SLAROM: a code for cell homogenization calculation of fast reactor
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
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
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
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
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
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
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
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
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
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
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
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)
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
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
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
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)
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
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
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.
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
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
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
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
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
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)
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
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
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
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
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)
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
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
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
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
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)
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
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
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'
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
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
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
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
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
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
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
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
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)
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)
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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.
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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.
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.)
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)
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
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
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)
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)
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
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
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
The MCU-RFFI/A and MCU-REA codes developed within the framework of the long term MCU project are widely used for calculations of neutron physic characteristics of WWER type reactors. Complete descriptions of the codes are available in both Russian and English. The codes are verified and validated by means of the comparison of calculated results with experimental data and mathematical benchmarks. The codes are licensed by Russian Nuclear and Criticality Safety Regulatory Body (Gosatomnadzor RF) (Code Passports: N 61 of 17.10.1966 and N 115 of 02.03.2000 accordingly)). The report gives examples of WWER reactor physic tasks important for practice solved using the codes from the MCU family. Some calculational results are given too. (Authors)
SAMDIST: A computer code for calculating statistical distributions for R-matrix resonance parameters
Leal, L.C.; 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
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
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)
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
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
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
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
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.
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)
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
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
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
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
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
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
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
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
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.
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
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
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
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
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
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...
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
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
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
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)
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
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)
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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...
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)
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 -
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
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
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
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
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
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
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
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
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
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
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)
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
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
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
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
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
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)
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
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
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)
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%.
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
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.
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
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
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
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
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
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)
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
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
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
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
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
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
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
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.
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
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)
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
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
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
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
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
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
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