocky shells and the metal core the code calls the PerpleX code (Connolly et al. 2005) to compute the mineralogy and material properties appropriate to that shell’s stoichiometry, pressure and temperature. Unique attributes of the code are as follows. The mineralogy is complete in the Fe-Mg-Si-O system, including species like FeSi and FeO in the core. We also include FeS (VII) in the core. We have also included an approximate phase diagram for water ice to account for an icy mantle. We also include the effects of adiabatic temperature profiles and a temperature jump at the core-mantle boundary. Finally, we have created a user-friendly interface allowing the code to be downloaded and used as a teaching tool. Results of the code and a demonstration of its use will be presented at the meeting.
Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors
International Nuclear Information System (INIS)
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
Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors
Energy Technology Data Exchange (ETDEWEB)
Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C
2006-10-15
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.
BOA, Beam Optics Analyzer A Particle-In-Cell Code
Energy Technology Data Exchange (ETDEWEB)
Thuc Bui
2007-12-06
The program was tasked with implementing time dependent analysis of charges particles into an existing finite element code with adaptive meshing, called Beam Optics Analyzer (BOA). BOA was initially funded by a DOE Phase II program to use the finite element method with adaptive meshing to track particles in unstructured meshes. It uses modern programming techniques, state-of-the-art data structures, so that new methods, features and capabilities are easily added and maintained. This Phase II program was funded to implement plasma simulations in BOA and extend its capabilities to model thermal electrons, secondary emissions, self magnetic field and implement a more comprehensive post-processing and feature-rich GUI. The program was successful in implementing thermal electrons, secondary emissions, and self magnetic field calculations. The BOA GUI was also upgraded significantly, and CCR is receiving interest from the microwave tube and semiconductor equipment industry for the code. Implementation of PIC analysis was partially successful. Computational resource requirements for modeling more than 2000 particles begin to exceed the capability of most readily available computers. Modern plasma analysis typically requires modeling of approximately 2 million particles or more. The problem is that tracking many particles in an unstructured mesh that is adapting becomes inefficient. In particular memory requirements become excessive. This probably makes particle tracking in unstructured meshes currently unfeasible with commonly available computer resources. Consequently, Calabazas Creek Research, Inc. is exploring hybrid codes where the electromagnetic fields are solved on the unstructured, adaptive mesh while particles are tracked on a fixed mesh. Efficient interpolation routines should be able to transfer information between nodes of the two meshes. If successfully developed, this could provide high accuracy and reasonable computational efficiency.
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.)
International Nuclear Information System (INIS)
This report describes the calculation procedure of the TRANCS code, which deals with fission product transport in fuel rod of high temperature gas-cooled reactor (HTGR). The fundamental equation modeled in the code is a cylindrical one-dimensional diffusion equation with generation and decay terms, and the non-stationary solution of the equation is obtained numerically by a finite difference method. The generation terms consist of the diffusional release from coated fuel particles, recoil release from outer-most coating layer of the fuel particle and generation due to contaminating uranium in the graphite matrix of the fuel compact. The decay term deals with neutron capture as well as beta decay. Factors affecting the computation error has been examined, and further extention of the code has been discussed in the fields of radial transport of fission products from graphite sleeve into coolant helium gas and axial transport in the fuel rod. (author)
International Nuclear Information System (INIS)
This report describes a reactor-containment code, ALICE, which uses an arbitrary Lagrangian-Eulerian method to describe the coolant motion, together with a Lagrangian method to analyze the response of the containment vessel and other solid media inside a reactor containment. The finite-difference formulation used to approximate the governing equations for the motion of the coolant can be solved in either an explicit or an implicit scheme; the finite-element formulation used to approximate the governing equations for the containment vessel and other solid media can be performed only in the explicit scheme. Thus, the ALICE code can perform two types of coupling calculations for the fluid and structure (implicit-explicit and explicit-explicit). The code is generalized so that it can apply to problems either in a two-dimensional Cartesian or in a two-dimensional cylindrical-coordinate system
Directory of Open Access Journals (Sweden)
Khedr Ahmed
2005-01-01
Full Text Available The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.
Energy Technology Data Exchange (ETDEWEB)
Glueckstern, P.; Reed, S.A.; Wilson, J.V.
1976-11-01
The reverse osmosis process has been used extensively for the conversion of brackish waters to potable water. The process is now nearing commercialization as a means for the conversion of seawater. The computer program (RO-75) is a Fortran code for the optimizatin of the design and economics of seawater reverse osmosis plants. The examples described are based on currently available, commercial membrane modules and prevailing prices. However, the code is very flexible and can be used to optimize plants utilizing future technological improvements and different economic parameters.
International Nuclear Information System (INIS)
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 P1- and Ssub(N)-calculations can be generated. Anisotropy of elastic scattering can be taken into account up to P5. 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.)
International Nuclear Information System (INIS)
The computer code BUSH has been developed for the calculation of steady state heat transfer in a rod bundle. For a given power, flow and geometry it can calculate the temperatures in the rods, coolant and shroud assuming that at any axial level each rod can be described by one temperature and the coolant fluid is also radially uniform at this level. Heat transfer by convection and radiation are handled and the geometry is flexible enough to model nearly all types of envisaged shroud design for the SUPERSARA test series. The modular way in which BUSH has been written makes it suitable for future development, either within the present BUSH framework or as part of a more advanced code
International Nuclear Information System (INIS)
The need of the experimental support for validation of the computational tools to be applied to analyze the mixing of diluted slugs has been recognized in various countries. The test series for the International Standard Problem ISP-43 provides a platform for experiences to be applied to the simulation of a well-defined test series. Test A and B of the UM2x4 loop test facility were calculated with the CFD Code CFX-4.3. The results show qualitatively good agreement with the experimental data for both tests. The structure of the flow field and the form of the propagating temperature perturbation front are well modeled by the CFD code. However, deviations occur at local positions. Comparative calculations with and without taking into account buoyancy have shown, that buoyancy effects are noticeable, but the mixing is mainly momentum controlled. (orig.)
International Nuclear Information System (INIS)
The program system KENOREST version 1998 will be presented, which is a useful tool for burnup and reactivity calculations for LWR fuel. The three-dimensional Monte Carlo code KENO-V.a is coupled with the one-dimensional GRS burnup program system OREST-98. The objective is to achieve a better modelling of plutonium and actinide build-up or burnout for advanced heterogeneous fuel assembly designs. Further objectives are directed to reliable calculations of the pin power distributions and of reactor safety parameters including axial and radial rod temperatures for fuel assemblies of modern design. The stand-alone-code KENO-V.a version is used without any changes in the program source. The OREST-98 system was developed to handle multirod problems and additional burnup dependent moderator conditions which can be applied to stretch-out simulations in the reactor. A new interface module RESPEFF between KENO and OREST transforms the 2-d or 3-d KENO flux results to the one-dimensional lattice code OREST in a fully automated manner to maintain reaction rate balance between the codes. First results for assembly multiplication factors, isotope inventories are compared with OECD results. (author)
International Nuclear Information System (INIS)
In this paper, multi-group microscopic cross-section uncertainty is propagated through the DRAGON (Version 4) lattice code, in order to perform uncertainty analysis on k∞ and 2-group homogenized macroscopic cross-sections predictions. A statistical methodology is employed for such purposes, where cross-sections of certain isotopes of various elements belonging to the 172 groups DRAGLIB library format, are considered as normal random variables. This library is based on JENDL-4 data, because JENDL-4 contains the largest amount of isotopic covariance matrixes among the different major nuclear data libraries. The aim is to propagate multi-group nuclide uncertainty by running the DRAGONv4 code 500 times, and to assess the output uncertainty of a test case corresponding to a 17 x 17 PWR fuel assembly segment without poison. The chosen sampling strategy for the current study is Latin Hypercube Sampling (LHS). The quasi-random LHS allows a much better coverage of the input uncertainties than simple random sampling (SRS) because it densely stratifies across the range of each input probability distribution. Output uncertainty assessment is based on the tolerance limits concept, where the sample formed by the code calculations infers to cover 95% of the output population with at least a 95% of confidence. This analysis is the first attempt to propagate parameter uncertainties of modern multi-group libraries, which are used to feed advanced lattice codes that perform state of the art resonant self-shielding calculations such as DRAGONv4. (authors)
Status and validation of the fast reactor lattice code KAPER-2 for slab and pin cells
International Nuclear Information System (INIS)
The cell programm KAPER for the calculation of the neutronics parameters of plate cells has been used frequently for the analysis of experiments performed in fast critical assemblies, especially at the facility SNEAK at Karlsruhe. KAPER determines cell-averaged group cross sections and diffusion coefficients. So far, it was restricted to slab geometry, because heterogeneity effects are usually much larger in plate-type critical assemblies than in fast power reactors. However, to allow an adequate prediction of the physics parameters of a sodium-cooled fast power reactor, especially in voided configurations, where streaming is important, a pin-cell code is required. Therefore, the version KAPER-2 was developed recently, which has options both for slab and for pin cells. In this paper, the method and the approximations used in KAPER-2 will be discussed. Probably the most serious limitation of the present version is the use of the simple Benoist formula to calculate anisotropic diffusion coefficients. Parallel with the code development, an effort is going on to check the validity of the different approximations by a Monte Carlo technique. The status of this validation effort, especially concerning the range of validity of the Benoist equation for slab lattices, will also be described. A more accurate method to calculate parallel streaming, similar to the method by Koehler and Ligou, is suggested, which may be interpreted as including parallel leakage in the collision probabilities. For pin lattices, the Benoist equation is compared with the most elaborate method available at Karlsruhe, the one developed by Eisemann. (author)
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.
Gamma Dose Calculations in the Target Service Cell of the SNS
International Nuclear Information System (INIS)
Calculations of the gamma dose rates inside and outside of the Target Service Cell (TSC) of the Spallation Neutron Source (SNS) are complicated by the large size of the structure, large volume of air (internal void), optical thickness of the enclosing walls, and multiplicity of radiation sources. Furthermore, a reasonably detailed distribution of the dose rate over the volume of the TSC, and on the outside of its walls is necessary in order to optimize electronic instrument locations, and plan access control. For all these reasons a deterministic transport method was preferred over Monte Carlo, The three- dimensional neutral particle transport code TORT was employed for this purpose with support from other peripheral codes in the Discrete Ordinates of Oak Ridge System (DOORS). The computational model for the TSC is described and the features of TORT and its companion codes that enable such a difficult calculation are discussed. Most prominent is the presence of severe ray effects in the air cavity of the TSC that persists in the transport through the concrete walls and is pronounced throughout the problem volume. Initial attempts at eliminating ray effects from the computed results using the newly developed three-dimensional uncollided flux and first collided source code GRTUNCL3D are described