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Sample records for process calculation code

  1. Calculation code revised MIXSET for Purex process

    International Nuclear Information System (INIS)

    Gonda, Kozo; Oka, Koichiro; Fukuda, Shoji.

    1979-02-01

    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), NH 2 OH, N 2 H 4 can be simulated. 1st Gr. (i) reduction of Pu(IV); U 4+ + 2Pu 4+ + 2H 2 O → UO 2 2+ + 2Pu 3+ + 4H + . (ii) oxidation of Pu(III); 2Pu 3+ + 3H + + NO 3 - → 2Pu 4+ + HNO 2 + H 2 O. (iii) oxidation of U(IV); U 4+ + NO 3 - + H 2 O → UO 2 2+ + H + + HNO 2 2U 4+ + O 2 + 2H 2 O → 2UO 2 2+ + 4H + . (iv) decomposition of HNO 2 ; HNO 2 + N 2 H 5 + → HN 3 + 2H 2 O + H + . (author)

  2. Calculation code MIXSET for Purex process

    International Nuclear Information System (INIS)

    Gonda, Kozo; Fukuda, Shoji.

    1977-09-01

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

  3. Calculation code PULCO for Purex process in pulsed column

    International Nuclear Information System (INIS)

    Gonda, Kozo; Matsuda, Teruo

    1982-03-01

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

  4. Calculation code of mass and heat transfer in a pulsed column for Purex process

    International Nuclear Information System (INIS)

    Tsukada, Takeshi; Takahashi, Keiki

    1993-01-01

    A calculation code for extraction behavior analysis in a pulsed column employed at an extraction process of a reprocessing plant was developed. This code was also combined with our previously developed calculation code for axial temperature profiles in a pulsed column. The one-dimensional dispersion model was employed for both of the extraction behavior analysis and the axial temperature profile analysis. The reported values of the fluid characteristics coefficient, the transfer coefficient and the diffusivities in the pulsed column were used. The calculated concentration profiles of HNO 3 , U and Pu for the steady state have a good agreement with the reported experimental results. The concentration and temperature profiles were calculated under the operation conditions which induce the abnormal U extraction behavior, i.e. U extraction zone is moved to the bottom of the column. Thought there is slight difference between calculated and experimental value, it is appeared that our developed code can be applied to the simulation under the normal operation condition and the relatively slowly transient condition. Pu accumulation phenomena was analyzed with this code and the accumulation tendency is similar to the reported analysis results. (author)

  5. Process of cross section generation for radiation shielding calculations, using the NJOY code

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1986-10-01

    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) [pt

  6. Implementation of decommissioning materials conditional clearance process to the OMEGA calculation code

    International Nuclear Information System (INIS)

    Zachar, Matej; Necas, Vladimir; Daniska, Vladimir

    2011-01-01

    The activities performed during nuclear installation decommissioning process inevitably lead to the production of large amount of radioactive material to be managed. Significant part of materials has such low radioactivity level that allows them to be released to the environment without any restriction for further use. On the other hand, for materials with radioactivity slightly above the defined unconditional clearance level, there is a possibility to release them conditionally for a specific purpose in accordance with developed scenario assuring that radiation exposure limits for population not to be exceeded. The procedure of managing such decommissioning materials, mentioned above, could lead to recycling and reuse of more solid materials and to save the radioactive waste repository volume. In the paper an a implementation of the process of conditional release to the OMEGA Code is analyzed in details; the Code is used for calculation of decommissioning parameters. The analytical approach in the material parameters assessment, firstly, assumes a definition of radiological limit conditions, based on the evaluation of possible scenarios for conditionally released materials, and their application to appropriate sorter type in existing material and radioactivity flow system. Other calculation procedures with relevant technological or economical parameters, mathematically describing e.g. final radiation monitoring or transport outside the locality, are applied to the OMEGA Code in the next step. Together with limits, new procedures creating independent material stream allow evaluation of conditional material release process during decommissioning. Model calculations evaluating various scenarios with different input parameters and considering conditional release of materials to the environment are performed to verify the implemented methodology. Output parameters and results of the model assessment are presented, discussed and conduced in the final part of the paper

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

    International Nuclear Information System (INIS)

    Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I.

    2015-01-01

    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.

  8. Verification test calculations for the Source Term Code Package

    International Nuclear Information System (INIS)

    Denning, R.S.; Wooton, R.O.; Alexander, C.A.; Curtis, L.A.; Cybulskis, P.; Gieseke, J.A.; Jordan, H.; Lee, K.W.; Nicolosi, S.L.

    1986-07-01

    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

  9. Effect of interpolation error in pre-processing codes on calculations of self-shielding factors and their temperature derivatives

    International Nuclear Information System (INIS)

    Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullan, D.E.

    1986-01-01

    We investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. We consider the 2.0347 to 3.3546 keV energy region for 238 U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems. (author)

  10. Effect of interpolation error in pre-processing codes on calculations of self-shielding factors and their temperature derivatives

    International Nuclear Information System (INIS)

    Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullen, D.E.

    1985-01-01

    The authors investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. They consider the 2.0347 to 3.3546 keV energy region for /sup 238/U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems

  11. The 1992 ENDF Pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1992-01-01

    This document summarizes the 1992 version of the ENDF pre-processing codes which are required for processing evaluated nuclear data coded in the format ENDF-4, ENDF-5, or ENDF-6. Included are the codes CONVERT, MERGER, LINEAR, RECENT, SIGMA1, LEGEND, FIXUP, GROUPIE, DICTION, MIXER, VIRGIN, COMPLOT, EVALPLOT, RELABEL. Some of the functions of these codes are: to calculate cross-sections from resonance parameters; to calculate angular distributions, group average, mixtures of cross-sections, etc; to produce graphical plottings and data comparisons. The codes are designed to operate on virtually any type of computer including PC's. They are available from the IAEA Nuclear Data Section, free of charge upon request, on magnetic tape or a set of HD diskettes. (author)

  12. Code ATOM for calculation of atomic characteristics

    International Nuclear Information System (INIS)

    Vainshtein, L.A.

    1990-01-01

    In applying atomic physics to problems of plasma diagnostics, it is necessary to determine some atomic characteristics, including energies and transition probabilities, for very many atoms and ions. Development of general codes for calculation of many types of atomic characteristics has been based on general but comparatively simple approximate methods. The program ATOM represents an attempt at effective use of such a general code. This report gives a brief description of the methods used, and the possibilities of and limitations to the code are discussed. Characteristics of the following processes can be calculated by ATOM: radiative transitions between discrete levels, radiative ionization and recombination, collisional excitation and ionization by electron impact, collisional excitation and ionization by point heavy particle (Born approximation only), dielectronic recombination, and autoionization. ATOM explores Born (for z=1) or Coulomb-Born (for z>1) approximations. In both cases exchange and normalization can be included. (N.K.)

  13. Electro-magnetic cascade calculation using EGS4 code

    International Nuclear Information System (INIS)

    Namito, Yoshihito; Hirayama, Hideo

    2001-01-01

    The outline of the general-purpose electron-photon transport code EGS4 (Electron-Gamma-Shower Version 4) is described. In section 1, the history of the electron photon Monte Carlo transport code toward EGS4 is described. In section 2, the features of the EGS4 and the physical processes treated, cross section preparation and language is explained. The upper energy limit of EGS4 is a few thousand GeV. The lower energy limit of EGS4 is 1 keV and 10 keV for photon and electron, respectively. In section 3, particle transport method in EGS4 code is discussed. The points are; condensed history method, continuous slowing down approximation and multiple scattering approximation. Order of the particle transport calculation is also mentioned. The switches to control scoring routine AUSGAB is listed. In section 4, the output from the code is described. In section 5, several benchmark calculations are described. (author)

  14. The MCEF code for nuclear evaporation and fission calculations

    International Nuclear Information System (INIS)

    Deppman, A.; Pina, S.R. de; Likhachev, V.P.; Mesa, J.; Arruda-Neto, J.D.T.; Rodriguez, O.; Goncalves, M.

    2001-11-01

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

  15. Two-dimensional sensitivity calculation code: SENSETWO

    International Nuclear Information System (INIS)

    Yamauchi, Michinori; Nakayama, Mitsuo; Minami, Kazuyoshi; Seki, Yasushi; Iida, Hiromasa.

    1979-05-01

    A SENSETWO code for the calculation of cross section sensitivities with a two-dimensional model has been developed, on the basis of first order perturbation theory. It uses forward neutron and/or gamma-ray fluxes and adjoint fluxes obtained by two-dimensional discrete ordinates code TWOTRAN-II. The data and informations of cross sections, geometry, nuclide density, response functions, etc. are transmitted to SENSETWO by the dump magnetic tape made in TWOTRAN calculations. The required input for SENSETWO calculations is thus very simple. The SENSETWO yields as printed output the cross section sensitivities for each coarse mesh zone and for each energy group, as well as the plotted output of sensitivity profiles specified by the input. A special feature of the code is that it also calculates the reaction rate with the response function used as the adjoint source in TWOTRAN adjoint calculation and the calculated forward flux from the TWOTRAN forward calculation. (author)

  16. ERRORJ. Covariance processing code. Version 2.2

    International Nuclear Information System (INIS)

    Chiba, Go

    2004-07-01

    ERRORJ is the covariance processing code that can produce covariance data of multi-group cross sections, which are essential for uncertainty analyses of nuclear parameters, such as neutron multiplication factor. The ERRORJ code can process the covariance data of cross sections including resonance parameters, angular and energy distributions of secondary neutrons. Those covariance data cannot be processed by the other covariance processing codes. ERRORJ has been modified and the version 2.2 has been developed. This document describes the modifications and how to use. The main topics of the modifications are as follows. Non-diagonal elements of covariance matrices are calculated in the resonance energy region. Option for high-speed calculation is implemented. Perturbation amount is optimized in a sensitivity calculation. Effect of the resonance self-shielding on covariance of multi-group cross section can be considered. It is possible to read a compact covariance format proposed by N.M. Larson. (author)

  17. Benchmark calculation of subchannel analysis codes

    International Nuclear Information System (INIS)

    1996-02-01

    In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)

  18. Development of the multistep compound process calculation code

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  19. Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

    A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.

  20. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

    Ryufuku, Hiroshi; Numakunai, Takao; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1979-03-01

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

  1. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.

    1997-06-01

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

  2. Burnup calculation code system COMRAD96

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu

    1997-06-01

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

  3. Nuclear Data Processing for Reactor Physics Calculation

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Pandiangan, Tumpal

    2003-01-01

    Nuclear data processing for reactor physics calculation has been done. Raw nuclear data cross-sections on file ENDF should be prepared and processed before it used in neutronic calculation. The processing code system such as NJOY-PC code has been used from linearization of nuclear cross-sections data and background contribution of resonance parameter (MF2) using RECONR module (0K) with energy range from 10 -5 to 10 7 eV. Afterward, the neutron cross-sections data should be processed and broadened to desire temperature (i.e. 293K) by using BROADR module. The Grouper and Therma modules will be applied for multi-groups calculation which suitable for WIMS/D4 (69 groups) and thermalization of nuclear constants. The final stage of processing nuclear cross-sections is updating WIMS/D4 library. The WIMSR module in NJOY-PC and WILLIE code will be applied in this stage. The evaluated nuclear data file, especially for 1 H 1 isotope, was taken from JENDL-3.2 and ENDF/B-VI for preliminary study. The results of nuclear data processing 1 H 1 shows that the old-WIMS (WIMS-lama) library have much discrepancies comparing with JENDL-3.2 or ENDF/B-VI files, especially in energy around 5 keV

  4. Development of codes for physical calculations of WWER

    International Nuclear Information System (INIS)

    Novikov, A.N.

    2000-01-01

    A package of codes for physical calculations of WWER reactors, used at the RRC 'Kurchatov Institute' is discussed including the purpose of these codes, approximations used, degree of data verification, possibilities of automation of calculations and presentation of results, trends of further development of the codes. (Authors)

  5. CONSUL code package application for LMFR core calculations

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  6. TEA: A CODE CALCULATING THERMOCHEMICAL EQUILIBRIUM ABUNDANCES

    Energy Technology Data Exchange (ETDEWEB)

    Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver, E-mail: jasmina@physics.ucf.edu [Planetary Sciences Group, Department of Physics, University of Central Florida, Orlando, FL 32816-2385 (United States)

    2016-07-01

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

  7. TEA: A CODE CALCULATING THERMOCHEMICAL EQUILIBRIUM ABUNDANCES

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  8. Improvement of calculation method for temperature coefficient of HTTR by neutronics calculation code based on diffusion theory. Analysis for temperature coefficient by SRAC code system

    International Nuclear Information System (INIS)

    Goto, Minoru; Takamatsu, Kuniyoshi

    2007-03-01

    The HTTR temperature coefficients required for the core dynamics calculations had been calculated from the HTTR core calculation results by the diffusion code with which the corrections had been performed using the core calculation results by the Monte-Carlo code MVP. This calculation method for the temperature coefficients was considered to have some issues to be improved. Then, the calculation method was improved to obtain the temperature coefficients in which the corrections by the Monte-Carlo code were not required. Specifically, from the point of view of neutron spectrum calculated by lattice calculations, the lattice model was revised which had been used for the calculations of the temperature coefficients. The HTTR core calculations were performed by the diffusion code with the group constants which were generated by the lattice calculations with the improved lattice model. The core calculations and the lattice calculations were performed by the SRAC code system. The HTTR core dynamics calculation was performed with the temperature coefficient obtained from the core calculation results. In consequence, the core dynamics calculation result showed good agreement with the experimental data and the valid temperature coefficient could be calculated only by the diffusion code without the corrections by Monte-Carlo code. (author)

  9. Development of the code package KASKAD for calculations of WWERs

    International Nuclear Information System (INIS)

    Bolobov, P.A.; Lazarenko, A.P.; Tomilov, M.Ju.

    2008-01-01

    The new version of software package for neutron calculation of WWER cores KASKAD 2007 consists of some calculating and service modules, which are integrated in the common framework. The package is based on the old version, which was expanded with some new functions and the new calculating modules, such as: -the BIPR-2007 code is the new one which performs calculation of power distribution in three-dimensional geometry for 2-group neutron diffusion calculation. This code is based on the BIPR-8KN model, provides all possibilities of BIPR-7A code and uses the same input data; -the PERMAK-2007 code is pin-by-pin few-group multilayer and 3-D code for neutron diffusion calculation; -graphical user interface for input data preparation of the TVS-M code. The report also includes some calculation results obtained with modified version of the KASKAD 2007 package. (Authors)

  10. CRACKEL: a computer code for CFR fuel management calculations

    International Nuclear Information System (INIS)

    Burstall, R.F.; Ball, M.A.; Thornton, D.E.J.

    1975-12-01

    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)

  11. Computer codes for the calculation of vibrations in machines and structures

    International Nuclear Information System (INIS)

    1989-01-01

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

  12. Improvement of the MSG code for the MONJU evaporators. Additional function of reverse flow calculation on water/steam model and animation for post processing

    International Nuclear Information System (INIS)

    Toda, Shin-ichi; Yoshikawa, Shinji; Oketani, Kazuhiro

    2003-05-01

    The improved version of the MSG code (Multi-dimensional Thermal-hydraulic Analysis Code for Steam Generators) has been released. It has been carried out to improve based on the original version in order to calculate reverse flow on water/steam side, and to animate the post-processing data. To calculate reverse flow locally, modification to set pressure at each divided node point of water/steam region in the helical-coil heat transfer tubes has been carried out. And the matrix solver has been also improved to treat a problem within practical calculation time against increasing the pressure points. In this case pressure and enthalpy have to be calculated simultaneously, however, it was found out that using the block-Jacobean method make a diagonal-dominant matrix, and solve the matrix efficiently with a relaxation method. As the result of calculations of a steady-state condition and a transient of SG blow down with manual trip operation, the improvement on calculation function of the MSG code was confirmed. And an animation function of temperature contour in the sodium shell side as a post processing has been added. Since the animation is very effective to understand thermal-hydraulic behavior on the sodium shell side of the SG, especially in case of transient condition, the analysis and evaluation of the calculation results will be enabled to be more quickly and effectively. (author)

  13. Progress on China nuclear data processing code system

    Science.gov (United States)

    Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu

    2017-09-01

    China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.

  14. Calculation of the real states of Ignalina NPP Unit 1 and Unit 2 RBMK-1500 reactors in the verification process of QUABOX/CUBBOX code

    International Nuclear Information System (INIS)

    Bubelis, E.; Pabarcius, R.; Demcenko, M.

    2001-01-01

    Calculations of the main neutron-physical characteristics of RBMK-1500 reactors of Ignalina NPP Unit 1 and Unit 2 were performed, taking real reactor core states as the basis for these calculations. Comparison of the calculation results, obtained using QUABOX/CUBBOX code, with experimental data and the calculation results, obtained using STEPAN code, showed that all the main neutron-physical characteristics of the reactors of Unit 1 and Unit 2 of Ignalina NPP are in the safe deviation range of die analyzed parameters, and that reactors of Ignalina NPP, during the process of the reactor core composition change, are operated in a safe and stable manner. (author)

  15. DNBR calculation in digital core protection system by a subchannel analysis code

    International Nuclear Information System (INIS)

    In, W. K.; Yoo, Y. J.; Hwang, T. H.; Ji, S. K.

    2001-01-01

    The DNBR calculation uncertainty and DNBR margin were evaluated in digital core protection system by a thermal-hydrualic subchannel analysis code MATRA. A simplified thermal-hydraulic code CETOP is used to calculate on-line DNBR in core protection system at a digital PWR. The DNBR tuning process against a best-estimate subchannel analysis code is required for CETOP to ensure accurate and conservative DNBR calculation but not necessary for MATRA. The DNBR calculations by MATRA and CETOP were performed for a large number of operating condition in Yonggwang nulcear units 3-4 where the digitial core protection system is initially implemented in Korea. MATRA resulted in a less negative mean value (i.e., reduce the overconservatism) and a somewhat larger standard deviation of the DNBR error. The uncertainty corrected minimum DNBR by MATRA was shown to be higher by 1.8% -9.9% that the CETOP DNBR

  16. Comprehensive nuclear model calculations: theory and use of the GNASH code

    International Nuclear Information System (INIS)

    Young, P.G.; Arthur, E.D.; Chadwick, M.B.

    1998-01-01

    The theory and operation of the nuclear reaction theory computer code GNASH is described, and detailed instructions are presented for code users. The code utilizes statistical Hauser-Feshbach theory with full angular momentum conservation and includes corrections for preequilibrium effects. This version is expected to be applicable for incident particle energies between 1 keV and 150 MeV and for incident photon energies to 140 MeV. General features of the code, the nuclear models that are utilized, input parameters needed to perform calculations, and the output quantities from typical problems are described in detail. A number of new features compared to previous versions are described in this manual, including the following: (1) inclusion of multiple preequilibrium processes, which allows the model calculations to be performed above 50 MeV; (2) a capability to calculate photonuclear reactions; (3) a method for determining the spin distribution of residual nuclei following preequilibrium reactions; and (4) a description of how preequilibrium spectra calculated with the FKK theory can be utilized (the 'FKK-GNASH' approach). The computational structure of the code and the subroutines and functions that are called are summarized as well. Two detailed examples are considered: 14-MeV neutrons incident on 93 Nb and 12-MeV neutrons incident on 238 U. The former example illustrates a typical calculation aimed at determining neutron, proton, and alpha emission spectra from 14-MeV reactions, and the latter example demonstrates use of the fission model in GNASH. Results from a variety of other cases are illustrated. (author)

  17. Code accuracy evaluation of ISP 35 calculations based on NUPEC M-7-1 test

    International Nuclear Information System (INIS)

    Auria, F.D.; Oriolo, F.; Leonardi, M.; Paci, S.

    1995-01-01

    Quantitative evaluation of code uncertainties is a necessary step in the code assessment process, above all if best-estimate codes are utilised for licensing purposes. Aiming at quantifying the code accuracy, an integral methodology based on the Fast Fourier Transform (FFT) has been developed at the University of Pisa (DCMN) and has been already applied to several calculations related to primary system test analyses. This paper deals with the first application of the FFT based methodology to containment code calculations based on a hydrogen mixing and distribution test performed in the NUPEC (Nuclear Power Engineering Corporation) facility. It is referred to pre-test and post-test calculations submitted for the International Standard Problem (ISP) n. 35. This is a blind exercise, simulating the effects of steam injection and spray behaviour on gas distribution and mixing. The result of the application of this methodology to nineteen selected variables calculated by ten participants are here summarized, and the comparison (where possible) of the accuracy evaluated for the pre-test and for the post-test calculations of a same user is also presented. (author)

  18. Verification of the LWRARC code for light-water-reactor afterheat rate calculations

    International Nuclear Information System (INIS)

    Murphy, B.D.

    1998-02-01

    This report describes verification studies carried out on the LWRARC (Light-Water-Reactor Afterheat Rate Calculations) computer code. The LWRARC code is proposed for automating the implementation of procedures specified in Draft Revision 1 of the U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 3.54, open-quotes Spent-Fuel Heat Generation in an Independent Spent-Fuel Storage Installation,close quotes which gives guidelines on the calculation of decay heat for spent nuclear fuel. Draft Regulatory Guide 3.54 allows one to estimate decay-heat values by means of a table lookup procedure with interpolation performed between table-entry values. The tabulated values of the relevant parameters span ranges that are appropriate for spent fuel from a boiling-water reactor (BWR) or a pressurized-water reactor (PWR), as the case may be, and decay-heat rates are obtained for spent fuel whose properties are within those parameter limits. In some instances, where these limits are either exceeded or where they approach critical regions, adjustments are invoked following table lookup. The LWRARC computer code is intended to replicate the manual process just described. In the code, the table lookup is done by entering a database and carrying out interpolations. The code then determines if adjustments apply, and, if this is the case, adjustment factors are calculated separately. The manual procedures in the Draft Regulatory Guide have been validated (i.e., they produce results that are good estimates of reality). The work reported in this document verifies that the LWRARC code replicates the manual procedures of the Draft Regulatory Guide, and that the code, taken together with the Draft Regulatory Guide, can support both verification and validation processes

  19. GRIMH3: A new reactor calculation code at Savannah River Site

    International Nuclear Information System (INIS)

    Le, T.T.; Pevey, R.E.

    1993-01-01

    The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex

  20. SCRAM reactivity calculations with the KIKO3D code

    International Nuclear Information System (INIS)

    Hordosy, G.; Kerszturi, A.; Maraczy, Cs.; Temesvari, E.

    1999-01-01

    Discrepancies between calculated static reactivities and measured reactivities evaluated with reactivity meters led to investigating SCRAM with the KIKO3D dynamic code, The time and space dependent neutron flux in the reactor core during the rod drop measurement was calculated by the KIKO3D nodal diffusion code. For calculating the ionisation chamber signals the Green function technique was applied. The Green functions of ionisation chambers were evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals during asymmetric SCRAM measurements were calculated and compared with measured data using the inverse point kinetics transformation. The sufficient agreement validates the KIKO3D code to determine the reactivities after SCRAM. (Authors)

  1. Calculation codes in radioprotection, radio-physics and dosimetry

    International Nuclear Information System (INIS)

    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.

    2010-01-01

    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 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 p (3)/K 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 the pulmonary

  2. OPAL reactor calculations using the Monte Carlo code serpent

    Energy Technology Data Exchange (ETDEWEB)

    Ferraro, Diego; Villarino, Eduardo [Nuclear Engineering Dept., INVAP S.E., Rio Negro (Argentina)

    2012-03-15

    In the present work the Monte Carlo cell code developed by VTT Serpent v1.1.14 is used to model the MTR fuel assemblies (FA) and control rods (CR) from OPAL (Open Pool Australian Light-water) reactor in order to obtain few-group constants with burnup dependence to be used in the already developed reactor core models. These core calculations are performed using CITVAP 3-D diffusion code, which is well-known reactor code based on CITATION. Subsequently the results are compared with those obtained by the deterministic calculation line used by INVAP, which uses the Collision Probability Condor cell-code to obtain few-group constants. Finally the results are compared with the experimental data obtained from the reactor information for several operation cycles. As a result several evaluations are performed, including a code to code cell comparison at cell and core level and calculation-experiment comparison at core level in order to evaluate the Serpent code actual capabilities. (author)

  3. Validation of the MCNP-DSP Monte Carlo code for calculating source-driven noise parameters of subcritical systems

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.

    1995-01-01

    This paper describes calculations performed to validate the modified version of the MCNP code, the MCNP-DSP, used for: the neutron and photon spectra of the spontaneous fission of californium 252; the representation of the detection processes for scattering detectors; the timing of the detection process; and the calculation of the frequency analysis parameters for the MCNP-DSP code

  4. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro

    2007-02-01

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

  5. Calculation code used in criticality analyses for the accident of JCO precipitation tank

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori

    2000-01-01

    In order to evaluate nuclear features on criticality accident formed at the nuclear fuel processing facility in Tokai Works of the JCO, Ltd. (JCO), in Tokai-mura, Ibaraki prefecture, dynamic analyses to calculate output change after occurring the accident as well as criticality analyses to calculate reactivity added to precipitation tank, were carried out according to scenario on accident formation. For the criticality analyses, a continuous energy Monte Carlo code MCNP was used to carry out calculation of reactivity fed into the precipitation tank as correctly as possible. And, SRAC code system was used for calculation on temperature and void reactivity coefficients, effective delayed neutron ratio beta eff , and instantaneous neutron generation time required for parameters controlling transition features at criticality accident. In addition, for the dynamic analyses, because of necessity of considering on volume expansion of solution fuels used as exothermic body and radiation decomposition gas forming into solution, output behavior, numbers of nuclear fission, and so forth at initial burst portion were calculated by using TRACE and quasi-regular code, at a center of AGNES-2 promoting on its development in JAERI. Here were reported on outlines and an analysis example on calculation code using for the nuclear features evaluation. (G.K.)

  6. Data processing codes for fatigue and tensile tests

    International Nuclear Information System (INIS)

    Sanchez Sarmiento, Gustavo; Iorio, A.F.; Crespi, J.C.

    1981-01-01

    The processing of fatigue and tensile tests data in order to obtain several parameters of engineering interest requires a considerable effort of numerical calculus. In order to reduce the time spent in this work and to establish standard data processing from a set of similar type tests, it is very advantageous to have a calculation code for running in a computer. Two codes have been developed in FORTRAN language; one of them predicts cyclic properties of materials from the monotonic and incremental or multiple cyclic step tests (ENSPRED CODE), and the other one reduces data coming from strain controlled low cycle fatigue tests (ENSDET CODE). Two examples are included using Zircaloy-4 material from different manufacturers. (author) [es

  7. Calculation code NIRVANA for free boundary MHD equilibrium

    International Nuclear Information System (INIS)

    Ninomiya, Hiromasa; Suzuki, Yasuo; Kameari, Akihisa

    1975-03-01

    The calculation method and code of solving the free boundary problem for MHD equilibrium has been developed. Usage of the code ''NIRVANA'' is described. The toroidal plasma current density determined as a function of the flux function PSI is substituted by a group of the ring currents, whereby the equation of MHD equilibrium is transformed into an integral equation. Either of the two iterative methods is chosen to solve the integral equation, depending on the assumptions made of the plasma surface points. Calculation of the magnetic field configurations is possible when the plasma surface coincides self-consistently with the magnetic flux including the separatrix points. The code is usable in calculation of the circular or non-circular shell-less Tokamak equilibrium. (auth.)

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

    International Nuclear Information System (INIS)

    Villarino, E.A.

    1990-01-01

    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) [es

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-04-15

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

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

    International Nuclear Information System (INIS)

    Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan

    2011-01-01

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

  11. Verification calculations for the WWER version of the TRANSURANUS code

    International Nuclear Information System (INIS)

    Elenkov, D.; Boneva, S.; Georgieva, M.; Georgiev, S.; Schubert, A.; Van Uffelen, P.

    2006-01-01

    The paper presents part of the work performed in the study project 'Research and Development for Licensing of Nuclear Fuel in Bulgaria'. The main objective of the project is to provide assistance for solving technical questions of the fuel licensing process in Bulgaria. One important issue is the extension of the predictive capabilities of fuel performance codes for Russian-type WWER reactors. In the last decade, a series of international projects has been based on the TRANSURANUS fuel performance code: Specific models for WWER fuel have been developed and implemented in the code in the late 90's. In 2000-2003, basic verification work was done by using experimental data of nuclear fuel irradiated in WWER-440 reactors. While the present paper focuses on the analysis of WWER-1000 standard fuel under normal operating conditions, the above study project covers additional tasks: 1) Post-irradiation calculations of ramp tests performed in the DR3 test reactor of the Risoe National Laboratory (five instrumented fuel rods of the Risoe 3 dataset contained in the IFPE database) using the TRANSURANUS code; 2) Compilation of cross-section libraries for isotope evolution calculations in WWER-440 and WWER-1000 fuel assemblies using the ORIGEN-S code; 3) Analysis of current situation and needs for an extension of the curriculum in Nuclear Engineering at the Technical University of Sofia. In this paper the post-irradiation calculations of steady-state irradiation experiments with nuclear fuel for Russian-type WWER-1000 reactors, using the latest release of the TRANSURANUS code (v1m1j03)are presented. Regarding a comprehensive verification of modern fuel performance codes, the burn-up region above 40 MWd/kgU is of increasing importance. A number of new phenomena emerge at high fuel burn-up, implying the need for enlarged databases of postirradiation examinations (PIE). For one fuel assembly irradiated in a WWER-1000 reactor with a rod discharge burn-up between 50 and 55 MWd

  12. FISPIN - a computer code for nuclide inventory calculations

    International Nuclear Information System (INIS)

    Burstall, R.F.

    1979-10-01

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

  13. Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

    2005-03-15

    Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too.

  14. Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

    International Nuclear Information System (INIS)

    Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

    2005-03-01

    Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too

  15. Multi-group diffusion perturbation calculation code. PERKY (2002)

    Energy Technology Data Exchange (ETDEWEB)

    Iijima, Susumu; Okajima, Shigeaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Perturbation calculation code based on the diffusion theory ''PERKY'' is designed for nuclear characteristic analyses of fast reactor. The code calculates reactivity worth on the multi-group diffusion perturbation theory in two or three dimensional core model and kinetics parameters such as effective delayed neutron fraction, prompt neutron lifetime and absolute reactivity scale factor ({rho}{sub 0} {delta}k/k) for FCA experiments. (author)

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

    International Nuclear Information System (INIS)

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

    1994-07-01

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

  17. HETERO code, heterogeneous procedure for reactor calculation

    International Nuclear Information System (INIS)

    Jovanovic, S.M.; Raisic, N.M.

    1966-11-01

    This report describes the procedure for calculating the parameters of heterogeneous reactor system taking into account the interaction between fuel elements related to established geometry. First part contains the analysis of single fuel element in a diffusion medium, and criticality condition of the reactor system described by superposition of elements interactions. the possibility of performing such analysis by determination of heterogeneous system lattice is described in the second part. Computer code HETERO with the code KETAP (calculation of criticality factor η n and flux distribution) is part of this report together with the example of RB reactor square lattice

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

    International Nuclear Information System (INIS)

    Fujieda, Hirobumi; Murakami, Yoshiki; Sugihara, Masayoshi.

    1992-11-01

    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)

  19. Development of M3C code for Monte Carlo reactor physics criticality calculations

    International Nuclear Information System (INIS)

    Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

    2015-06-01

    The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

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

  1. GNASH: a preequilibrium, statistical nuclear-model code for calculation of cross sections and emission spectra

    International Nuclear Information System (INIS)

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

    1977-11-01

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

  2. Direct calculation of current drive efficiency in FISIC code

    International Nuclear Information System (INIS)

    Wright, J.C.; Phillips, C.K.; Bonoli, P.T.

    1996-01-01

    Two-dimensional RF modeling codes use a parameterization (1) of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform quasi-linear diffusion coefficient and requires a priori knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by the full wave code, FISIC (2). Current profiles are calculated using the adjoint formulation (3). Comparisons between the two formulations are presented. copyright 1996 American Institute of Physics

  3. Audit calculations of accidents analysis for second unit of Ignalina NPP with ATHLET code

    International Nuclear Information System (INIS)

    Adomavicius, A.; Belousov, A.; Ognerubov, V.

    2004-01-01

    Background of thermo hydraulic processes audit calculations in the frame of RSR-2 project is presented. Assumptions for the design based accident - RBMK-1500 group distributor header break analysis and modeling are presented. Audit calculations by ATHLET code and evaluation of results were provided. (author)

  4. Thermal hydraulic calculation of STORM facility using GOTHIC code

    International Nuclear Information System (INIS)

    Pevec, D.; Grgic, D.; Prah, M.

    1995-01-01

    Benchmark calculation CTI defined in frame of STORM experimental programme is used to prove that the GOTHIC code is capable to predict behaviour of experimental facility with reasonable accuracy. GOTHIC code is developed mainly for containment calculation. In this situation it is successfully used for calculation of one dimensional flow of steam and noncondensable mixture. Steady state distributions of pressure, temperature and the velocity of gas along facility are consistent with results obtained by other benchmark participants. (author)

  5. Post-processing of the TRAC code's results

    International Nuclear Information System (INIS)

    Baron, J.H.; Neuman, D.

    1987-01-01

    The TRAC code serves for the analysis of accidents in nuclear installations from the thermohydraulic point of view. A program has been developed with the aim of processing the information rapidly generated by the code, with screening graph capacity, both in high and low resolution, or either in paper through printer or plotter. Although the programs are intended to be used after the TRAC runs, they may be also used even when the program is running so as to observe the calculation process. The advantages of employing this type of tool, its actual capacity and its possibilities of expansion according to the user's needs are herein described. (Author)

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

    International Nuclear Information System (INIS)

    Morel, J.E.

    1987-01-01

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

  7. Application of fuel management calculation codes for CANDU reactor

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun

    2003-01-01

    Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III

  8. Consistent Code Qualification Process and Application to WWER-1000 NPP

    International Nuclear Information System (INIS)

    Berthon, A.; Petruzzi, A.; Giannotti, W.; D'Auria, F.; Reventos, F.

    2006-01-01

    Calculation analysis by application of the system codes are performed to evaluate the NPP or the facility behavior during a postulated transient or to evaluate the code capability. The calculation analysis constitutes a process that involves the code itself, the data of the reference plant, the data about the transient, the nodalization, and the user. All these elements affect one each other and affect the results. A major issue in the use of mathematical model is constituted by the model capability to reproduce the plant or facility behavior under steady state and transient conditions. These aspects constitute two main checks that must be satisfied during the qualification process. The first of them is related to the realization of a scheme of the reference plant; the second one is related to the capability to reproduce the transient behavior. The aim of this paper is to describe the UMAE (Uncertainty Method based on Accuracy Extrapolation) methodology developed at University of Pisa for qualifying a nodalization and analysing the calculated results and to perform the uncertainty evaluation of the system code by the CIAU code (Code with the capability of Internal Assessment of Uncertainty). The activity consists with the re-analysis of the Experiment BL-44 (SBLOCA) performed in the LOBI facility and the analysis of a Kv-scaling calculation of the WWER-1000 NPP nodalization taking as reference the test BL-44. Relap5/Mod3.3 has been used as thermal-hydraulic system code and the standard procedure adopted at University of Pisa has been applied to show the capability of the code to predict the significant aspects of the transient and to obtain a qualified nodalization of the WWER-1000 through a systematic qualitative and quantitative accuracy evaluation. The qualitative accuracy evaluation is based on the selection of Relevant Thermal-hydraulic Aspects (RTAs) and is a prerequisite to the application of the Fast Fourier Transform Based Method (FFTBM) which quantifies

  9. Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation

    International Nuclear Information System (INIS)

    Keco, M.; Debrecin, N.; Grgic, D.

    2005-01-01

    Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-02-01

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

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

    International Nuclear Information System (INIS)

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

    1979-02-01

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

  12. SWAT4.0 - The integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

    International Nuclear Information System (INIS)

    Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki

    2015-03-01

    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)

  13. Calculation of neutron spectra produced in neutron generator target: Code testing.

    Science.gov (United States)

    Gaganov, V V

    2018-03-01

    DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. Verification of 3-D generation code package for neutronic calculations of WWERs

    International Nuclear Information System (INIS)

    Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.

    2000-01-01

    Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)

  15. Description of the CAREM Reactor Neutronic Calculation Codes

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel

    2000-01-01

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

  16. SIMCRI: a simple computer code for calculating nuclear criticality parameters

    International Nuclear Information System (INIS)

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

    1986-03-01

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

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

    International Nuclear Information System (INIS)

    Fink, J.K.

    1979-12-01

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

  18. Development of a BWR core burn-up calculation code COREBN-BWR

    International Nuclear Information System (INIS)

    Morimoto, Yuichi; Okumura, Keisuke

    1992-05-01

    In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

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

    International Nuclear Information System (INIS)

    Nowakowski, Pedro Mariano

    2004-01-01

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

  20. Procedure and code for calculating black control rods taking into account epithermal absorption, code CAS-1

    International Nuclear Information System (INIS)

    Martinc, R.; Trivunac, N.; Zivkovic, Z.

    1964-12-01

    This report describes the computer code CAS-1, calculation method and procedure applied for calculating the black control rods taking into account the epithermal neutron absorption. Results obtained for supercell method applied for regular lattice reflected in the multiplication medium is part of this report in addition to the computer code manual

  1. Research on Primary Shielding Calculation Source Generation Codes

    Science.gov (United States)

    Zheng, Zheng; Mei, Qiliang; Li, Hui; Shangguan, Danhua; Zhang, Guangchun

    2017-09-01

    Primary Shielding Calculation (PSC) plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF) for the source particle sample code of the J Monte Carlo Transport (JMCT) code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three dimensional (3D) power distributions in xyz geometry to source distributions in r θ z geometry for the J Discrete Ordinate Transport (JSNT) code. Validation on PSC model of Qinshan No.1 nuclear power plant (NPP), CAP1400 and CAP1700 reactors are performed. Numerical results show that the theoretical model and the codes are both correct.

  2. Code Betal to calculation Alpha/Beta activities in environmental samples

    International Nuclear Information System (INIS)

    Romero, L.; Travesi, A.

    1983-01-01

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

  3. EMPIRE-II statistical model code for nuclear reaction calculations

    Energy Technology Data Exchange (ETDEWEB)

    Herman, M [International Atomic Energy Agency, Vienna (Austria)

    2001-12-15

    EMPIRE II is a nuclear reaction code, comprising various nuclear models, and designed for calculations in the broad range of energies and incident particles. A projectile can be any nucleon or Heavy Ion. The energy range starts just above the resonance region, in the case of neutron projectile, and extends up to few hundreds of MeV for Heavy Ion induced reactions. The code accounts for the major nuclear reaction mechanisms, such as optical model (SCATB), Multistep Direct (ORION + TRISTAN), NVWY Multistep Compound, and the full featured Hauser-Feshbach model. Heavy Ion fusion cross section can be calculated within the simplified coupled channels approach (CCFUS). A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers (BARFIT), moments of inertia (MOMFIT), and {gamma}-ray strength functions. Effects of the dynamic deformation of a fast rotating nucleus can be taken into account in the calculations. The results can be converted into the ENDF-VI format using the accompanying code EMPEND. The package contains the full EXFOR library of experimental data. Relevant EXFOR entries are automatically retrieved during the calculations. Plots comparing experimental results with the calculated ones can be produced using X4TOC4 and PLOTC4 codes linked to the rest of the system through bash-shell (UNIX) scripts. The graphic user interface written in Tcl/Tk is provided. (author)

  4. INLUX-DBR - A calculation code to calculate indoor natural illuminance inside buildings under various sky conditions

    International Nuclear Information System (INIS)

    Ferraro, V.; Igawa, N.; Marinelli, V.

    2010-01-01

    A calculation code, named INLUX-DBR, is presented, which is a modified version of INLUX code, able to predict the illuminance distribution on the inside surfaces of a room with six walls and a window, and on the work plane. At each desired instant the code solves the system of the illuminance equations of each surface element, characterized by the latter's reflection coefficient and its view factors toward the other elements. In the model implemented in the code, the sky-diffuse luminance distribution, the sun beam light and the light reflected from the ground toward the room are considered. The code was validated by comparing the calculated values of illuminance with the experimental values measured inside a scale model (1:5) of a building room, in various sky conditions of overcast, clear and intermediate days. The validation is performed using the sky luminance data measured by a sky scanner and the measured beam illuminance of the sun as input data. A comparative analysis of some of the well-known calculation models of sky luminance, namely Perez, Igawa and CIE models was also carried out, comparing the code predictions and the measured values of inside illuminance in the scale model.

  5. Parameter calculation tool for the application of radiological dose projection codes

    International Nuclear Information System (INIS)

    Galindo G, I. F.; Vergara del C, J. A.; Galvan A, S. J.; Tijerina S, F.

    2016-09-01

    The use of specialized codes to estimate the radiation dose projection to an emergency postulated event at a nuclear power plant requires that certain plant data be available according to the event being simulated. The calculation of the possible radiological release is the critical activity to carry out the emergency actions. However, not all of the plant data required are obtained directly from the plant but need to be calculated. In this paper we present a computational tool that calculates the plant data required to use the radiological dose estimation codes. The tool provides the required information when there is a gas emergency venting event in the primary containment atmosphere, whether well or dry well and also calculates the time in which the spent fuel pool would be discovered in the event of a leak of water on some of the walls or floor of the pool. The tool developed has mathematical models for the processes involved such as: compressible flow in pipes considering area change and for constant area, taking into account the effects of friction and for the case of the spent fuel pool hydraulic models to calculate the time in which a container is emptied. The models implemented in the tool are validated with data from the literature for simulated cases. The results with the tool are very similar to those of reference. This tool will also be very supportive so that in postulated emergency cases can use the radiological dose estimation codes to adequately and efficiently determine the actions to be taken in a way that affects as little as possible. (Author)

  6. Detailed resonance absorption calculations with the Monte Carlo code MCNP and collision probability version of the slowing down code ROLAIDS

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Janssen, A.J.

    1994-01-01

    Very accurate Mote Carlo calculations with Monte Carlo Code have been performed to serve as reference for benchmark calculations on resonance absorption by U 238 in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (authors). 9 refs., 5 figs., 2 tabs

  7. FAST CALCULATION OF THE LOMB-SCARGLE PERIODOGRAM USING GRAPHICS PROCESSING UNITS

    International Nuclear Information System (INIS)

    Townsend, R. H. D.

    2010-01-01

    I introduce a new code for fast calculation of the Lomb-Scargle periodogram that leverages the computing power of graphics processing units (GPUs). After establishing a background to the newly emergent field of GPU computing, I discuss the code design and narrate key parts of its source. Benchmarking calculations indicate no significant differences in accuracy compared to an equivalent CPU-based code. However, the differences in performance are pronounced; running on a low-end GPU, the code can match eight CPU cores, and on a high-end GPU it is faster by a factor approaching 30. Applications of the code include analysis of long photometric time series obtained by ongoing satellite missions and upcoming ground-based monitoring facilities, and Monte Carlo simulation of periodogram statistical properties.

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

    International Nuclear Information System (INIS)

    Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

    2002-01-01

    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)

  9. Development of system analysis code for pyrochemical process using molten salt electrorefining

    International Nuclear Information System (INIS)

    Tozawa, K.; Matsumoto, T.; Kakehi, I.

    2000-04-01

    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 development of an 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. The cathode processor calculation code with distillation process was developed. A code validation calculation has been conducted on the basic of the benchmark problem for natural convection in a square cavity. Results by using the present code agreed well for the velocity-temperature fields, the maximum velocity and its location with the benchmark solution published in a paper. The functions have been added to advance the reality in simulation and to increase the efficiency in utilization. The test run has been conducted using the code with the above modification for an axisymmetric enclosed vessel simulating a cathode processor, and the capability of the distillation process simulation with the code has been confirmed. An analytical model for cooling behavior of the pyrochemical process cell was developed. The analytical model was selected by comparing benchmark analysis with detailed analysis on engineering workstation. Flow and temperature distributions were confirmed by the result of steady state analysis. In the result of transient cooling analysis, an initial transient peak of temperature occurred at balanced heat condition in the steady-state analysis. Final gas temperature distribution was dependent on gas circulation flow in transient condition. Then there were different final gas temperature distributions on the basis of the result of steady-state analysis. This phenomenon has a potential for it's own metastable condition. Therefore it was necessary to design gas cooling flow pattern without cooling gas circulation

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

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Naito, Yoshitaka; Yamakawa, Yasuhiro.

    1980-11-01

    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(NO 3 ) 4 aqueous solution, Pu metal or PuO 2 -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)

  11. INLUX-DBR - A calculation code to calculate indoor natural illuminance inside buildings under various sky conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ferraro, V.; Igawa, N.; Marinelli, V. [Mechanical Engineering Department, University of Calabria, 87036 Arcavacata di Rende (CS) (Italy)

    2010-09-15

    A calculation code, named INLUX-DBR, is presented, which is a modified version of INLUX code, able to predict the illuminance distribution on the inside surfaces of a room with six walls and a window, and on the work plane. At each desired instant the code solves the system of the illuminance equations of each surface element, characterized by the latter's reflection coefficient and its view factors toward the other elements. In the model implemented in the code, the sky-diffuse luminance distribution, the sun beam light and the light reflected from the ground toward the room are considered. The code was validated by comparing the calculated values of illuminance with the experimental values measured inside a scale model (1:5) of a building room, in various sky conditions of overcast, clear and intermediate days. The validation is performed using the sky luminance data measured by a sky scanner and the measured beam illuminance of the sun as input data. A comparative analysis of some of the well-known calculation models of sky luminance, namely Perez, Igawa and CIE models was also carried out, comparing the code predictions and the measured values of inside illuminance in the scale model. (author)

  12. Discrete-ordinates electron transport calculations using standard neutron transport codes

    International Nuclear Information System (INIS)

    Morel, J.E.

    1979-01-01

    The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

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

    International Nuclear Information System (INIS)

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

    1980-04-01

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

  14. Methodology of personnel exposure calculation and optimisation within the decommissioning planning code OMEGA

    International Nuclear Information System (INIS)

    Vasko, Marek; Daniska, Vladimir; Rehak, Ivan; Necas, Vladimir

    2011-01-01

    Calculation of personnel exposure is a one of the main parameters being evaluated within the pre-decommissioning plans together with other decommissioning drivers such as costs, manpower, amounts of RAW and conventional waste and amount of discharged gaseous and liquid effluents. Alongside with manpower, the exposure is an indicator of the decommissioning process for need of staff, and quantifies impact of decommissioning on personnel from the radio hygienic point of view. At the same time it indicates suitability of individual work procedures use for decommissioning activities. For this reason it is important to estimate as precise as possible demands on personnel exposure even during preparatory decommissioning phase to quantify impact of decommissioning on personnel and eventually optimize the decommissioning process, if needed. The most appropriate way of staff exposure estimation during decommissioning preparatory phases is its calculation based on radiological and physical characteristics of equipment to be decommissioned and also quantitative and qualitative characterisation of typical decommissioning activities. On one hand, the methodology of exposure calculation should allow as much as possible realistic description and algorithmisation of exposure ways during decommissioning activities. On the other hand the calculation have to be systematic, well-arranged and clearly definable by appropriate mathematic relations. Calculation can be made by various approaches using more or less sophisticated software solutions from classic MS Excel sheets up to the complex calculation codes. In this paper, a methodology used for personnel exposure calculation and optimization implemented within the complex computer code OMEGA developed at DECOM, a.s. is described. (author)

  15. Burnup calculation methodology in the serpent 2 Monte Carlo code

    International Nuclear Information System (INIS)

    Leppaenen, J.; Isotalo, A.

    2012-01-01

    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)

  16. Refuelling design and core calculations at NPP Paks: codes and methods

    International Nuclear Information System (INIS)

    Pos, I.; Nemes, I.; Javor, E.; Korpas, L.; Szecsenyi, Z.; Patai-Szabo, S.

    2001-01-01

    This article gives a brief review of the computer codes used in the fuel management practice at NPP Paks. The code package consist of the HELIOS neutron and gamma transport code for preparation of few-group cross section library, the CERBER code to determine the optimal core loading patterns and the C-PORCA code for detailed reactor physical analysis of different reactor states. The last two programs have been developed at the NPP Paks. HELIOS gives sturdy basis for our neutron physical calculation, CERBER and C-PORCA programs have been enhanced in great extent for last years. Methods and models have become more detailed and accurate as regards the calculated parameters and space resolution. Introduction of a more advanced data handling algorithm arbitrary move of fuel assemblies can be followed either in the reactor core or storage pool. The new interactive WINDOWS applications allow easier and more reliable use of codes. All these computer code developments made possible to handle and calculate new kind of fuels as profiled Russian and BNFL fuel with burnable poison or to support the reliable reuse of fuel assemblies stored in the storage pool. To extend thermo-hydraulic capability, with KFKI contribution the COBRA code will also be coupled to the system (Authors)

  17. The calculation of coolant leak rate through the cracks using RELAP5 code

    International Nuclear Information System (INIS)

    Krungeleviciute, V.; Kaliatka, A.

    2001-01-01

    For reason to choose method of leak detection first of all it is necessary to perform evaluating thermal-hydraulic calculations. These calculations allow to determine flow rate of discharged coolant. For coolant leak rate calculations through possible cracks in Ignalina NPP pipes SQUIRT and RELAP5 thermal-hydraulic codes were used. SQUIRT is well known as computer program that predicts the leakage for cracked pipes in NPP. As this code calculates only water (at subcooled or saturated conditions) leak rate, RELAP5 code model, that calculates water and steam leak rate, was created. For model validation comparison of SQUIRT, RELAP5 and experimental results was performed. Analysis shows RELAP5 code model suitability for calculations of leak through through-wall cracks in pipes. (author)

  18. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    Suyama, Kenya; Hirakawa, Naohiro; Iwasaki, Tomohiko.

    1997-11-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

  19. Kinetic parameters evaluation of PWRs using static cell and core calculation codes

    International Nuclear Information System (INIS)

    Jahanbin, Ali; Malmir, Hessam

    2012-01-01

    Highlights: ► In this study, we have calculated effective delayed neutron fraction and prompt neutron lifetime in PWRs. ► New software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. ► This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. ► The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. - Abstract: In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs, using static cell and core calculation codes, is reported. A new software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. Using the WIMS cell calculation code, multigroup microscopic cross-sections and number densities of different materials can be generated in a binary file. By the use of BORGES code, these binary-form cross-sections and number densities are converted to a format readable by the CITATION core calculation code, by which the kinetic parameters can be finally obtained. This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. Benchmarking of the results against the final safety analysis report (FSAR) of the aforementioned reactors shows very good agreements with these published documents.

  20. Parallel processing of Monte Carlo code MCNP for particle transport problem

    Energy Technology Data Exchange (ETDEWEB)

    Higuchi, Kenji; Kawasaki, Takuji

    1996-06-01

    It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)

  1. Mathematical simulation of processes in horizontal steam generator and the program of calculation of its characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Titov, V.F.; Zorin, V.M.; Gorburov, V.I. [OKB Gidropress, Moscow Energy Inst. (Russian Federation)

    1995-12-31

    On the basis of mathematical models describing the processes in horizontal steam generator (SG) the code giving the possibility to calculate the hydrodynamical characteristics in any point of water volume, has been developed. The code simulates the processes in SG in the stationary (or quasi-stationary) mode or operation only. The code may be used as a next step to calculations of the SG characteristics in the non-stationary modes of operation.

  2. Mathematical simulation of processes in horizontal steam generator and the program of calculation of its characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Titov, V F; Zorin, V M; Gorburov, V I [OKB Gidropress, Moscow Energy Inst. (Russian Federation)

    1996-12-31

    On the basis of mathematical models describing the processes in horizontal steam generator (SG) the code giving the possibility to calculate the hydrodynamical characteristics in any point of water volume, has been developed. The code simulates the processes in SG in the stationary (or quasi-stationary) mode or operation only. The code may be used as a next step to calculations of the SG characteristics in the non-stationary modes of operation.

  3. Development of throughflow calculation code for axial flow compressors

    International Nuclear Information System (INIS)

    Kim, Ji Hwan; Kim, Hyeun Min; No, Hee Cheon

    2005-01-01

    The power conversion systems of the current HTGRs are based on closed Brayton cycle and major concern is thermodynamic performance of the axial flow helium gas turbines. Particularly, the helium compressor has some unique design challenges compared to the air-breathing compressor such as high hub-to-tip ratios throughout the machine and a large number of stages due to the physical property of the helium and thermodynamic cycle. Therefore, it is necessary to develop a design and analysis code for helium compressor that can estimate the design point and off-design performance accurately. KAIST nuclear system laboratory has developed a compressor design and analysis code by means of throughflow calculation and several loss models. This paper presents the outline of the development of a throughflow calculation code and its verification results

  4. Fuel behaviour calculations with version 2.0 of the code FUROM

    International Nuclear Information System (INIS)

    Kulacsy, K.

    2011-01-01

    The fuel modelling code FUROM (FUel ROd Model), suitable for calculating the normal operation condition behaviour of PWR and WWER fuels, has been developed at AEKI for several years. In 2010 the new version of the code, FUROM-2.0 was released. Calculations performed with this version and results are presented. (author)

  5. Perspective on the audit calculation for SFR using TRACE code

    Energy Technology Data Exchange (ETDEWEB)

    Shin, An Dong; Choi, Yong Won; Bang, Young Suk; Bae, Moo Hoon; Huh, Byung Gil; Seol, Kwang One [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    Korean Sodium Cooled Fast Reactor (SFR) is being developed by KAERI. The Prototype SFR will be a first SFR applied for licensing. KINS started research programs for preparing new concept design licensing recently. Safety analysis for the certain reactor is based on the computational estimation with conservatism and/or uncertainty of modeling. For the audit calculation for sodium cooled fast reactor (SFR), TRACE code is considered as one of analytical tool for SFR since TRACE code have already sodium related properties and models in it and have experience in the liquid metal coolant system area in abroad. Applicability of TRACE code for SFR is prechecked before real audit calculation. In this study, Demonstration Fast Reactor (DFR) 600 steady state conditions is simulated for identification of area of modeling improvements of TRACE code.

  6. User effects on the transient system code calculations. Final report

    International Nuclear Information System (INIS)

    Aksan, S.N.; D'Auria, F.

    1995-01-01

    Large thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are accomplished by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies has the influence of the code user on the calculated results been directly addressed. In this paper, the results of the investigations on the user effects for the thermal-hydraulic transient system codes is presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects are discussed in detail and general recommendations and conclusions are presented to control and limit them

  7. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    International Nuclear Information System (INIS)

    Ahnert, C.; Arayones, J.M.

    1985-01-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

  8. Hamor-2: a computer code for LWR inventory calculation

    International Nuclear Information System (INIS)

    Guimaraes, L.N.F.; Marzo, M.A.S.

    1985-01-01

    A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt

  9. Optical model calculations with the code ECIS95

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, B V [Departamento de Fisica, Instituto Tecnologico da Aeronautica, Centro Tecnico Aeroespacial (Brazil)

    2001-12-15

    The basic features of elastic and inelastic scattering within the framework of the spherical and deformed nuclear optical models are discussed. The calculation of cross sections, angular distributions and other scattering quantities using J. Raynal's code ECIS95 is described. The use of the ECIS method (Equations Couplees en Iterations Sequentielles) in coupled-channels and distorted-wave Born approximation calculations is also reviewed. (author)

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

    International Nuclear Information System (INIS)

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

    1982-10-01

    This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc

  11. Data calculation program for RELAP 5 code

    International Nuclear Information System (INIS)

    Silvestre, Larissa J.B.; Sabundjian, Gaiane

    2015-01-01

    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)

  12. Data calculation program for RELAP 5 code

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  13. OLIFE: Tight Binding Code for Transmission Coefficient Calculation

    Science.gov (United States)

    Mijbil, Zainelabideen Yousif

    2018-05-01

    A new and human friendly transport calculation code has been developed. It requires a simple tight binding Hamiltonian as the only input file and uses a convenient graphical user interface to control calculations. The effect of magnetic field on junction has also been included. Furthermore the transmission coefficient can be calculated between any two points on the scatterer which ensures high flexibility to check the system. Therefore Olife can highly be recommended as an essential tool for pretesting studying and teaching electron transport in molecular devices that saves a lot of time and effort.

  14. Method for calculating internal radiation and ventilation with the ADINAT heat-flow code

    International Nuclear Information System (INIS)

    Butkovich, T.R.; Montan, D.N.

    1980-01-01

    One objective of the spent fuel test in Climax Stock granite (SFTC) is to correctly model the thermal transport, and the changes in the stress field and accompanying displacements from the application of the thermal loads. We have chosen the ADINA and ADINAT finite element codes to do these calculations. ADINAT is a heat transfer code compatible to the ADINA displacement and stress analysis code. The heat flow problem encountered at SFTC requires a code with conduction, radiation, and ventilation capabilities, which the present version of ADINAT does not have. We have devised a method for calculating internal radiation and ventilation with the ADINAT code. This method effectively reproduces the results from the TRUMP multi-dimensional finite difference code, which correctly models radiative heat transport between drift surfaces, conductive and convective thermal transport to and through air in the drifts, and mass flow of air in the drifts. The temperature histories for each node in the finite element mesh calculated with ADINAT using this method can be used directly in the ADINA thermal-mechanical calculation

  15. CITOPP, CITMOD, CITWI, Processing codes for CITATION Code

    International Nuclear Information System (INIS)

    Albarhoum, M.

    2008-01-01

    Description of program or function: CITOPP processes the output file of the CITATION 3-D diffusion code. The program can plot axial, radial and circumferential flux distributions (in cylindrical geometry) in addition to the multiplication factor convergence. The flux distributions can be drawn for each group specified by the program and visualized on the screen. CITMOD processes both the output and the input files of the CITATION 3-D diffusion code. CITMOD can visualize both axial, and radial-angular models of the reactor described by CITATION input/output files. CITWI processes the input file (CIT.INP) of CITATION 3-D diffusion code. CIT.INP is processed to deduce the dimensions of the cell whose cross sections can be representative of the homonym reactor component in section 008 of CIT.INP

  16. Introduction to reactor lattice calculations by the WIMSD code

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1998-01-01

    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)

  17. Two-dimensional core calculation research for fuel management optimization based on CPACT code

    International Nuclear Information System (INIS)

    Chen Xiaosong; Peng Lianghui; Gang Zhi

    2013-01-01

    Fuel management optimization process requires rapid assessment for the core layout program, and the commonly used methods include two-dimensional diffusion nodal method, perturbation method, neural network method and etc. A two-dimensional loading patterns evaluation code was developed based on the three-dimensional LWR diffusion calculation program CPACT. Axial buckling introduced to simulate the axial leakage was searched in sub-burnup sections to correct the two-dimensional core diffusion calculation results. Meanwhile, in order to get better accuracy, the weight equivalent volume method of the control rod assembly cross-section was improved. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Krepper, E

    1998-10-01

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

  19. Theoretical calculation possibilities of the computer code HAMMER

    International Nuclear Information System (INIS)

    Onusic Junior, J.

    1978-06-01

    With the aim to know the theoretical calculation possibilities of the computer code HAMMER, developed at Savanah River Laboratory, a analysis of the crytical cells assembly of the kind utilized in PWR reactors is made. (L.F.S.) [pt

  20. Parallel processing of two-dimensional Sn transport calculations

    International Nuclear Information System (INIS)

    Uematsu, M.

    1997-01-01

    A parallel processing method for the two-dimensional S n transport code DOT3.5 has been developed to achieve a drastic reduction in computation time. In the proposed method, parallelization is achieved with angular domain decomposition and/or space domain decomposition. The calculational speed of parallel processing by angular domain decomposition is largely influenced by frequent communications between processing elements. To assess parallelization efficiency, sample problems with up to 32 x 32 spatial meshes were solved with a Sun workstation using the PVM message-passing library. As a result, parallel calculation using 16 processing elements, for example, was found to be nine times as fast as that with one processing element. As for parallel processing by geometry segmentation, the influence of processing element communications on computation time is small; however, discontinuity at the segment boundary degrades convergence speed. To accelerate the convergence, an alternate sweep of angular flux in conjunction with space domain decomposition and a two-step rescaling method consisting of segmentwise rescaling and ordinary pointwise rescaling have been developed. By applying the developed method, the number of iterations needed to obtain a converged flux solution was reduced by a factor of 2. As a result, parallel calculation using 16 processing elements was found to be 5.98 times as fast as the original DOT3.5 calculation

  1. Implementation of a dry process fuel cycle model into the DYMOND code

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Jeong, Chang Joon; Choi, Hang Bok

    2004-01-01

    For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada Deuterium Uranium (CANDU) reactor, direct use of spent Pressurized Water Reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-through and DUPIC fuel cycles

  2. Verification and validation of XSDRNPM code for tank waste calculations

    International Nuclear Information System (INIS)

    ROGERS, C.A.

    1999-01-01

    This validation study demonstrates that the XSDRNPM computer code accurately calculates the infinite neutron multiplication for water-moderated systems of low enriched uranium, plutonium, and iron. Calculations are made on a 200 MHz Brvo MS 5200M personal

  3. Comparison of calculations of a reflected reactor with diffusion, SN and Monte Carlo codes

    International Nuclear Information System (INIS)

    McGregor, B.

    1975-01-01

    A diffusion theory code, POW, was compared with a Monte Carlo transport theory code, KENO, for the calculation of a small C/ 235 U cylindrical core with a graphite reflector. The calculated multiplication factors were in good agreement but differences were noted in region-averaged group fluxes. A one-dimensional spherical geometry was devised to approximate cylindrical geometry. Differences similar to those already observed were noted when the region-averaged fluxes from a diffusion theory (POW) calculation were compared with an SN transport theory (ANAUSN) calculation for the spherical model. Calculations made with SN and Monte Carlo transport codes were in good agreement. It was concluded that observed flux differences were attributable to the POW code, and were not inconsistent with inherent diffusion theory approximations. (author)

  4. WIPP Benchmark calculations with the large strain SPECTROM codes

    International Nuclear Information System (INIS)

    Callahan, G.D.; DeVries, K.L.

    1995-08-01

    This report provides calculational results from the updated Lagrangian structural finite-element programs SPECTROM-32 and SPECTROM-333 for the purpose of qualifying these codes to perform analyses of structural situations in the Waste Isolation Pilot Plant (WIPP). Results are presented for the Second WIPP Benchmark (Benchmark II) Problems and for a simplified heated room problem used in a parallel design calculation study. The Benchmark II problems consist of an isothermal room problem and a heated room problem. The stratigraphy involves 27 distinct geologic layers including ten clay seams of which four are modeled as frictionless sliding interfaces. The analyses of the Benchmark II problems consider a 10-year simulation period. The evaluation of nine structural codes used in the Benchmark II problems shows that inclusion of finite-strain effects is not as significant as observed for the simplified heated room problem, and a variety of finite-strain and small-strain formulations produced similar results. The simplified heated room problem provides stratigraphic complexity equivalent to the Benchmark II problems but neglects sliding along the clay seams. The simplified heated problem does, however, provide a calculational check case where the small strain-formulation produced room closures about 20 percent greater than those obtained using finite-strain formulations. A discussion is given of each of the solved problems, and the computational results are compared with available published results. In general, the results of the two SPECTROM large strain codes compare favorably with results from other codes used to solve the problems

  5. The use of the codes from MCU family for calculations of WWER type reactors

    International Nuclear Information System (INIS)

    Abagijan, L.P.; Alexeyev, N.I.; Bryzgalov, V.I.; Gomin, E.A.; Glushkov, A.E.; Gorodkov, S.S.; Gurevich, M.I.; Kalugin, M.A.; Marin, S.V.; Shkarovsky, D.A.; Yudkevich, M.S.

    2000-01-01

    The MCU-RFFI/A and MCU-REA codes developed within the framework of the long term MCU project are widely used for calculations of neutron physic characteristics of WWER type reactors. Complete descriptions of the codes are available in both Russian and English. The codes are verified and validated by means of the comparison of calculated results with experimental data and mathematical benchmarks. The codes are licensed by Russian Nuclear and Criticality Safety Regulatory Body (Gosatomnadzor RF) (Code Passports: N 61 of 17.10.1966 and N 115 of 02.03.2000 accordingly)). The report gives examples of WWER reactor physic tasks important for practice solved using the codes from the MCU family. Some calculational results are given too. (Authors)

  6. PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation

    International Nuclear Information System (INIS)

    Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.

    1995-01-01

    1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-11-01

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

  8. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    International Nuclear Information System (INIS)

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-01-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

  9. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    Science.gov (United States)

    Karim, Julia Abdul

    2008-05-01

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

  10. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    International Nuclear Information System (INIS)

    Karim, Julia Abdul

    2008-01-01

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained

  11. The 1996 ENDF pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1996-01-01

    The codes are named 'the Pre-processing' codes, because they are designed to pre-process ENDF/B data, for later, further processing for use in applications. This is a modular set of computer codes, each of which reads and writes evaluated nuclear data in the ENDF/B format. Each code performs one or more independent operations on the data, as described below. These codes are designed to be computer independent, and are presently operational on every type of computer from large mainframe computer to small personal computers, such as IBM-PC and Power MAC. The codes are available from the IAEA Nuclear Data Section, free of charge upon request. (author)

  12. Effect of difference between group constants processed by codes TIMS and ETOX on integral quantities

    International Nuclear Information System (INIS)

    Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi.

    1978-06-01

    Group constants of 235 U, 238 U, 239 Pu, 240 Pu and 241 Pu have been produced with the processing code TIMS using the evaluated nuclear data of JENDL-1. The temperature and composition dependent self-shielding factors have been calculated for the two cases with and without considering mutual interference resonant nuclei. By using the group constants set produced by the TIMS code, the integral quantities, i.e. multiplication factor, Na-void reactivity effect and Doppler reactivity effect, are calculated and compared with those calculated with the use of the cross sections set produced by the ETOX code to evaluate accuracy of the approximate calculation method in ETOX. There is much difference in self-shielding factor in each energy group between the two codes. For the fast reactor assemblies under study, however, the integral quantities calculated with these two sets are in good agreement with each other, because of eventual cancelation of errors. (auth.)

  13. A NEM diffusion code for fuel management and time average core calculation

    International Nuclear Information System (INIS)

    Mishra, Surendra; Ray, Sherly; Kumar, A.N.

    2005-01-01

    A computer code based on Nodal expansion method has been developed for solving two groups three dimensional diffusion equation. This code can be used for fuel management and time average core calculation. Explicit Xenon and fuel temperature estimation are also incorporated in this code. TAPP-4 phase-B physics experimental results were analyzed using this code and a code based on FD method. This paper gives the comparison of the observed data and the results obtained with this code and FD code. (author)

  14. Comparison of computer code calculations with FEBA test data

    International Nuclear Information System (INIS)

    Zhu, Y.M.

    1988-06-01

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

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

    International Nuclear Information System (INIS)

    2003-01-01

    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.)

  16. Hot particle dose calculations using the computer code VARSKIN Mod 2

    International Nuclear Information System (INIS)

    Durham, J.S.

    1991-01-01

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

  17. Confidence level in the calculations of HCDA consequences using large codes

    International Nuclear Information System (INIS)

    Nguyen, D.H.; Wilburn, N.P.

    1979-01-01

    The probabilistic approach to nuclear reactor safety is playing an increasingly significant role. For the liquid-metal fast breeder reactor (LMFBR) in particular, the ultimate application of this approach could be to determine the probability of achieving the goal of a specific line-of-assurance (LOA). Meanwhile a more pressing problem is one of quantifying the uncertainty in a calculated consequence for hypothetical core disruptive accident (HCDA) using large codes. Such uncertainty arises from imperfect modeling of phenomenology and/or from inaccuracy in input data. A method is presented to determine the confidence level in consequences calculated by a large computer code due to the known uncertainties in input invariables. A particular application was made to the initial time of pin failure in a transient overpower HCDA calculated by the code MELT-IIIA in order to demonstrate the method. A probability distribution function (pdf) for the time of failure was first constructed, then the confidence level for predicting this failure parameter within a desired range was determined

  18. Some benchmark calculations for VVER-1000 assemblies by WIMS-7B code

    International Nuclear Information System (INIS)

    Sultanov, N.V.

    2001-01-01

    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 K eff 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.)

  19. Radiation transport simulation in gamma irradiator systems using E G S 4 Monte Carlo code and dose mapping calculations based on point kernel technique

    International Nuclear Information System (INIS)

    Raisali, G.R.

    1992-01-01

    A series of computer codes based on point kernel technique and also Monte Carlo method have been developed. These codes perform radiation transport calculations for irradiator systems having cartesian, cylindrical and mixed geometries. The monte Carlo calculations, the computer code 'EGS4' has been applied to a radiation processing type problem. This code has been acompanied by a specific user code. The set of codes developed include: GCELLS, DOSMAPM, DOSMAPC2 which simulate the radiation transport in gamma irradiator systems having cylinderical, cartesian, and mixed geometries, respectively. The program 'DOSMAP3' based on point kernel technique, has been also developed for dose rate mapping calculations in carrier type gamma irradiators. Another computer program 'CYLDETM' as a user code for EGS4 has been also developed to simulate dose variations near the interface of heterogeneous media in gamma irradiator systems. In addition a system of computer codes 'PRODMIX' has been developed which calculates the absorbed dose in the products with different densities. validation studies of the calculated results versus experimental dosimetry has been performed and good agreement has been obtained

  20. The concerted calculation of the BN-600 reactor for the deterministic and stochastic codes

    Science.gov (United States)

    Bogdanova, E. V.; Kuznetsov, A. N.

    2017-01-01

    The solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC “Kurchatov Institute” in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.

  1. Calculation of static harmonics of a nuclear reactor using CITATION code

    International Nuclear Information System (INIS)

    Belchior Junior, A.; Moreira, J.M.L.

    1989-01-01

    The CITATION code, which solves the multigroup diffusion equation by the finite difference method, calculates the fundamental λ-mode (harmonic) for nuclear reactors. In this work, two fission source correction methods are attempted to obtain higher λ-modes through the CITATION code. The two methods are compared, their advantages and disadvantages analysed and verified against analytical solutions. Two dimensional harmonic modes are calculated for the IEA-R1 research reactor and for the ANGRA-I power reactor. The results are shown in graphics and tables. (author) [pt

  2. THIDA: code system for calculation of the exposure dose rate around a fusion device

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Igarashi, Masahito.

    1978-12-01

    A code system THIDA has been developed for calculation of the exposure dose rates around a fusion device. It consists of the following: one- and two-dimensional discrete ordinate transport codes; induced activity calculation code; activation chain, activation cross section, radionuclide gamma-ray energy/intensity and gamma-ray group constant files; and gamma ray flux to exposure dose rate conversion coefficients. (author)

  3. PWR core follow calculations using the ELCOS code system

    International Nuclear Information System (INIS)

    Grimm, P.; Paratte, J.M.

    1990-01-01

    The ELCOS code system developed at PSI is used to simulate a cycle of a PWR in which one fifth of the assemblies are MOX fuel. The reactor and the calculational methods are briefly described. The calculated critical boron concentrations and power distributions are compared with the measurements at the plant. Although the critical boron concentration is somewhat overpredicted and the computed power distributions are slightly flatter than the measured ones the results of the calculations agree generally well with the measured data. (author) 1 tab., 8 figs., 6 refs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    Vitruk, S.G.; Korsun, A.S.; Ushakov, P.A.

    1995-01-01

    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

  6. Recent development for the ITS code system: Parallel processing and visualization

    International Nuclear Information System (INIS)

    Fan, W.C.; Turner, C.D.; Halbleib, J.A. Sr.; Kensek, R.P.

    1996-01-01

    A brief overview is given for two software developments related to the ITS code system. These developments provide parallel processing and visualization capabilities and thus allow users to perform ITS calculations more efficiently. Timing results and a graphical example are presented to demonstrate these capabilities

  7. The 1989 ENDF pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.; McLaughlin, P.K.

    1989-12-01

    This document summarizes the 1989 version of the ENDF pre-processing codes which are required for processing evaluated nuclear data coded in the format ENDF-4, ENDF-5, or ENDF-6. The codes are available from the IAEA Nuclear Data Section, free of charge upon request. (author)

  8. Computer code selection criteria for flow and transport code(s) to be used in undisturbed vadose zone calculations for TWRS environmental analyses

    International Nuclear Information System (INIS)

    Mann, F.M.

    1998-01-01

    The Tank Waste Remediation System (TWRS) is responsible for the safe storage, retrieval, and disposal of waste currently being held in 177 underground tanks at the Hanford Site. In order to successfully carry out its mission, TWRS must perform environmental analyses describing the consequences of tank contents leaking from tanks and associated facilities during the storage, retrieval, or closure periods and immobilized low-activity tank waste contaminants leaving disposal facilities. Because of the large size of the facilities and the great depth of the dry zone (known as the vadose zone) underneath the facilities, sophisticated computer codes are needed to model the transport of the tank contents or contaminants. This document presents the code selection criteria for those vadose zone analyses (a subset of the above analyses) where the hydraulic properties of the vadose zone are constant in time the geochemical behavior of the contaminant-soil interaction can be described by simple models, and the geologic or engineered structures are complicated enough to require a two-or three dimensional model. Thus, simple analyses would not need to use the fairly sophisticated codes which would meet the selection criteria in this document. Similarly, those analyses which involve complex chemical modeling (such as those analyses involving large tank leaks or those analyses involving the modeling of contaminant release from glass waste forms) are excluded. The analyses covered here are those where the movement of contaminants can be relatively simply calculated from the moisture flow. These code selection criteria are based on the information from the low-level waste programs of the US Department of Energy (DOE) and of the US Nuclear Regulatory Commission as well as experience gained in the DOE Complex in applying these criteria. Appendix table A-1 provides a comparison between the criteria in these documents and those used here. This document does not define the models (that

  9. Comparison of thick-target (alpha,n yield calculation codes

    Directory of Open Access Journals (Sweden)

    Fernandes Ana C.

    2017-01-01

    Full Text Available Neutron production yields and energy distributions from (α,n reactions in light elements were calculated using three different codes (SOURCES, NEDIS and USD and compared with the existing experimental data in the 3.5-10 MeV alpha energy range. SOURCES and NEDIS display an agreement between calculated and measured yields in the decay series of 235U, 238U and 232Th within ±10% for most materials. The discrepancy increases with alpha energy but still an agreement of ±20% applies to all materials with reliable elemental production yields (the few exceptions are identified. The calculated neutron energy distributions describe the experimental data, with NEDIS retrieving very well the detailed features. USD generally underestimates the measured yields, in particular for compounds with heavy elements and/or at high alpha energies. The energy distributions exhibit sharp peaks that do not match the observations. These findings may be caused by a poor accounting of the alpha particle energy loss by the code. A big variability was found among the calculated neutron production yields for alphas from Sm decay; the lack of yield measurements for low (~2 MeV alphas does not allow to conclude on the codes’ accuracy in this energy region.

  10. The Calculation of Flooding Level using CFX Code

    International Nuclear Information System (INIS)

    Oh, Seo Bin; Kim, Keon Yeop; Lee, Hyung Ho

    2015-01-01

    The plant design should consider internal flooding by postulated pipe ruptures, component failures, actuation of spray systems, and improper system alignment. The flooding causes failure of safety-related equipment and affects the integrity of the structure. The safety-related equipment should be installed above the flood level for protection against flooding effects. Conservative estimates of the flood level are important when a DBA occurs. The flooding level can be calculated simply applying Bernoulli's equation. However, in this study, a realistic calculation is performed with ANSYS CFX code. In calculation with CFX, air-core vortex phenomena, and turbulent flow can be simulated, which cannot be calculated analytically. The flooding level is evaluated by analytical calculation and CFX analysis for an assumed condition. The flood level is calculated as 0.71m and 1.1m analytically and with CFX simulation, respectively. Comparing the analytical calculation and simulation, they are similar, but the analytical calculation is not conservative. There are many factors reducing the drainage capacity such as air-core vortex, intake of air, and turbulent flow. Therefore, in case of flood level evaluation by analytical calculation, a sufficient safety margin should be considered

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-10-01

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

  13. Development and application of MASKA-LM code for calculation of thermal hydraulics and mass transfer of lead cooled fast reactors

    International Nuclear Information System (INIS)

    Vladimir Ya Kumaev; Andrei A Lebezov; Victor V Alexeev

    2005-01-01

    Full text of publication follows: The report is devoted to the development and application of the two-dimensional MASKA-LM computer code intended for numerical calculations of lead coolant flows, temperatures and transport of impurities in BREST-type reactors of the integral design. The description of heat and mass transfer in liquid metal systems, proceeding in the coolant and at the interface 'coolant - structural materials', is a complex problem involving the joint simulation of thermal-hydraulic, physical and chemical processes in view of the real configuration of the reactor circuit. The report presents the state-of-the-art in the development of the two-dimensional code MASKA-LM and the results of trial calculations of heat and mass transfer in the primary circuit of the lead cooled reactor. The set of governing equations to be solved is based on the porous body model and describes the thermal-hydraulic processes in the reactor as a whole. The numerical method for solution of the governing equations is discussed. To check the code workability and study the technique by the way of solution of a particular task, calculations were performed in reference to the chosen version of the lead cooled BREST reactor under design. The examined domain of the reactor was simulated by a porous body with the parameters corresponding to those of the real reactor medium in terms of heat generation, resistance and the geometry of the hydraulic path of coolant. Analysis of the calculated two-dimensional fields of velocities, pressure and temperatures shows the existence of a complex coolant flow with stagnant and vortex zones. A nonuniform distribution of the coolant flow rate along the core radius was obtained. The results of calculations of the impurity transport of iron, oxygen and magnetite in the primary reactor circuit are discussed as well. The developed code MASKA-LM allows one to evaluate the issue of components of structural materials into coolant as impurities, their

  14. A Monte Carlo neutron transport code for eigenvalue calculations on a dual-GPU system and CUDA environment

    Energy Technology Data Exchange (ETDEWEB)

    Liu, T.; Ding, A.; Ji, W.; Xu, X. G. [Nuclear Engineering and Engineering Physics, Rensselaer Polytechnic Inst., Troy, NY 12180 (United States); Carothers, C. D. [Dept. of Computer Science, Rensselaer Polytechnic Inst. RPI (United States); Brown, F. B. [Los Alamos National Laboratory (LANL) (United States)

    2012-07-01

    Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis. Its lengthy computation time can be reduced by general-purpose computing on Graphics Processing Units (GPU), one of the latest parallel computing techniques under development. The method of porting a regular transport code to GPU is usually very straightforward due to the 'embarrassingly parallel' nature of MC code. However, the situation becomes different for eigenvalue calculation in that it will be performed on a generation-by-generation basis and the thread coordination should be explicitly taken care of. This paper presents our effort to develop such a GPU-based MC code in Compute Unified Device Architecture (CUDA) environment. The code is able to perform eigenvalue calculation under simple geometries on a multi-GPU system. The specifics of algorithm design, including thread organization and memory management were described in detail. The original CPU version of the code was tested on an Intel Xeon X5660 2.8 GHz CPU, and the adapted GPU version was tested on NVIDIA Tesla M2090 GPUs. Double-precision floating point format was used throughout the calculation. The result showed that a speedup of 7.0 and 33.3 were obtained for a bare spherical core and a binary slab system respectively. The speedup factor was further increased by a factor of {approx}2 on a dual GPU system. The upper limit of device-level parallelism was analyzed, and a possible method to enhance the thread-level parallelism was proposed. (authors)

  15. A Monte Carlo neutron transport code for eigenvalue calculations on a dual-GPU system and CUDA environment

    International Nuclear Information System (INIS)

    Liu, T.; Ding, A.; Ji, W.; Xu, X. G.; Carothers, C. D.; Brown, F. B.

    2012-01-01

    Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis. Its lengthy computation time can be reduced by general-purpose computing on Graphics Processing Units (GPU), one of the latest parallel computing techniques under development. The method of porting a regular transport code to GPU is usually very straightforward due to the 'embarrassingly parallel' nature of MC code. However, the situation becomes different for eigenvalue calculation in that it will be performed on a generation-by-generation basis and the thread coordination should be explicitly taken care of. This paper presents our effort to develop such a GPU-based MC code in Compute Unified Device Architecture (CUDA) environment. The code is able to perform eigenvalue calculation under simple geometries on a multi-GPU system. The specifics of algorithm design, including thread organization and memory management were described in detail. The original CPU version of the code was tested on an Intel Xeon X5660 2.8 GHz CPU, and the adapted GPU version was tested on NVIDIA Tesla M2090 GPUs. Double-precision floating point format was used throughout the calculation. The result showed that a speedup of 7.0 and 33.3 were obtained for a bare spherical core and a binary slab system respectively. The speedup factor was further increased by a factor of ∼2 on a dual GPU system. The upper limit of device-level parallelism was analyzed, and a possible method to enhance the thread-level parallelism was proposed. (authors)

  16. Development and preliminary validation of flux map processing code MAPLE

    International Nuclear Information System (INIS)

    Li Wenhuai; Zhang Xiangju; Dang Zhen; Chen Ming'an; Lu Haoliang; Li Jinggang; Wu Yuanbao

    2013-01-01

    The self-reliant flux map processing code MAPLE was developed by China General Nuclear Power Corporation (CGN). Weight coefficient method (WCM), polynomial expand method (PEM) and thin plane spline (TPS) method were applied to fit the deviation between measured and predicted detector signal results for two-dimensional radial plane, to interpolate or extrapolate the non-instrumented location deviation. Comparison of results in the test cases shows that the TPS method can better capture the information of curved fitting lines than the other methods. The measured flux map data of the Lingao Nuclear Power Plant were processed using MAPLE as validation test cases, combined with SMART code. Validation results show that the calculation results of MAPLE are reasonable and satisfied. (authors)

  17. Computer code for calculating personnel doses due to tritium exposures

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  18. The Coding Process and Its Challenges

    Directory of Open Access Journals (Sweden)

    Judith A. Holton, Ph.D.

    2010-02-01

    Full Text Available Coding is the core process in classic grounded theory methodology. It is through coding that the conceptual abstraction of data and its reintegration as theory takes place. There are two types of coding in a classic grounded theory study: substantive coding, which includes both open and selective coding procedures, and theoretical coding. In substantive coding, the researcher works with the data directly, fracturing and analysing it, initially through open coding for the emergence of a core category and related concepts and then subsequently through theoretical sampling and selective coding of data to theoretically saturate the core and related concepts. Theoretical saturation is achieved through constant comparison of incidents (indicators in the data to elicit the properties and dimensions of each category (code. This constant comparing of incidents continues until the process yields the interchangeability of indicators, meaning that no new properties or dimensions are emerging from continued coding and comparison. At this point, the concepts have achieved theoretical saturation and the theorist shifts attention to exploring the emergent fit of potential theoretical codes that enable the conceptual integration of the core and related concepts to produce hypotheses that account for relationships between the concepts thereby explaining the latent pattern of social behaviour that forms the basis of the emergent theory. The coding of data in grounded theory occurs in conjunction with analysis through a process of conceptual memoing, capturing the theorist’s ideation of the emerging theory. Memoing occurs initially at the substantive coding level and proceeds to higher levels of conceptual abstraction as coding proceeds to theoretical saturation and the theorist begins to explore conceptual reintegration through theoretical coding.

  19. An IBM-1620 code for calculation of isotopic composition of irradiated thorium (ISOCOM-2)

    International Nuclear Information System (INIS)

    Soliman, R.H.; Karchava, G.; Hamouda, I.

    1978-01-01

    The present work gives a description of an IBM-1620 code to calculate the isotopic composition during the irradiation of a nuclear fuel, which initially contains 232 Th. The numerical results on test calculations are presented. The code has been in operation since 1968

  20. Calculation of the void reactivity of CANDU lattices using the SCALE code system

    Energy Technology Data Exchange (ETDEWEB)

    Valko, J. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Feher, S. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Slobben, J. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1995-11-01

    The reactivity effect of coolant voiding in CANDU-type fuel lattices has been calculated with different methods using the SCALE code system. The known positive void reactivity coefficient of the original lattice was correctly obtained. A modified fuel bundle containing dysprosium and slightly enriched uranium to eliminate the positive reactivity effect was also calculated. Owing to the increased heterogeneity of this modified fuel the one-dimensional cylindrical calculation with XSDRN proved to be inadequate. Code options allowing bundle geometry were successfully used for the calculation of the strongly space dependent flux and spectrum changes which determine the void reactivity. (orig.).

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

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.; Panayotov, D.; Ilieva, B.

    2001-08-01

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

  2. User manual of FUNF code for fissile material data calculation

    International Nuclear Information System (INIS)

    Zhang, Jingshang

    2006-03-01

    The FUNF code (2005 version) is used to calculate fast neutron reaction data of fissile materials with incident energies from about 1 keV up to 20 MeV. The first version of the FUNF code was completed in 1994. the code has been developed continually since that time and has often been used as an evaluation tool for setting up CENDL and for analyzing the measurements of fissile materials. During these years many improvements have been made. In this manual, the format of the input parameter files and the output files, as well as the functions of flag used in FUNF code, are introduced in detail, and the examples of the format of input parameters files are given. FUNF code consists of the spherical optical model, the Hauser-Feshbach model, and the unified Hauser-Feshbach and exciton model. (authors)

  3. A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)

    1997-12-31

    Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.

  4. A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

    International Nuclear Information System (INIS)

    Marguet, S.D.

    1997-01-01

    Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF's neutronic code COCCINELLE uses the Rowland's formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission's products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on 'low' configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.)

  5. Contributions to the validation of advanced codes for accident analysis calculations with 3-dimensional neutron kinetics. STC with the Ukraine. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Rohde, U.; Khalimonchuk, V.; Kuchin, A.; Seidel, A.

    2000-10-01

    In the frame of a project of scientific-technical cooperation funded by BMBF/BMWi, the coupled code ATHLET-DYN3D has been transferred to the Scientific and Technical Centre on Nuclear and Radiation Safety Kiev (Ukraine). This program code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermohydraulics code system ATHLET. For the purpose of validation of this coupled code, a measurement data base has been generated. In the data base suitable experimental data for operational transients from NPPs are collected. The data collection and documentation was performed in accordance with a directive about requirements to measurement data for code validation, which has been elaborated within the project. The validation calculations have been performed for two selected transients. The results of these calculations were compared with measurement values from the data base. The function of the code DYN3D was expanded with a subroutine for reactivity coefficients calculation. Using this modification of the code DYN3D, investigations of reactivity contributions on different operational processes can be performed. (orig.) [de

  6. Research on pre-processing of QR Code

    Science.gov (United States)

    Sun, Haixing; Xia, Haojie; Dong, Ning

    2013-10-01

    QR code encodes many kinds of information because of its advantages: large storage capacity, high reliability, full arrange of utter-high-speed reading, small printing size and high-efficient representation of Chinese characters, etc. In order to obtain the clearer binarization image from complex background, and improve the recognition rate of QR code, this paper researches on pre-processing methods of QR code (Quick Response Code), and shows algorithms and results of image pre-processing for QR code recognition. Improve the conventional method by changing the Souvola's adaptive text recognition method. Additionally, introduce the QR code Extraction which adapts to different image size, flexible image correction approach, and improve the efficiency and accuracy of QR code image processing.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

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

  8. Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)

    2014-07-01

    A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm{sup 2}s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)

  9. Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

    International Nuclear Information System (INIS)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

    2014-01-01

    A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm 2 s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)

  10. DYNREL - the reference calculation (coupled code utilization on analysis of RIA-transient)

    International Nuclear Information System (INIS)

    Strmensky, C.; Darilek, P.

    2003-01-01

    DYNREL is coupled code, comprising DYN3D and RELAP5 programs. The coupled code has been developed during four years. Now DYNREL is tested on selected RIA and thermo-hydraulic transient calculations. This material describes some results from selected RIA transient calculation (initiated by control rod movement). DYNREL modelled the whole nuclear reactors. The core is modeled as 313 or 349 independent thermo-hydraulic channels with 10 or 20 axial layers. Thermo-hydraulic part contains about 700 components that covered the six loops' model of nuclear power plant in detail. The calculated results are compared with DYN3D/M3, DYN3D/H1.1 results (Authors)

  11. Licensing in BE system code calculations. Applications and uncertainty evaluation by CIAU method

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco

    2007-01-01

    The evaluation of uncertainty constitutes the necessary supplement of Best Estimate (BE) calculations performed to understand accident scenarios in water cooled nuclear reactors. The needs come from the imperfection of computational tools on the one side and from the interest in using such tool to get more precise evaluation of safety margins. In the present paper the approaches to uncertainty are outlined and the CIAU (Code with capability of Internal Assessment of Uncertainty) method proposed by the University of Pisa is described including ideas at the basis and results from applications. Two approaches are distinguished that are characterized as 'propagation of code input uncertainty' and 'propagation of code output errors'. For both methods, the thermal-hydraulic code is at the centre of the process of uncertainty evaluation: in the former case the code itself is adopted to compute the error bands and to propagate the input errors, in the latter case the errors in code application to relevant measurements are used to derive the error bands. The CIAU method exploits the idea of the 'status approach' for identifying the thermal-hydraulic conditions of an accident in any Nuclear Power Plant (NPP). Errors in predicting such status are derived from the comparison between predicted and measured quantities and, in the stage of the application of the method, are used to compute the uncertainty. (author)

  12. Approach to the calculation of energy deposition in a container of fuel irradiated by the neutronic codes coupling fluid-dynamics

    International Nuclear Information System (INIS)

    Hueso, C.; Aleman, A.; Colomer, C.; Fabbri, M.; Martin, M.; Saellas, J.

    2013-01-01

    In this work identifies a possible area of improvement through the creation of a code of coupling between deposition energy codes which calculate neutron (MCNP), and data from heading into fluid dynamics (ANSYS-Fluent) or codes thermomechanical, called MAFACS (Monte Carlo ANSYS Fluent Automatic Coupling Software), being possible to so summarize the process by shortening the needs of computing time, increasing the precision of the results and therefore improving the design of the components.

  13. A comparison of FEMAXI-III code calculations with irradiation experiments

    International Nuclear Information System (INIS)

    Ito, K.; Sogame, M.; Ichikawa, M.; Nakajima, T.

    1981-01-01

    The FEMAXI-III code calculations were compared with in-pile diameter measurements in the Halden Boiling Water Reactor, in order to check the ability to analyse the pellet-cladding mechanical interaction. The results showed generally good agreement between calculations and measurements. The Studsvik INTER-RAMP Experiments were also analysed to examine the predictability of fuel rod failures. Good agreement was obtained between calculated and measured fission gas x release. The threshold stress to cause failure was estimated by means of FEMAXI-III. (author)

  14. Calculations of Sobol indices for the Gaussian process metamodel

    Energy Technology Data Exchange (ETDEWEB)

    Marrel, Amandine [CEA, DEN, DTN/SMTM/LMTE, F-13108 Saint Paul lez Durance (France)], E-mail: amandine.marrel@cea.fr; Iooss, Bertrand [CEA, DEN, DER/SESI/LCFR, F-13108 Saint Paul lez Durance (France); Laurent, Beatrice [Institut de Mathematiques, Universite de Toulouse (UMR 5219) (France); Roustant, Olivier [Ecole des Mines de Saint-Etienne (France)

    2009-03-15

    Global sensitivity analysis of complex numerical models can be performed by calculating variance-based importance measures of the input variables, such as the Sobol indices. However, these techniques, requiring a large number of model evaluations, are often unacceptable for time expensive computer codes. A well-known and widely used decision consists in replacing the computer code by a metamodel, predicting the model responses with a negligible computation time and rending straightforward the estimation of Sobol indices. In this paper, we discuss about the Gaussian process model which gives analytical expressions of Sobol indices. Two approaches are studied to compute the Sobol indices: the first based on the predictor of the Gaussian process model and the second based on the global stochastic process model. Comparisons between the two estimates, made on analytical examples, show the superiority of the second approach in terms of convergence and robustness. Moreover, the second approach allows to integrate the modeling error of the Gaussian process model by directly giving some confidence intervals on the Sobol indices. These techniques are finally applied to a real case of hydrogeological modeling.

  15. Calculations of Sobol indices for the Gaussian process metamodel

    International Nuclear Information System (INIS)

    Marrel, Amandine; Iooss, Bertrand; Laurent, Beatrice; Roustant, Olivier

    2009-01-01

    Global sensitivity analysis of complex numerical models can be performed by calculating variance-based importance measures of the input variables, such as the Sobol indices. However, these techniques, requiring a large number of model evaluations, are often unacceptable for time expensive computer codes. A well-known and widely used decision consists in replacing the computer code by a metamodel, predicting the model responses with a negligible computation time and rending straightforward the estimation of Sobol indices. In this paper, we discuss about the Gaussian process model which gives analytical expressions of Sobol indices. Two approaches are studied to compute the Sobol indices: the first based on the predictor of the Gaussian process model and the second based on the global stochastic process model. Comparisons between the two estimates, made on analytical examples, show the superiority of the second approach in terms of convergence and robustness. Moreover, the second approach allows to integrate the modeling error of the Gaussian process model by directly giving some confidence intervals on the Sobol indices. These techniques are finally applied to a real case of hydrogeological modeling

  16. Benchmark calculation for GT-MHR using HELIOS/MASTER code package and MCNP

    International Nuclear Information System (INIS)

    Lee, Kyung Hoon; Kim, Kang Seog; Noh, Jae Man; Song, Jae Seung; Zee, Sung Quun

    2005-01-01

    The latest research associated with the very high temperature gas-cooled reactor (VHTR) is focused on the verification of a system performance and safety under operating conditions for the VHTRs. As a part of those, an international gas-cooled reactor program initiated by IAEA is going on. The key objectives of this program are the validation of analytical computer codes and the evaluation of benchmark models for the projected and actual VHTRs. New reactor physics analysis procedure for the prismatic VHTR is under development by adopting the conventional two-step procedure. In this procedure, a few group constants are generated through the transport lattice calculations using the HELIOS code, and the core physics analysis is performed by the 3-dimensional nodal diffusion code MASTER. We evaluated the performance of the HELIOS/MASTER code package through the benchmark calculations related to the GT-MHR (Gas Turbine-Modular Helium Reactor) to dispose weapon plutonium. In parallel, MCNP is employed as a reference code to verify the results of the HELIOS/MASTER procedure

  17. Initial Assessment of Parallelization of Monte Carlo Calculation using Graphics Processing Units

    International Nuclear Information System (INIS)

    Choi, Sung Hoon; Joo, Han Gyu

    2009-01-01

    Monte Carlo (MC) simulation is an effective tool for calculating neutron transports in complex geometry. However, because Monte Carlo simulates each neutron behavior one by one, it takes a very long computing time if enough neutrons are used for high precision of calculation. Accordingly, methods that reduce the computing time are required. In a Monte Carlo code, parallel calculation is well-suited since it simulates the behavior of each neutron independently and thus parallel computation is natural. The parallelization of the Monte Carlo codes, however, was done using multi CPUs. By the global demand for high quality 3D graphics, the Graphics Processing Unit (GPU) has developed into a highly parallel, multi-core processor. This parallel processing capability of GPUs can be available to engineering computing once a suitable interface is provided. Recently, NVIDIA introduced CUDATM, a general purpose parallel computing architecture. CUDA is a software environment that allows developers to manage GPU using C/C++ or other languages. In this work, a GPU-based Monte Carlo is developed and the initial assessment of it parallel performance is investigated

  18. Load balancing in highly parallel processing of Monte Carlo code for particle transport

    International Nuclear Information System (INIS)

    Higuchi, Kenji; Takemiya, Hiroshi; Kawasaki, Takuji

    1998-01-01

    In parallel processing of Monte Carlo (MC) codes for neutron, photon and electron transport problems, particle histories are assigned to processors making use of independency of the calculation for each particle. Although we can easily parallelize main part of a MC code by this method, it is necessary and practically difficult to optimize the code concerning load balancing in order to attain high speedup ratio in highly parallel processing. In fact, the speedup ratio in the case of 128 processors remains in nearly one hundred times when using the test bed for the performance evaluation. Through the parallel processing of the MCNP code, which is widely used in the nuclear field, it is shown that it is difficult to attain high performance by static load balancing in especially neutron transport problems, and a load balancing method, which dynamically changes the number of assigned particles minimizing the sum of the computational and communication costs, overcomes the difficulty, resulting in nearly fifteen percentage of reduction for execution time. (author)

  19. A computer code for calculation of radioactive nuclide generation and depletion, decay heat and γ ray spectrum. FPGS90

    International Nuclear Information System (INIS)

    Ihara, Hitoshi; Katakura, Jun-ichi; Nakagawa, Tsuneo

    1995-11-01

    In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting γ ray and β ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted γ ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library 'JNDC Nuclear Data Library of Fission Products - second version -', which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author)

  20. A computer code for calculation of radioactive nuclide generation and depletion, decay heat and {gamma} ray spectrum. FPGS90

    Energy Technology Data Exchange (ETDEWEB)

    Ihara, Hitoshi; Katakura, Jun-ichi; Nakagawa, Tsuneo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1995-11-01

    In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting {gamma} ray and {beta} ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted {gamma} ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library `JNDC Nuclear Data Library of Fission Products - second version -`, which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author).

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

    International Nuclear Information System (INIS)

    Mann, F.M.

    1979-07-01

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

  2. Calculation of fluid-structure interaction for reactor safety with the Cassiopee code

    International Nuclear Information System (INIS)

    Graveleau, J.L.; Louvet, P.D.

    1979-01-01

    The cassiopee code is an eulerian-lagrangian coupled code for computations where the hydrodynamic is coupled with structural domains. It is completely explicit. The fluid zones may be computed either in lagrangian or in eulerian coordinates; thin shells can be computed wih their flexural behaviour; elastic plastic zones must be calculated in a lagrangian way. This code is under development in Cadarache. Its purpose is to compute the hypothetical core disruptive accident of a LMFBR when lagrangian codes are not sufficient. This paper contains a description of the code and two examples of computations, one of which has been compared with experimental results

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

    International Nuclear Information System (INIS)

    Laborde, J.C.; Prevost, C.; Vendel, J.

    1995-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  7. Stability analysis by ERATO code

    International Nuclear Information System (INIS)

    Tsunematsu, Toshihide; Takeda, Tatsuoki; Matsuura, Toshihiko; Azumi, Masafumi; Kurita, Gen-ichi

    1979-12-01

    Problems in MHD stability calculations by ERATO code are described; which concern convergence property of results, equilibrium codes, and machine optimization of ERATO code. It is concluded that irregularity on a convergence curve is not due to a fault of the ERATO code itself but due to inappropriate choice of the equilibrium calculation meshes. Also described are a code to calculate an equilibrium as a quasi-inverse problem and a code to calculate an equilibrium as a result of a transport process. Optimization of the code with respect to I/O operations reduced both CPU time and I/O time considerably. With the FACOM230-75 APU/CPU multiprocessor system, the performance is about 6 times as high as with the FACOM230-75 CPU, showing the effectiveness of a vector processing computer for the kind of MHD computations. This report is a summary of the material presented at the ERATO workshop 1979(ORNL), supplemented with some details. (author)

  8. A Source Term Calculation for the APR1400 NSSS Auxiliary System Components Using the Modified SHIELD Code

    International Nuclear Information System (INIS)

    Park, Hong Sik; Kim, Min; Park, Seong Chan; Seo, Jong Tae; Kim, Eun Kee

    2005-01-01

    The SHIELD code has been used to calculate the source terms of NSSS Auxiliary System (comprising CVCS, SIS, and SCS) components of the OPR1000. Because the code had been developed based upon the SYSTEM80 design and the APR1400 NSSS Auxiliary System design is considerably changed from that of SYSTEM80 or OPR1000, the SHIELD code cannot be used directly for APR1400 radiation design. Thus the hand-calculation is needed for the portion of design changes using the results of the SHIELD code calculation. In this study, the SHIELD code is modified to incorporate the APR1400 design changes and the source term calculation is performed for the APR1400 NSSS Auxiliary System components

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  10. Code-B-1 for stress/strain calculation for TRISO fuel particle (Contract research)

    International Nuclear Information System (INIS)

    Aihara, Jun; Ueta, Shohei; Shibata, Taiju; Sawa, Kazuhiro

    2011-12-01

    We have developed Code-B-1 for the prediction of the failure probabilities of the coated fuel particles for the high temperature gas-cooled reactors (HTGRs) under operation by modification of an existing code. A finite element method (FEM) is employed for the stress calculation part and Code-B-1 can treat the plastic deformation of the coating layer of the coated fuel particles which the existing code cannot treat. (author)

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

    International Nuclear Information System (INIS)

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

    1986-11-01

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

  12. Calculation of criticality of the AP600 reactor with KENO V.a code

    Energy Technology Data Exchange (ETDEWEB)

    Krumbein, A; Caner, M; Shapira, M [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center

    1996-12-01

    The Westinghouse AP600 PWR has been modeled using the KENO V.a three dimensional Monte Carlo criticality program of the SCALE-PC code system. These calculations and the use of a Monte Carlo neutron transport code such as KENO will provide us with an independent check on our WIMS/CITATION calculations for the AP600 as well as for other reactors. It will also enable us to model more complicated geometries. (authors).

  13. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    International Nuclear Information System (INIS)

    Wan, C.; Cao, L.; Wu, H.; Zu, T.; Shen, W.

    2015-01-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  14. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Wan, C.; Cao, L.; Wu, H.; Zu, T., E-mail: chenghuiwan@stu.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: tiejun@mail.xjtu.edu.cn [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Shen, W., E-mail: Wei.Shen@cnsc-ccsn.gc.ca [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-07-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  15. Implementation of an implicit method into heat conduction calculation of TRAC-PF1/MOD2 code

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Abe, Yutaka; Ohnuki, Akira; Murao, Yoshio

    1990-08-01

    A two-dimensional unsteady heat conduction equation is solved in the TRAC-PF/MOD2 code to calculate temperature transients in fuel rod. A large CPU time is often required to get stable solution of temperature transients in the TRAC calculation with a small axial node size (less than 1.0 mm), because the heat conduction equation is discretized explicitly. To eliminate the restriction of the maximum time step size by the heat conduction calculation, an implicit method for solving the heat condition equation was developed and implemented into the TRAC code. Several assessment calculations were performed with the original and modified TRAC codes. It is confirmed that the implicit method is reliable and is successfully implemented into the TRAC code through comparison with theoretical solutions and assessment calculation results. It is demonstrated that the implicit method makes the heat conduction calculation practical even for the analyses of temperature transients with the axial node size less than 0.1 mm. (author)

  16. LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory

    International Nuclear Information System (INIS)

    Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.

    1976-04-01

    The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented

  17. Simulation codes of chemical separation process of spent fuel reprocessing. Tool for process development and safety research

    International Nuclear Information System (INIS)

    Asakura, Toshihide; Sato, Makoto; Matsumura, Masakazu; Morita, Yasuji

    2005-01-01

    This paper reviews the succeeding development and utilization of Extraction System Simulation Code for Advanced Reprocessing (ESSCAR). From the viewpoint of development, more tests with spent fuel and calculations should be performed with better understanding of the physico-chemical phenomena in a separation process. From the viewpoint of process safety research on fuel cycle facilities, it is important to know the process behavior of a key substance; being highly reactive but existing only trace amount. (author)

  18. Code-Mixing and Code Switchingin The Process of Learning

    Directory of Open Access Journals (Sweden)

    Diyah Atiek Mustikawati

    2016-09-01

    Full Text Available This study aimed to describe a form of code switching and code mixing specific form found in the teaching and learning activities in the classroom as well as determining factors influencing events stand out that form of code switching and code mixing in question.Form of this research is descriptive qualitative case study which took place in Al Mawaddah Boarding School Ponorogo. Based on the analysis and discussion that has been stated in the previous chapter that the form of code mixing and code switching learning activities in Al Mawaddah Boarding School is in between the use of either language Java language, Arabic, English and Indonesian, on the use of insertion of words, phrases, idioms, use of nouns, adjectives, clauses, and sentences. Code mixing deciding factor in the learning process include: Identification of the role, the desire to explain and interpret, sourced from the original language and its variations, is sourced from a foreign language. While deciding factor in the learning process of code, includes: speakers (O1, partners speakers (O2, the presence of a third person (O3, the topic of conversation, evoke a sense of humour, and just prestige. The significance of this study is to allow readers to see the use of language in a multilingual society, especially in AL Mawaddah boarding school about the rules and characteristics variation in the language of teaching and learning activities in the classroom. Furthermore, the results of this research will provide input to the ustadz / ustadzah and students in developing oral communication skills and the effectiveness of teaching and learning strategies in boarding schools.

  19. Parallel processing of structural integrity analysis codes

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.

    1996-01-01

    Structural integrity analysis forms an important role in assessing and demonstrating the safety of nuclear reactor components. This analysis is performed using analytical tools such as Finite Element Method (FEM) with the help of digital computers. The complexity of the problems involved in nuclear engineering demands high speed computation facilities to obtain solutions in reasonable amount of time. Parallel processing systems such as ANUPAM provide an efficient platform for realising the high speed computation. The development and implementation of software on parallel processing systems is an interesting and challenging task. The data and algorithm structure of the codes plays an important role in exploiting the parallel processing system capabilities. Structural analysis codes based on FEM can be divided into two categories with respect to their implementation on parallel processing systems. The first category codes such as those used for harmonic analysis, mechanistic fuel performance codes need not require the parallelisation of individual modules of the codes. The second category of codes such as conventional FEM codes require parallelisation of individual modules. In this category, parallelisation of equation solution module poses major difficulties. Different solution schemes such as domain decomposition method (DDM), parallel active column solver and substructuring method are currently used on parallel processing systems. Two codes, FAIR and TABS belonging to each of these categories have been implemented on ANUPAM. The implementation details of these codes and the performance of different equation solvers are highlighted. (author). 5 refs., 12 figs., 1 tab

  20. New developments on Monte Carlo simulation code for the calculation of Atom Displacements Induced rates by High Energy Electrons in Solid Materials

    International Nuclear Information System (INIS)

    Damiani, Daniela D.; Cruz, Carlos M.; Pinnera, Ibrahin; Abreu, Yamiel; Leyva, Antonio

    2015-01-01

    New developments and simulations on regard to the interactions of incident gamma radiation over solids materials using the MCSAD (Monte Carlo Simulation of Atom Displacement) code are presented. In this code Monte Carlo algorithms are applied in order to sample all electrons and gamma interaction processes occurring during their transport through a solid target, especially those connected to the output of atom displacements events. Particularly, it is calculated the limit angle to elastic scattering for the electrons on a new approach, which allows correctly the splitting of the electron single processes at higher scattering angles. On this way, the probability of single electron scattering processes transferring high recoil atomic energy leading to atom displacement effects is calculated and consequently sampled in the MCSAD code. In addition, it is considered some other new theoretical aspects in order to improve previous versions, like the one concerning the selection of threshold energy for displacements at a given atom site in dependence of the atom recoil direction. (Author)

  1. Calculation code of heterogeneity effects for analysis of small sample reactivity worth

    International Nuclear Information System (INIS)

    Okajima, Shigeaki; Mukaiyama, Takehiko; Maeda, Akio.

    1988-03-01

    The discrepancy between experimental and calculated central reactivity worths has been one of the most significant interests for the analysis of fast reactor critical experiment. Two effects have been pointed out so as to be taken into account in the calculation as the possible cause of the discrepancy; one is the local heterogeneity effect which is associated with the measurement geometry, the other is the heterogeneity effect on the distribution of the intracell adjoint flux. In order to evaluate these effects in the analysis of FCA actinide sample reactivity worth the calculation code based on the collision probability method was developed. The code can handle the sample size effect which is one of the local heterogeneity effects and also the intracell adjoint heterogeneity effect. (author)

  2. PCRELAP5: data calculation program for RELAP 5 code

    International Nuclear Information System (INIS)

    Silvestre, Larissa Jacome Barros

    2016-01-01

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

  3. Development of a nuclear spallation simulation code and calculations of primary spallation products

    International Nuclear Information System (INIS)

    Nishida, Takahiko; Nakahara, Yasuaki; Tsutsui, Tsuneo

    1986-08-01

    In order to make evaluations of computational models for the nuclear spallation reaction from a nuclear physics point of view, a simulation code NUCLEUS has been developed by modifying and combining the Monte Carlo codes NMTC/JAERI and NMTA/JAERI for calculating only the nuclear spallation reaction (intranuclear cascade + evaporation and/or fast fission) between a nucleus and a projectile without taking into consideration of internuclear transport. New several plotting routines have been provided for the rapid process of much more event data, obtained by using the ARGUS plotting system. The results obtained by our code can be directly compared with the experimental results using by thin foil experiments in which internuclear multiple collisions have little effects, and will serve to upgrade the calculational methods and the values of nuclear parameters currently used in the calculations. Some discussions are done about the preliminary computational results obtained by using NUCLEUS. The mass distribution and charge dispersion of reaction products are examined in some detail for the nuclear spallation reaction between incident protons and target nuclei, such as U, Pb and Ag, in the energy range from 0.5 GeV to 3.0 GeV. These results show that the distribution of reaction products ceases to change its form as the proton energy increases over about 2 GeV. The same tendency is seen in the energy dependence of the number of primary particles emitted from a nucleus. After spallation reactions, a variety of nuclei, especially many neutron deficient nuclides with nuclear charges nearly equal to ones of a target nucleus, are produced. Due to their short lifetime most of them will change to stable nuclides in due time. Finally, some important issues are discussed to improve the present simulation method. (author)

  4. Uncertainties in calculations of nuclear design code system for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Shindo, R.; Yamashita, K.; Murata, I.

    1991-01-01

    The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs

  5. Comparison of computer code calculations with experimental results obtained in the NSPP series of experiments

    International Nuclear Information System (INIS)

    Tobias, M.L.

    1987-01-01

    Experiments were done on several aerosols in air atmospheres at varying temperatures and humidity conditions of interest in forming a data base for testing aerosol behavior models used as part of the process of evaluating the ''source term'' in light water reactor accidents. This paper deals with the problems of predicting the observed experimental data for suspended aerosol concentration with aerosol calculational codes. Comparisons of measured versus predicted data are provided

  6. EFFDOS - a FORTRAN-77-code for the calculation of the effective dose equivalent

    International Nuclear Information System (INIS)

    Baer, M.; Honcu, S.; Huebschmann, W.

    1984-01-01

    The FORTRAN-77-code EFFDOS calculates the effective dose equivalent according to ICRP 26 due to the longterm emission of radionuclides into the atmosphere for the following exposure pathways: inhalation, ingestion, γ-ground irradiation (γ-irradiation by radionuclides deposited on the ground) and β- or γ-submersion (irradiation by the passing radioactive cloud). For calculating the effective dose equivalent at a single spot it is necessary to put in the diffusion factor and - if need be - the washout factor; otherwise EFFDOS calculates the input data for the computer codes ISOLA III and WOLGA-1, which then are enabled to compute the atmospheric diffusion, ground deposition and local dose equivalent distribution for the requested exposure pathway. Atmospheric diffusion, deposition and radionuclide transfer are calculated according to the ''Allgemeine Berechnungsgrundlage ....'' recommended by the German Fed. Ministry of Interior. A sample calculated is added. (orig.) [de

  7. Aerosol sampling and Transport Efficiency Calculation (ASTEC) and application to surtsey/DCH aerosol sampling system: Code version 1.0: Code description and user's manual

    International Nuclear Information System (INIS)

    Yamano, N.; Brockmann, J.E.

    1989-05-01

    This report describes the features and use of the Aerosol Sampling and Transport Efficiency Calculation (ASTEC) Code. The ASTEC code has been developed to assess aerosol transport efficiency source term experiments at Sandia National Laboratories. This code also has broad application for aerosol sampling and transport efficiency calculations in general as well as for aerosol transport considerations in nuclear reactor safety issues. 32 refs., 31 figs., 7 tabs

  8. Opacity calculations for extreme physical systems: code RACHEL

    Science.gov (United States)

    Drska, Ladislav; Sinor, Milan

    1996-08-01

    Computer simulations of physical systems under extreme conditions (high density, temperature, etc.) require the availability of extensive sets of atomic data. This paper presents basic information on a self-consistent approach to calculations of radiative opacity, one of the key characteristics of such systems. After a short explanation of general concepts of the atomic physics of extreme systems, the structure of the opacity code RACHEL is discussed and some of its applications are presented.

  9. MVP/GMVP version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

    2017-03-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

  10. Application of MCNP code in shielding calculation of minitype fast reactor

    International Nuclear Information System (INIS)

    He Keyu; Han Weishi

    2008-01-01

    An accurate shielding calculation model has been set up for the minitype sodium-cooled fast reactor (MFR) based on MCNP code and particular calculation of its primary shielding parameters has been carried out. The results indicate that the photon and neutron flux density of MFR has rapidly fallen to a low-level. The material for the shielding layer outside of main container is primarily of carbon steel, which can be design as a shielding structure satisfying the safety code. The sodium activation in primary circuit is extremely limited and it is simple to shield from. Both the output of helium in reflector and burn up of boron-10 in control rod are very small. These materials can be used for several cycle lives. (authors)

  11. PUFF-III: A Code for Processing ENDF Uncertainty Data Into Multigroup Covariance Matrices

    International Nuclear Information System (INIS)

    Dunn, M.E.

    2000-01-01

    PUFF-III is an extension of the previous PUFF-II code that was developed in the 1970s and early 1980s. The PUFF codes process the Evaluated Nuclear Data File (ENDF) covariance data and generate multigroup covariance matrices on a user-specified energy grid structure. Unlike its predecessor, PUFF-III can process the new ENDF/B-VI data formats. In particular, PUFF-III has the capability to process the spontaneous fission covariances for fission neutron multiplicity. With regard to the covariance data in File 33 of the ENDF system, PUFF-III has the capability to process short-range variance formats, as well as the lumped reaction covariance data formats that were introduced in ENDF/B-V. In addition to the new ENDF formats, a new directory feature is now available that allows the user to obtain a detailed directory of the uncertainty information in the data files without visually inspecting the ENDF data. Following the correlation matrix calculation, PUFF-III also evaluates the eigenvalues of each correlation matrix and tests each matrix for positive definiteness. Additional new features are discussed in the manual. PUFF-III has been developed for implementation in the AMPX code system, and several modifications were incorporated to improve memory allocation tasks and input/output operations. Consequently, the resulting code has a structure that is similar to other modules in the AMPX code system. With the release of PUFF-III, a new and improved covariance processing code is available to process ENDF covariance formats through Version VI

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  13. Comparison of EGS4 and MCNP Monte Carlo codes when calculating radiotherapy depth doses.

    Science.gov (United States)

    Love, P A; Lewis, D G; Al-Affan, I A; Smith, C W

    1998-05-01

    The Monte Carlo codes EGS4 and MCNP have been compared when calculating radiotherapy depth doses in water. The aims of the work were to study (i) the differences between calculated depth doses in water for a range of monoenergetic photon energies and (ii) the relative efficiency of the two codes for different electron transport energy cut-offs. The depth doses from the two codes agree with each other within the statistical uncertainties of the calculations (1-2%). The relative depth doses also agree with data tabulated in the British Journal of Radiology Supplement 25. A discrepancy in the dose build-up region may by attributed to the different electron transport algorithims used by EGS4 and MCNP. This discrepancy is considerably reduced when the improved electron transport routines are used in the latest (4B) version of MCNP. Timing calculations show that EGS4 is at least 50% faster than MCNP for the geometries used in the simulations.

  14. Development and validation of a criticality calculation scheme based on French deterministic transport codes

    International Nuclear Information System (INIS)

    Santamarina, A.

    1991-01-01

    A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)

  15. Calculation of loading on pipes during filling processes

    International Nuclear Information System (INIS)

    Thiele, Thomas; Swidersky, Harald

    2013-01-01

    Filling processes in pipe systems do normally not belong to load design cases for which the integrity of pipelines and their mountings are verified with fluid- and structure-dynamic analysis. However, their frequency of occurrence is several times higher than those of the postulated incident-induced transients. That is why they have to be taken into consideration within fatigue analysis. The loading on pipes or rather on their mountings during filling processes originates from differences in the density of the transported fluids, e.g. at transport of gas slugs within water flow. The exposure time of the flow momentum force is fixed by the height of the flow velocity and by the length of discontinuities in the pipeline sections. Filling procedures frequently end with a pressure surge which was caused by the impingement and decelaration of the water plug at orifices in pipe systems. The calculation of such processes with 1D fluid-dynamic or rather thermal-hydraulic programs requires an idealization of the real form of the two phase flow or respectively of the two phase interface. In the past, several two phase flow regime maps were developed and implemented in codes for this. In this paper, the applicability of the thermo-hydraulic program RELAP5/MOD3.3 which is established in nuclear engineering is examined in order to calculate realistic loads from plug flows during the filling processes. For this, post-test calculations of experiments have been performed and the results have been compared with the experimental results as well as with the classical analytical approach according to Joukowsky. The comparison shows that, dependent on the discretization, the calculated loads are indeed partly underestimated, though the calculation results according to the Joukowsky-approach lie above the measurements. (orig.)

  16. SUNF, Simplified UNF Code, Fast Neutron Calculation by Unified Hauser-Feshbach Theory

    International Nuclear Information System (INIS)

    Zhang Jingshang

    2001-01-01

    1 - Description of program or function: The SUNF code is the simplified version of UNF code and is based on the unified Hauser-Feshbach and exciton model. SUNF code has been developed for calculations of fast neutron data for structural materials with neutron energies below 20 MeV. Besides elastic scattering channel, the code may handle decay sequence up to (n,3n) reaction, including 14 reaction channels. The energy spectra can be obtained and the output form is in the ENDF/B-6 format, but in file 5 form. For the ENDF-B-6 output, the incident energies are divided into two types: only cross section calculation; and those including neutron energy spectra. 2 - Methods: Gaussian integration is used for all numerical integration. 3 - Restrictions on the complexity of the problem: The incident energies of neutrons are from 1 KeV to 20 MeV. There are two parameters in this code: incident neutron energies number 'NEL'; and the number of discrete levels of residual nuclei for the first particle emissions 'NLV'. The users can set the values of NEL and NLV according to the storage size of the computer used. The number of discrete levels of residual nuclei for the multi-particle emissions is not greater than 20

  17. FORTRAN Code for Glandular Dose Calculation in Mammography Using Sobol-Wu Parameters

    Directory of Open Access Journals (Sweden)

    Mowlavi A A

    2007-07-01

    Full Text Available Background: Accurate computation of the radiation dose to the breast is essential to mammography. Various the thicknesses of breast, the composition of the breast tissue and other variables affect the optimal breast dose. Furthermore, the glandular fraction, which refers to the composition of the breasts, as partitioned between radiation-sensitive glandular tissue and the adipose tissue, also has an effect on this calculation. Fatty or fibrous breasts would have a lower value for the glandular fraction than dense breasts. Breast tissue composed of half glandular and half adipose tissue would have a glandular fraction in between that of fatty and dense breasts. Therefore, the use of a computational code for average glandular dose calculation in mammography is a more effective means of estimating the dose of radiation, and is accurate and fast. Methods: In the present work, the Sobol-Wu beam quality parameters are used to write a FORTRAN code for glandular dose calculation in molybdenum anode-molybdenum filter (Mo-Mo, molybdenum anode-rhodium filter (Mo-Rh and rhodium anode-rhodium filter (Rh-Rh target-filter combinations in mammograms. The input parameters of code are: tube voltage in kV, half-value layer (HVL of the incident x-ray spectrum in mm, breast thickness in cm (d, and glandular tissue fraction (g. Results: The average glandular dose (AGD variation against the voltage of the mammogram X-ray tube for d = 4 cm, HVL = 0.34 mm Al and g=0.5 for the three filter-target combinations, as well as its variation against the glandular fraction of breast tissue for kV=25, HVL=0.34, and d=4 cm has been calculated. The results related to the average glandular absorbed dose variation against HVL for kV = 28, d=4 cm and g= 0.6 are also presented. The results of this code are in good agreement with those previously reported in the literature. Conclusion: The code developed in this study calculates the glandular dose quickly, and it is complete and

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-05-01

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

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

    International Nuclear Information System (INIS)

    Daures, J.; Gouriou, J.; Bordy, J.M.

    2010-01-01

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

  20. Covariance data processing code. ERRORJ

    International Nuclear Information System (INIS)

    Kosako, Kazuaki

    2001-01-01

    The covariance data processing code, ERRORJ, was developed to process the covariance data of JENDL-3.2. ERRORJ has the processing functions of covariance data for cross sections including resonance parameters, angular distribution and energy distribution. (author)

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

    International Nuclear Information System (INIS)

    Rittmann, P.D.

    1990-12-01

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

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

    International Nuclear Information System (INIS)

    Kastner, W.; Erve, M.; Henzel, N.; Stellwag, B.

    1990-01-01

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

  3. FEAST: a two-dimensional non-linear finite element code for calculating stresses

    International Nuclear Information System (INIS)

    Tayal, M.

    1986-06-01

    The computer code FEAST calculates stresses, strains, and displacements. The code is two-dimensional. That is, either plane or axisymmetric calculations can be done. The code models elastic, plastic, creep, and thermal strains and stresses. Cracking can also be simulated. The finite element method is used to solve equations describing the following fundamental laws of mechanics: equilibrium; compatibility; constitutive relations; yield criterion; and flow rule. FEAST combines several unique features that permit large time-steps in even severely non-linear situations. The features include a special formulation for permitting many finite elements to simultaneously cross the boundary from elastic to plastic behaviour; accomodation of large drops in yield-strength due to changes in local temperature and a three-step predictor-corrector method for plastic analyses. These features reduce computing costs. Comparisons against twenty analytical solutions and against experimental measurements show that predictions of FEAST are generally accurate to ± 5%

  4. User's manual for a process model code

    International Nuclear Information System (INIS)

    Kern, E.A.; Martinez, D.P.

    1981-03-01

    The MODEL code has been developed for computer modeling of materials processing facilities associated with the nuclear fuel cycle. However, it can also be used in other modeling applications. This report provides sufficient information for a potential user to apply the code to specific process modeling problems. Several examples that demonstrate most of the capabilities of the code are provided

  5. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. The spread of the results in the international calculation amounted to ± 12,000 pcm in the realistic fuel dissolver exercise n degrees 19 proposed by BNFL, and to ± 25,000 pcm in the benchmark n degrees 20 in which fissile material in solid form is surrounded by fissile material in solution. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat latter effect, permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates solicited from the participants. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism (NITAWL in the international SCALE package) to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. Improvements in the up-dated 1990 contributions, as do recent complementary reference calculations (MCNP, VIM, ultrafine slowing-down CGM calculation), confirm the need to use rigorous self-shielding methods in criticality design-oriented codes. 6 refs., 11 figs., 3 tabs

  6. Simulation calculations using the code Geant III for the EUROGAM device

    Energy Technology Data Exchange (ETDEWEB)

    Beck, F A; Curien, D; Duchene, G; France, G de; Wei, L [Strasbourg-1 Univ., 67 (France). Centre de Recherches Nucleaires

    1992-08-01

    Simulation calculations are good tools to determine, at a low cost, the characteristics of a detector. It enables to change the geometry of the counter in an iterative way to optimize its response leading to the best performances for the whole multi-detector device. This kind of calculations have been performed using the Geant III code for the EUROGAM device. (author). 3 tabs., 5 figs.

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

    International Nuclear Information System (INIS)

    El-Osery, I.A.

    1981-01-01

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

  8. Cell-assembly coding in several memory processes.

    Science.gov (United States)

    Sakurai, Y

    1998-01-01

    The present paper discusses why the cell assembly, i.e., an ensemble population of neurons with flexible functional connections, is a tenable view of the basic code for information processes in the brain. The main properties indicating the reality of cell-assembly coding are neurons overlaps among different assemblies and connection dynamics within and among the assemblies. The former can be detected as multiple functions of individual neurons in processing different kinds of information. Individual neurons appear to be involved in multiple information processes. The latter can be detected as changes of functional synaptic connections in processing different kinds of information. Correlations of activity among some of the recorded neurons appear to change in multiple information processes. Recent experiments have compared several different memory processes (tasks) and detected these two main properties, indicating cell-assembly coding of memory in the working brain. The first experiment compared different types of processing of identical stimuli, i.e., working memory and reference memory of auditory stimuli. The second experiment compared identical processes of different types of stimuli, i.e., discriminations of simple auditory, simple visual, and configural auditory-visual stimuli. The third experiment compared identical processes of different types of stimuli with or without temporal processing of stimuli, i.e., discriminations of elemental auditory, configural auditory-visual, and sequential auditory-visual stimuli. Some possible features of the cell-assembly coding, especially "dual coding" by individual neurons and cell assemblies, are discussed for future experimental approaches. Copyright 1998 Academic Press.

  9. 3-D extension C5G7 MOX benchmark calculation using threedant code

    International Nuclear Information System (INIS)

    Kim, H.Ch.; Han, Ch.Y.; Kim, J.K.; Na, B.Ch.

    2005-01-01

    It pursued the benchmark on deterministic 3-D MOX fuel assembly transport calculations without spatial homogenization (C5G7 MOX Benchmark Extension). The goal of this benchmark is to provide a more through test results for the abilities of current available 3-D methods to handle the spatial heterogeneities of reactor core. The benchmark requires solutions in the form of normalized pin powers as well as the eigenvalue for each of the control rod configurations; without rod, with A rods, and with B rods. In this work, the DANTSYS code package was applied to analyze the 3-D Extension C5G7 MOX Benchmark problems. The THREEDANT code within the DANTSYS code package, which solves the 3-D transport equation in x-y-z, and r-z-theta geometries, was employed to perform the benchmark calculations. To analyze the benchmark with the THREEDANT code, proper spatial and angular approximations were made. Several calculations were performed to investigate the effects of the different spatial approximations on the accuracy. The results from these sensitivity studies were analyzed and discussed. From the results, it is found that the 4*4 grid per pin cell is sufficiently refined so that very little benefit is obtained by increasing the mesh size. (authors)

  10. PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tsuneo; Sugi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iijima, Shungo; Nishigori, Takeo

    1999-06-01

    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,{gamma}), (n,n`), (n,p), (n,{alpha}), (n,d), (n,t), (n,{sup 3}He), (n,2n), (n,n`p), (n,n`{alpha}), (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)

  11. US/JAERI calculational benchmarks for nuclear data and codes intercomparison. Article 8

    International Nuclear Information System (INIS)

    Youssef, M.Z.; Jung, J.; Sawan, M.E.; Nakagawa, M.; Mori, T.; Kosako, K.

    1986-01-01

    Prior to analyzing the integral experiments performed at the FNS facility at JAERI, both US and JAERI's analysts have agreed upon four calculational benchmark problems proposed by JAERI to intercompare results based on various codes and data base used independently by both countries. To compare codes the same data base is used (ENDF/B-IV). To compare nuclear data libraries, common codes were applied. Some of the benchmarks chosen were geometrically simple and consisted of a single material to clearly identify sources of discrepancies and thus help in analysing the integral experiments

  12. Post-test calculation and uncertainty analysis of the experiment QUENCH-07 with the system code ATHLET-CD

    International Nuclear Information System (INIS)

    Austregesilo, Henrique; Bals, Christine; Trambauer, Klaus

    2007-01-01

    In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B 4 C oxidation do not affect significantly the total calculated hydrogen release rates

  13. Resolution of the multigroup scattering equation in a one-dimensional geometry and subsidiary calculations: the MUDE code; Resolution de l'equation multigroupe de la diffusion dans une geometrie a une dimension et calculs annexes: code MUDE

    Energy Technology Data Exchange (ETDEWEB)

    Bore, C; Dandeu, Y; Saint-Amand, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    MUDE is a nuclear code written in FORTRAN II for IBM 7090-7094. It resolves a system of difference equations approximating to the one-dimensional multigroup neutron scattering problem. More precisely, this code makes it possible to: 1. Calculate the critical condition of a reactor (k{sub eff}, critical radius, critical composition) and the corresponding fluxes; 2. Calculate the associated fluxes and various subsidiary results; 3. Carry out perturbation calculations; 4. Study the propagation of fluxes at a distance; 5. Estimate the relative contributions of the cross sections (macroscopic or microscopic); 6. Study the changes with time of the composition of the reactor. (authors) [French] MUDE est un code nucleaire ecrit en FORTRAN II pour IBM 7090-7094. Il resout un systeme d'equations aux differences approchant le probleme de diffusion neutronique multigroupe a une dimension. Plus precisement ce code permet de: 1. Calculer la condition critique d'un reacteur (k{sub eff}, rayon critique, composition critique) et les flux correspondants; 2. Calculer les flux adjoints et divers resultats connexes; 3. Effectuer des calculs de perturbation; 4. Etudier la propagation des flux a longue distance; 5. Ponderer des sections efficaces (macroscopiques ou microscopiques); 6. Etudier l'evolution de la composition du reacteur au cours du temps. (auteurs)

  14. Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

    Science.gov (United States)

    Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

    2018-02-01

    The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

  15. Calculation of pellet radial power distributions with a Monte Carlo burnup code

    International Nuclear Information System (INIS)

    Suzuki, Motomu; Yamamoto, Toru; Nakata, Tetsuo

    2010-01-01

    The Japan Nuclear Energy Safety Organization (JNES) has been working on an irradiation test program of high-burnup MOX fuel at Halden Boiling Water Reactor (HBWR). MOX and UO 2 fuel rods had been irradiated up to about 64 GWd/t (rod avg.) as a Japanese utilities research program (1st phase), and using those fuel rods, in-situ measurement of fuel pellet centerline temperature was done during the 2nd phase of irradiation as the JNES test program. As part of analysis of the temperature data, power distributions in a pellet radial direction were analyzed by using a Monte Carlo burnup code MVP-BURN. In addition, the calculated results of deterministic burnup codes SRAC and PLUTON for the same problem were compared with those of MVP-BURN to evaluate their accuracy. Burnup calculations with an assembly model were performed by using MVP-BURN and those with a pin cell model by using SRAC and PLUTON. The cell pitch and, therefore, fuel to moderator ratio in the pin cell calculation was determined from the comparison of neutron energy spectra with those of MVP-BURN. The fuel pellet radial distributions of burnup and fission reaction rates at the end of the 1st phase irradiation were compared between the three codes. The MVP-BURN calculation results show a large peaking in the burnup and fission rates in the pellet outer region for the UO 2 and MOX pellets. The SRAC calculations give very close results to those of the MVP-BURN. On the other hand, the PLUTON calculations show larger burnup for the UO 2 and lower burnup for the MOX pellets in the pellet outer region than those of MVP-BURN, which lead to larger fission rates for the UO 2 and lower fission rates for the MOX pellets, respectively. (author)

  16. An evaluation of calculation parameters in the EGSnrc/BEAMnrc Monte Carlo codes and their effect on surface dose calculation

    International Nuclear Information System (INIS)

    Kim, Jung-Ha; Hill, Robin; Kuncic, Zdenka

    2012-01-01

    The Monte Carlo (MC) method has proven invaluable for radiation transport simulations to accurately determine radiation doses and is widely considered a reliable computational measure that can substitute a physical experiment where direct measurements are not possible or feasible. In the EGSnrc/BEAMnrc MC codes, there are several user-specified parameters and customized transport algorithms, which may affect the calculation results. In order to fully utilize the MC methods available in these codes, it is essential to understand all these options and to use them appropriately. In this study, the effects of the electron transport algorithms in EGSnrc/BEAMnrc, which are often a trade-off between calculation accuracy and efficiency, were investigated in the buildup region of a homogeneous water phantom and also in a heterogeneous phantom using the DOSRZnrc user code. The algorithms and parameters investigated include: boundary crossing algorithm (BCA), skin depth, electron step algorithm (ESA), global electron cutoff energy (ECUT) and electron production cutoff energy (AE). The variations in calculated buildup doses were found to be larger than 10% for different user-specified transport parameters. We found that using BCA = EXACT gave the best results in terms of accuracy and efficiency in calculating buildup doses using DOSRZnrc. In addition, using the ESA = PRESTA-I option was found to be the best way of reducing the total calculation time without losing accuracy in the results at high energies (few keV ∼ MeV). We also found that although choosing a higher ECUT/AE value in the beam modelling can dramatically improve computation efficiency, there is a significant trade-off in surface dose uncertainty. Our study demonstrates that a careful choice of user-specified transport parameters is required when conducting similar MC calculations. (note)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  19. DISP1 code

    International Nuclear Information System (INIS)

    Vokac, P.

    1999-12-01

    DISP1 code is a simple tool for assessment of the dispersion of the fission product cloud escaping from a nuclear power plant after an accident. The code makes it possible to tentatively check the feasibility of calculations by more complex PSA3 codes and/or codes for real-time dispersion calculations. The number of input parameters is reasonably low and the user interface is simple enough to allow a rapid processing of sensitivity analyses. All input data entered through the user interface are stored in the text format. Implementation of dispersion model corrections taken from the ARCON96 code enables the DISP1 code to be employed for assessment of the radiation hazard within the NPP area, in the control room for instance. (P.A.)

  20. About the application of MCNP4 code in nuclear reactor core design calculations

    International Nuclear Information System (INIS)

    Svarny, J.

    2000-01-01

    This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

  1. HETERO code, heterogeneous procedure for reactor calculation; Program Hetero, heterogeni postupak proracuna reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, S M; Raisic, N M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1966-11-15

    This report describes the procedure for calculating the parameters of heterogeneous reactor system taking into account the interaction between fuel elements related to established geometry. First part contains the analysis of single fuel element in a diffusion medium, and criticality condition of the reactor system described by superposition of elements interactions. the possibility of performing such analysis by determination of heterogeneous system lattice is described in the second part. Computer code HETERO with the code KETAP (calculation of criticality factor {eta}{sub n} and flux distribution) is part of this report together with the example of RB reactor square lattice.

  2. Development of a two-dimensional simulation code (koad) including atomic processes for beam direct energy conversion

    International Nuclear Information System (INIS)

    Yamamoto, Y.; Yoshikawa, K.; Hattori, Y.

    1987-01-01

    A two-dimensional simulation code for the beam direct energy conversion called KVAD (Kyoto University Advanced DART) including various loss mechanisms has been developed, and shown excellent agreement with the authors' experiments using the He + beams. The beam direct energy converter (BDC) is the device to recover the kinetic energy of unneutralized ions in the neutral beam injection (NBI) system directly into electricity. The BDC is very important and essential not only to the improvements of NBI system efficiency, but also to the relaxation of high heat flux problems on the beam dump with increase of injection energies. So far no simulation code could have successfully predicted BDC experimental results. The KUAD code applies, an optimized algorithm for vector processing, the finite element method (FEM) for potential calculation, and a semi-automatic method for spatial segmentations. Since particle trajectories in the KVAD code are analytically solved, very high speed tracings of the particle could be achieved by introducing an adjacent element matrix to identify the neighboring triangle elements and electrodes. Ion space charges are also analytically calculated by the Cloud in Cell (CIC) method, as well as electron space charges. Power losses due to atomic processes can be also evaluated in the KUAD code

  3. Calculation of mixed HEU-LEU cores for the HOR research reactor with the scale code system

    International Nuclear Information System (INIS)

    Leege, P.F.A. de; Gibcus, H.P.M.; Hoogenboom, J.E.; Vries, J.W. de

    1997-01-01

    The HOR reactor of Interfaculty Reactor Institute (IRI), Delft, The Netherlands, will be converted to use low enriched fuel (LEU) assemblies. As there are still many usable high enriched (HEU) fuel assemblies present, there will be a considerable reactor operation time with mixed cores with both HEU and LEU fuel assemblies. At IRI a comprehensive reactor physics code system and evaluated nuclear data is implemented for detailed core calculations. One of the backbones of the IRI code system is the well-known SCALE code system package. Full core calculations are performed with the diffusion theory code BOLD VENTURE, the nodal code SILWER, and the Monte Carlo code KENO Va. Results are displayed of a strategy from a HEU core to a mixed HEU-LEU core and eventually a LEU core. (author)

  4. OPT-TWO: Calculation code for two-dimensional MOX fuel models in the optimum concentration distribution

    International Nuclear Information System (INIS)

    Sato, Shohei; Okuno, Hiroshi; Sakai, Tomohiro

    2007-08-01

    OPT-TWO is a calculation code which calculates the optimum concentration distribution, i.e., the most conservative concentration distribution in the aspect of nuclear criticality safety, of MOX (mixed uranium and plutonium oxide) fuels in the two-dimensional system. To achieve the optimum concentration distribution, we apply the principle of flattened fuel importance distribution with which the fuel system has the highest reactivity. Based on this principle, OPT-TWO takes the following 3 calculation steps iteratively to achieve the optimum concentration distribution with flattened fuel importance: (1) the forward and adjoint neutron fluxes, and the neutron multiplication factor, with TWOTRAN code which is a two-dimensional neutron transport code based on the SN method, (2) the fuel importance, and (3) the quantity of the transferring fuel. In OPT-TWO, the components of MOX fuel are MOX powder, uranium dioxide powder and additive. This report describes the content of the calculation, the computational method, and the installation method of the OPT-TWO, and also describes the application method of the criticality calculation of OPT-TWO. (author)

  5. Nuclear Characteristics of SPNDs and Preliminary Calculation of Hybrid Fixed Incore Detector with Monte Carlo Code

    International Nuclear Information System (INIS)

    Koo, Bon Seung; Lee, Kyung Hoon; Song, Jae Seung; Park, Sang Yoon

    2013-01-01

    In this paper, the basic nuclear characteristics of major emitter materials were surveyed. In addition, preliminary calculations of Cobalt-Vanadium fixed incore detector were performed using the Monte Carlo code. Calculational results were cross-checked by KARMA. KARMA is a two-dimensional multigroup transport theory code developed by the KAERI and approved by Korean regularity agency to be employed as a nuclear design tool for a Korean commercial pressurizer water reactor. The nuclear characteristics of the major emitter materials were surveyed, and preliminary calculations of the hybrid fixed incore detector were performed with the MCNP code. The eigenvalue and pin-by-pin fission power distributions were calculated and showed good agreement with the KARMA calculation results. As future work, gamma power distributions as well as several types of XS of the emitter, insulator, and collector regions for a Co-V ICI assembly will be evaluated and compared

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

    International Nuclear Information System (INIS)

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

    1981-04-01

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

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

    International Nuclear Information System (INIS)

    Kikuchi, Shiroh; Takeuchi, Suehiro

    1987-11-01

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

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

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1984-12-01

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

  9. NAUAHYGROS - A code for calculating aerosol behavior in nuclear power plant containments following a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Sher, R. [Rudolph Sher Associates, Stanford, CA (United States); Li, J. [Polestar Applied Technology, Inc., Los Altos, CA (United States)

    1995-02-01

    NAUAHYGROS is a computer code to calculate the behavior of fission product and other aerosol particles in the containment of a nuclear reactor following a severe accident. It is an extension of the German code NAUA, which has been in widespread use for many years. Early versions of NAUA treated various aerosol phenomena in dry atmospheres, including aerosol agglomeration, diffusion (plateout), and settling processes. Later versions added treatments of steam condensation on particles in saturated or supersaturated containment atmospheres. The importance of these condensation effects on aerosol removal rates was demonstrated in large scale simulated containment tests. The additional features incorporated in NAUAHYGROS include principally a treatment of steam condensation on hygroscopic aerosols, which can grow as a result of steam condensation even in superheated atmospheres, and improved modelling of steam condensation on the walls of the containment. The code has been validated against the LACE experiments.

  10. A point-kernel shielding code for calculations of neutron and secondary gamma-ray 1cm dose equivalents: PKN

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Tanaka, Shun-ichi

    1991-09-01

    A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)

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

    International Nuclear Information System (INIS)

    Yamaguchi, Yasuhiro

    1991-01-01

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

  12. Implementation of refined core thermal-hydraulic calculation feature in the MARS/MASTER code

    International Nuclear Information System (INIS)

    Joo, H. K.; Jung, J. J.; Cho, B. O.; Ji, S. K.; Lee, W. J.; Jang, M. H.

    2000-01-01

    As an effort to enhance the fidelity of the core thermal/hydraulic calculation in the MARS/MASTER code, a best-estimate system/core coupled code, the COBRA-III module of MASTER is activated that enables refined core T/H calculations. Since the COBRA-III module is capable of using fuel-assembly sized nodes, the resolution of the T/H solution is high so that accurate incorporation of local T/H feedback effects becomes possible. The COBRA-III module is utilized such that the refined core T/H calculation is performed using the coarse-mesh flow boundary conditions specified by MARS at both ends of the core. The results of application to the OECD MSLB benchmark analysis indicate that the local peaking factor can be reduced by upto 15% with the refined calculation through the accurate representation of the local Doppler effect evaluation, although the prediction of the global transient behaviors such as the total core power change remain essentially unaffected

  13. MVP/GMVP Version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods (Translated document)

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

    2017-03-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

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

    International Nuclear Information System (INIS)

    Heeb, C.M.

    1991-03-01

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

  15. Comparison of a semi-empirical method with some model codes for gamma-ray spectrum calculation

    Energy Technology Data Exchange (ETDEWEB)

    Sheng, Fan; Zhixiang, Zhao [Chinese Nuclear Data Center, Beijing, BJ (China)

    1996-06-01

    Gamma-ray spectra calculated by a semi-empirical method are compared with those calculated by the model codes such as GNASH, TNG, UNF and NDCP-1. The results of the calculations are discussed. (2 tabs., 3 figs.).

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

    International Nuclear Information System (INIS)

    Asai, Kiyoshi; Shinozawa, Naohisa; Ishikawa, Hirohiko; Chino, Masamichi; Hayashi, Takashi

    1983-02-01

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

  17. EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

    International Nuclear Information System (INIS)

    Jones, D.B.

    1986-01-01

    EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

  18. Calculation Of Fuel Burnup And Radionuclide Inventory In The Syrian Miniature Neutron Source Reactor Using The GETERA Code

    International Nuclear Information System (INIS)

    Khattab, K.; Dawahra, S.

    2011-01-01

    Calculations of the fuel burnup and radionuclide inventory in the Syrian Miniature Neutron Source Reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer and has a bigger library of isotopes and more accurate. (author)

  19. Development of EASYQAD version β: A Visualization Code System for QAD-CGGP-A Gamma and Neutron Shielding Calculation Code

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Soon Young; Shin, Chang Ho; Kim, Jong Kyung

    2007-01-01

    EASYQAD had been previously developed by using MATLAB GUI (Graphical User Interface) in order to perform conveniently gamma and neutron shielding calculations at Hanyang University. It had been completed as version α of radiation shielding analysis code. In this study, EASYQAD was upgraded to version β with many additional functions and more user-friendly graphical interfaces. For general users to run it on Windows XP environment without any MATLAB installation, this version was developed into a standalone code system

  20. Computer code for shielding calculations of x-rays rooms

    International Nuclear Information System (INIS)

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

    2015-01-01

    The building an effective barrier against ionizing radiation present in radiographic rooms requires consideration of many variables. The methodology used for thickness specification of primary and secondary, barrier of a traditional radiographic room, considers the following factors: Use Factor, Occupational Factor, distance between the source and the wall, Workload, Kerma in the air and distance between the patient and the source. With these data it was possible to develop a computer code, which aims to identify and use variables in functions obtained through graphics regressions provided by NCRP-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) report, for shielding calculation of room walls, and the walls of the dark room and adjacent areas. With the implemented methodology, it was made a code validation by comparison of results with a study case provided by the report. The obtained values for thickness comprise different materials such as concrete, lead and glass. After validation it was made a case study of an arbitrary radiographic room.The development of the code resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in september/2011. (authors)

  1. Carmen system: a code block for neutronic PWR calculation by diffusion theory with spacedependent feedback effects

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J.M.

    1982-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Shirakawa, Toshihiko; Hatanaka, Koichiro

    2001-11-01

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

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

    International Nuclear Information System (INIS)

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

    1983-02-01

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

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

    International Nuclear Information System (INIS)

    Fry, C.J.

    1985-08-01

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

  5. Advanced local dose rate calculations with the Monte Carlo code MCNP for plutonium nitrate storage containers

    International Nuclear Information System (INIS)

    Quade, U.

    1994-01-01

    Neutron- und Gamma dose rate calculations were performed for the storage containers filled with plutonium nitrate of the MOX fabrication facility of Siemens. For the particle transport calculations the Monte Carlo Code MCNP 4.2 was used. The calculated results were compared with experimental dose rate measurements. It can be stated that the choice of the code system was appropriate since all aspects of the many facettes of the problem were well reproduced in the calculations. The position dependency as well as the influence of the shieldings, the reflections and the mutual influences of the sources were well described by the calculations for the gamma and for the neutron dose rates. However, good agreement with the experimental results on the gamma dose rates could only be reached when the lead shielding of the detector was integrated into the geometry modelling of the calculations. For some few cases of thick shieldings and soft gamma ray sources the statistics of the calculational results were not sufficient. In such cases more elaborate variance reduction methods must be applied in future calculations. Thus the MCNP code in connection with NGSRC has been proven as an effective tool for the solution of this type of problems. (orig./HP) [de

  6. Calculation of nuclear data for incident energies to 200 MeV with the FKK-GNASH code system

    International Nuclear Information System (INIS)

    Chadwick, M.B.; Young, P.G.

    1993-02-01

    We describe how the FKK-GNASH code system has been extended to calculate nucleon-induced reactions up to 200 MeV, and used to predict (p,xn) and (p,xp) cross sections on 208 Pb at incident energies of 25, 45, 80 and 160 MeV, for an intermediate energy code intercomparison. Details of the reaction mechanisms calculated by FKK-GNASH are given, and the calculational procedure is described

  7. Decay Heat Calculations for Reactors: Development of a Computer Code ADWITA

    International Nuclear Information System (INIS)

    Raj, Devesh

    2015-01-01

    Estimation of release of energy (decay heat) over an extended period of time after termination of neutron induced fission is necessary for determining the heat removal requirements when the reactor is shutdown, and for fuel storage and transport facilities as well as for accident studies. A Fuel Cycle Analysis Code, ADWITA (Activation, Decay, Waste Incineration and Transmutation Analysis) which can generate inventory based on irradiation history and calculate radioactivity and decay heat for extended period of cooling, has been written. The method and data involved in Fuel Cycle Analysis Code ADWITA and some results obtained shall also be presented. (author)

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

    International Nuclear Information System (INIS)

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

    1992-11-01

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

  9. Description of ground motion data processing codes: Volume 3

    International Nuclear Information System (INIS)

    Sanders, M.L.

    1988-02-01

    Data processing codes developed to process ground motion at the Nevada Test Site for the Weapons Test Seismic Investigations Project are used today as part of the program to process ground motion records for the Nevada Nuclear Waste Storage Investigations Project. The work contained in this report documents and lists codes and verifies the ''PSRV'' code. 39 figs

  10. Verification of computer code for calculation of coolant radiolysis in the VVER reactor core with regard for boiling in its upper part

    Energy Technology Data Exchange (ETDEWEB)

    Arkhipov, O.P.; Kabakchi, S.A. [OKB Gidropress, Podolsk, Moscow (Russian Federation)

    2010-07-01

    Code Bora for WWER coolant radiolysis calculation considering single jets boiling in the reactor core top part is developed on the basis of computer codes MOPABA-H2 (radiolysis of aqueous solutions) and SteamRad (radiolysis of vapor). Physico-chemical processes taking place in boiling core coolant are complex and diversified. Still, for the solution of certain problems their simulation can be simplified. The approach of reasonable simplification was used for development of code Bora: mathematical model assumed is purposed for simulation of phenomena only in the area of interest; the number of simulated chemical reactions and particles shall be reasonably minimum; complexity of interphase mass transfer calculation procedure shall be adequate to actually available accuracy of modeling. The analysis of new experimental initial yields of water radiolysis products data and kinetic parameters of elementary chemical reactions with their participation has been carried out. Some changes have been introduced in the mechanism of liquid water and aqueous solutions of ammonia radiolysis have been significantly revised on the basis of this analysis. Examples of the calculations provided for code Bora verification are presented. Despite of very simple simulation of interphase mass transfer, Bora allows to obtain average chemical composition of two-phase coolant at BWR core outlet with the accuracy sufficient for engineering calculations. The report also presents the results of two-phase coolant chemical composition test calculation for reactor core top part coolant boiling in pressurized water reactor. (author)

  11. YNOGK: A NEW PUBLIC CODE FOR CALCULATING NULL GEODESICS IN THE KERR SPACETIME

    Energy Technology Data Exchange (ETDEWEB)

    Yang Xiaolin; Wang Jiancheng, E-mail: yangxl@ynao.ac.cn [National Astronomical Observatories, Yunnan Observatory, Chinese Academy of Sciences, Kunming 650011 (China)

    2013-07-01

    Following the work of Dexter and Agol, we present a new public code for the fast calculation of null geodesics in the Kerr spacetime. Using Weierstrass's and Jacobi's elliptic functions, we express all coordinates and affine parameters as analytical and numerical functions of a parameter p, which is an integral value along the geodesic. This is the main difference between our code and previous similar ones. The advantage of this treatment is that the information about the turning points does not need to be specified in advance by the user, and many applications such as imaging, the calculation of line profiles, and the observer-emitter problem, become root-finding problems. All elliptic integrations are computed by Carlson's elliptic integral method as in Dexter and Agol, which guarantees the fast computational speed of our code. The formulae to compute the constants of motion given by Cunningham and Bardeen have been extended, which allow one to readily handle situations in which the emitter or the observer has an arbitrary distance from, and motion state with respect to, the central compact object. The validation of the code has been extensively tested through applications to toy problems from the literature. The source FORTRAN code is freely available for download on our Web site http://www1.ynao.ac.cn/{approx}yangxl/yxl.html.

  12. YNOGK: A NEW PUBLIC CODE FOR CALCULATING NULL GEODESICS IN THE KERR SPACETIME

    International Nuclear Information System (INIS)

    Yang Xiaolin; Wang Jiancheng

    2013-01-01

    Following the work of Dexter and Agol, we present a new public code for the fast calculation of null geodesics in the Kerr spacetime. Using Weierstrass's and Jacobi's elliptic functions, we express all coordinates and affine parameters as analytical and numerical functions of a parameter p, which is an integral value along the geodesic. This is the main difference between our code and previous similar ones. The advantage of this treatment is that the information about the turning points does not need to be specified in advance by the user, and many applications such as imaging, the calculation of line profiles, and the observer-emitter problem, become root-finding problems. All elliptic integrations are computed by Carlson's elliptic integral method as in Dexter and Agol, which guarantees the fast computational speed of our code. The formulae to compute the constants of motion given by Cunningham and Bardeen have been extended, which allow one to readily handle situations in which the emitter or the observer has an arbitrary distance from, and motion state with respect to, the central compact object. The validation of the code has been extensively tested through applications to toy problems from the literature. The source FORTRAN code is freely available for download on our Web site http://www1.ynao.ac.cn/~yangxl/yxl.html.

  13. Verification of RRC Ki code package for neutronic calculations of WWER core with GD

    International Nuclear Information System (INIS)

    Aleshin, S.S.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Pavlov, V.I.; Pavlovitchev, A.M.; Sidorenko, V.D.; Tsvetkov, V.M.

    2001-01-01

    The report presented is concerned with verification results of TVS-M/PERMAK-A/BIPR-7A code package for WWERs neutronic calculation as applied to calculation of systems containing U-GD pins. The verification is based on corresponded benchmark calculations, data critical experiments and on operation data obtained WWER units with Gd. The comparison results are discussed (Authors)

  14. Implantation of a new calculation method of fuel depletion in the CITHAM code

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1985-01-01

    It is evaluated the accuracy of the linear aproximation method used in the CITHAN code to obtain the solution of depletion equations. Results are compared with the Benchmark problem. The convenience of depletion chain before criticality calculations is analysed. The depletion calculation was modified using linear combination technic of linear chains. (M.C.K.) [pt

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

    International Nuclear Information System (INIS)

    Gren, Milan

    2010-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Blackburn, P E

    1977-12-01

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

  17. BARS - a heterogeneous code for 3D pin-by-pin LWR steady-state and transient calculation

    International Nuclear Information System (INIS)

    Avvakumov, A.V.; Malofeev, V.M.

    2000-01-01

    A 3D pin-by-pin dynamic model for LWR detailed calculation was developed. The model is based on a coupling of the BARS neutronic code with the RELAP5/MOD3.2 thermal hydraulic code. This model is intended to calculate a fuel cycle, a xenon transient, and a wide range of reactivity initiated accidents in a WWER and a PWR. Galanin-Feinberg heterogeneous method was realized in the BARS code. Some results for a validation of the heterogeneous method are presented for reactivity coefficients, a pin-by-pin power distribution, and a fast pulse transient. (Authors)

  18. A fast, user-friendly code for calculating magnetohydrodynamic equilibria

    International Nuclear Information System (INIS)

    Haney, S.W.; Freidberg, J.P.; Solomon, C.J.

    1995-01-01

    Using variational techniques, we have developed a fast, user-friendly code for computing approximate, but highly accurate fixed boundary magnetohydrodynamic equilibria for tokamak plasmas. The variational procedure simplifies the problem---a two-dimensional nonlinear partial differential equation---to a set of nonlinear algebraic equations. The reduced problem can be readily solved on workstations or personal computers. This allows us to exploit sophisticated graphical user interfaces that make supplying calculation data and viewing results easy. This ease-of-use, along with the semianalytic nature of our calculation, allows researchers to routinely incorporate equilibrium information into their work. It also provides a tool for educators teaching fusion theory. We describe the variational formulation, the speed and accuracy of the computer implementation, and the design and operation of a user-friendly graphical interface

  19. Introduction of SCIENCE code package

    International Nuclear Information System (INIS)

    Lu Haoliang; Li Jinggang; Zhu Ya'nan; Bai Ning

    2012-01-01

    The SCIENCE code package is a set of neutronics tools based on 2D assembly calculations and 3D core calculations. It is made up of APOLLO2F, SMART and SQUALE and used to perform the nuclear design and loading pattern analysis for the reactors on operation or under construction of China Guangdong Nuclear Power Group. The purpose of paper is to briefly present the physical and numerical models used in each computation codes of the SCIENCE code pack age, including the description of the general structure of the code package, the coupling relationship of APOLLO2-F transport lattice code and SMART core nodal code, and the SQUALE code used for processing the core maps. (authors)

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

    International Nuclear Information System (INIS)

    Simon-Cornu, Marie; Mourlon, Christophe; Bordy, J.M.; Daures, J.; Dusiac, D.; Moignau, F.; Gouriou, J.; Million, M.; Moreno, B.; Chabert, I.; Lazaro, D.; Barat, E.; Dautremer, T.; Montagu, T.; Agelou, M.; De Carlan, L.; Patin, D.; Le Loirec, C.; Dupuis, P.; Gassa, F.; Guerin, L.; Batalla, A.; Leni, Pierre-Emmanuel; Laurent, Remy; Gschwind, Regine; Makovicka, Libor; Henriet, Julien; Salomon, Michel; Vivier, Alain; Lopez, Gerald; Dossat, C.; Pourrouquet, P.; Thomas, J.C.; Sarie, I.; Peyrard, P.F.; Chatry, N.; Lavielle, D.; Loze, R.; Brun, E.; Damian, F.; Diop, C.; Dumonteil, E.; Hugot, F.X.; Jouanne, C.; Lee, Y.K.; Malvagi, F.; Mazzolo, A.; Petit, O.; Trama, J.C.; Visonneau, T.; Zoia, A.; Courageot, Estelle; Gaillard-Lecanu, Emmanuelle; Kutschera, Reinald; Le Meur, Gaelle; Uzio, Fabien; De Conto, Celine; Gschwind, Regine; Makovicka, Libor; Farah, Jad; Martinetti, Florent; Sayah, Rima; Donadille, Laurent; Herault, Joel; Delacroix, Sabine; Nauraye, Catherine; Lee, Choonsik; Bolch, Wesley; Clairand, Isabelle; Horodynski, Jean-Michel; Pauwels, Nicolas; Robert, Pierre; VOLLAIRE, Joachim; Nicoletti, C.; Kitsos, S.; Tardy, M.; Marchaud, G.; Stankovskiy, Alexey; Van Den Eynde, Gert; Fiorito, Luca; Malambu, Edouard; Dreuil, Serge; Mougeot, X.; Be, M.M.; Bisch, C.; Villagrasa, C.; Dos Santos, M.; Clairand, I.; Karamitros, M.; Incerti, S.; Petitguillaume, Alice; Franck, Didier; Desbree, Aurelie; Bernardini, Michela; Labriolle-Vaylet, Claire de; Gnesin, Silvano; Leadermann, Jean-Pascal; Paterne, Loic; Bochud, Francois O.; Verdun, Francis R.; Baechler, Sebastien; Prior, John O.; Thomassin, Alain; Arial, Emmanuelle; Laget, Michael; Masse, Veronique; Saldarriaga Vargas, Clarita; Struelens, Lara; Vanhavere, Filip; Perier, Aurelien; Courageot, Estelle; Gaillard-Lecanu, Emmanuelle; Le-Meur, Gaelle; Monier, Catherine; Thers, Dominique; Le-Guen, Bernard; Blond, Serge; Cordier, Gerard; Le Roy, Maiwenn; De Carlan, Loic; Bordy, Jean-Marc; Caccia, Barbara; Andenna, Claudio; Charimadurai, Arun; Selvam, T Palani; Czarnecki, Damian; Zink, Klemens; Gschwind, Regine; Martin, Eric; Huot, Nicolas; Zoubair, Mariam; El Bardouni, Tarek; Lazaro, Delphine; Barat, Eric; Dautremer, Thomas; Montagu, Thierry; Chabert, Isabelle; Guerin, Lucie; Batalla, Alain; Moignier, C.; Huet, C.; Bassinet, C.; Baumann, M.; Barraux, V.; Sebe-Mercier, K.; Loiseau, C.; Batalla, A.; Makovicka, L.; Desnoyers, Yvon; Juhel, Gabriel; Mattera, Christophe; Tempier, Maryline

    2014-03-01

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

  1. Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system

    International Nuclear Information System (INIS)

    Dunn, M.E.; Greene, N.M.; Leal, L.C.

    1999-01-01

    Modern techniques for the establishment of criticality safety for fissile systems invariably require the use of neutronic transport codes with applicable cross-section data. Accurate cross-section data are essential for solving the Boltzmann Transport Equation for fissile systems. In the absence of applicable critical experimental data, the use of independent calculational methods is crucial for the establishment of subcritical limits. Moreover, there are various independent modern transport codes available to the criticality safety analyst (e.g., KENO V.a., MCNP, and MONK). In contrast, there is currently only one complete software package that processes data from the Version 6 format of the Evaluated Nuclear Data File (ENDF) to a format useable by criticality safety codes. To facilitate independent cross-section processing, Oak Ridge National Laboratory (ORNL) is upgrading the AMPX code system to enable independent processing of Version 6 formats using state-of-the-art procedures. The AMPX code system has been in continuous use at ORNL since the early 1970s and is the premier processor for providing multigroup cross sections for criticality safety analysis codes. Within the AMPX system, the module POLIDENT is used to access the resonance parameters in File 2 of an ENDF/B library, generate point cross-section data, and combine the cross sections with File 3 point data. At the heart of any point cross-section processing code is the generation of a suitable energy mesh for representing the data. The purpose of this work is to facilitate the AMPX upgrade through the development of a new and innovative energy meshing technique for processing point cross-section data

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Calculation of the effective dose from natural radioactivity in soil using MCNP code.

    Science.gov (United States)

    Krstic, D; Nikezic, D

    2010-01-01

    Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this work. Calculations have been done for the most common natural radionuclides in soil (238)U, (232)Th series and (40)K. A ORNL human phantoms and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs. The effective dose was calculated according to ICRP 74 recommendations. Conversion factors of effective dose per air kerma were determined. Results obtained here were compared with other authors. Copyright 2009 Elsevier Ltd. All rights reserved.

  5. Using of the Serpent code based on the Monte-Carlo method for calculation of the VVER-1000 fuel assembly characteristics

    Directory of Open Access Journals (Sweden)

    V. V. Galchenko

    2016-12-01

    Full Text Available The description of calculation scheme of fuel assembly for preparation of few-group characteristics is considered with help of Serpent code. This code uses the Monte-Carlo method and energy continuous microscopic data libraries. Serpent code is devoted for calculation of fuel assembly characteristics, burnup calculations and preparation of few-group homogenized macroscopic cross-sections. The results of verification simulations in comparison with other codes (WIMS, HELIOS, NESSEL etc., which are used for neutron-physical analysis of VVER type fuel, are presented.

  6. Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

    1987-02-01

    Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

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

    International Nuclear Information System (INIS)

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

    1980-07-01

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

  8. Timing comparison of two-dimensional discrete-ordinates codes for criticality calculations

    International Nuclear Information System (INIS)

    Miller, W.F. Jr.; Alcouffe, R.E.; Bosler, G.E.; Brinkley, F.W. Jr.; O'dell, R.D.

    1979-01-01

    The authors compare two-dimensional discrete-ordinates neutron transport computer codes to solve reactor criticality problems. The fundamental interest is in determining which code requires the minimum Central Processing Unit (CPU) time for a given numerical model of a reasonably realistic fast reactor core and peripherals. The computer codes considered are the most advanced available and, in three cases, are not officially released. The conclusion, based on the study of four fast reactor core models, is that for this class of problems the diffusion synthetic accelerated version of TWOTRAN, labeled TWOTRAN-DA, is superior to the other codes in terms of CPU requirements

  9. Improvement of the spallation-reaction simulation code by considering both the high-momentum intranuclear nucleons and the preequilibrium process

    International Nuclear Information System (INIS)

    Ishibashi, K.; Miura, Y.; Sakae, T.

    1990-01-01

    In the present study, intranuclear nucleons with a high momentum are introduced into intranuclear cascade calculation, and the preequilibrium effects are considered at the end of the cascade process. The improvements made in the HETC (High Energy Transport Code) are outlined, focusing on intranuclear nucleons with a high momentum, and termination of the intranuclear cascade process. Discussion is made of the cutoff energy, and Monte Carlo calculations based on an excitation model are presented and analyzed. The experimental high energy neutrons in the backward direction are successfully reproduced. The preequilibrium effect is considered in a local manner, and this is introduced as a simple probability density function for terminating the intranuclear cascade process. The resultant neutron spectra reproduce the shoulders of the experimental data in the region of 20 to 50 MeV. The exciton model is coded with a Monte Carlo algorithm. The results of the exciton model calculation is not so appreciable except for intermediate energy neutrons in the backward direction. (N.K.)

  10. Data processing with microcode designed with source coding

    Science.gov (United States)

    McCoy, James A; Morrison, Steven E

    2013-05-07

    Programming for a data processor to execute a data processing application is provided using microcode source code. The microcode source code is assembled to produce microcode that includes digital microcode instructions with which to signal the data processor to execute the data processing application.

  11. Investigating the Simulink Auto-Coding Process

    Science.gov (United States)

    Gualdoni, Matthew J.

    2016-01-01

    Model based program design is the most clear and direct way to develop algorithms and programs for interfacing with hardware. While coding "by hand" results in a more tailored product, the ever-growing size and complexity of modern-day applications can cause the project work load to quickly become unreasonable for one programmer. This has generally been addressed by splitting the product into separate modules to allow multiple developers to work in parallel on the same project, however this introduces new potentials for errors in the process. The fluidity, reliability and robustness of the code relies on the abilities of the programmers to communicate their methods to one another; furthermore, multiple programmers invites multiple potentially differing coding styles into the same product, which can cause a loss of readability or even module incompatibility. Fortunately, Mathworks has implemented an auto-coding feature that allows programmers to design their algorithms through the use of models and diagrams in the graphical programming environment Simulink, allowing the designer to visually determine what the hardware is to do. From here, the auto-coding feature handles converting the project into another programming language. This type of approach allows the designer to clearly see how the software will be directing the hardware without the need to try and interpret large amounts of code. In addition, it speeds up the programming process, minimizing the amount of man-hours spent on a single project, thus reducing the chance of human error as well as project turnover time. One such project that has benefited from the auto-coding procedure is Ramses, a portion of the GNC flight software on-board Orion that has been implemented primarily in Simulink. Currently, however, auto-coding Ramses into C++ requires 5 hours of code generation time. This causes issues if the tool ever needs to be debugged, as this code generation will need to occur with each edit to any part of

  12. Calculation of the RSG-GAS core using computer code citation-3D

    International Nuclear Information System (INIS)

    Taryo, T.; Rokhmadi

    1998-01-01

    Since core reactivity is one of the reactor safety parameters, this R and D has been carried out. To carry out the R and D, the code called WIMSD4 was used respectively for generating cross section and diffusion parameters. The code CITATION was then applied to estimate core reactivity in the RSG-GAS core. To verify the result of the calculation, data and information of the RSG-GAS Typical Working Core Were used. To Prove the codes reliably used, the case of all control elements down in the reactor core and that of all control rods up in the core were applied. The result taking into account those cases showed respectively that K eff are less and greater than unity (K eff eff >1)

  13. Development of continuous energy Monte Carlo burn-up calculation code MVP-BURN

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Nakagawa, Masayuki; Sasaki, Makoto

    2001-01-01

    Burn-up calculations based on the continuous energy Monte Carlo method became possible by development of MVP-BURN. To confirm the reliably of MVP-BURN, it was applied to the two numerical benchmark problems; cell burn-up calculations for High Conversion LWR lattice and BWR lattice with burnable poison rods. Major burn-up parameters have shown good agreements with the results obtained by a deterministic code (SRAC95). Furthermore, spent fuel composition calculated by MVP-BURN was compared with measured one. Atomic number densities of major actinides at 34 GWd/t could be predicted within 10% accuracy. (author)

  14. Arabic Natural Language Processing System Code Library

    Science.gov (United States)

    2014-06-01

    Adelphi, MD 20783-1197 This technical note provides a brief description of a Java library for Arabic natural language processing ( NLP ) containing code...for training and applying the Arabic NLP system described in the paper "A Cross-Task Flexible Transition Model for Arabic Tokenization, Affix...and also English) natural language processing ( NLP ), containing code for training and applying the Arabic NLP system described in Stephen Tratz’s

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

    International Nuclear Information System (INIS)

    Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I.; Santamarina, A.

    2005-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    International Nuclear Information System (INIS)

    Bellido, Luis F.

    1995-07-01

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

  18. The modified high-energy transport code, HETC, and design calculations for the SSC [Superconducting Super Collider

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Alsmiller, F.S.; Gabriel, T.A.; Hermann, O.W.; Bishop, B.L.

    1988-01-01

    The proposed Superconducting Super Collider (SSC) will have two circulating proton beams, each with an energy of 20 TeV. In order to perform detector and shield design calculations at these higher energies that are as accurate as possible, it is necessary to incorporate in the calculations the best available information on differential particle production from hadron-nucleus collisions. In this paper, the manner in which this has been done in the High-Energy Transport Code HETC will be described and calculated results obtained with the modified code will be compared with experimental data. 10 refs., 1 fig

  19. Applicability of vector processing to large-scale nuclear codes

    International Nuclear Information System (INIS)

    Ishiguro, Misako; Harada, Hiroo; Matsuura, Toshihiko; Okuda, Motoi; Ohta, Fumio; Umeya, Makoto.

    1982-03-01

    To meet the growing trend of computational requirements in JAERI, introduction of a high-speed computer with vector processing faculty (a vector processor) is desirable in the near future. To make effective use of a vector processor, appropriate optimization of nuclear codes to pipelined-vector architecture is vital, which will pose new problems concerning code development and maintenance. In this report, vector processing efficiency is assessed with respect to large-scale nuclear codes by examining the following items: 1) The present feature of computational load in JAERI is analyzed by compiling the computer utilization statistics. 2) Vector processing efficiency is estimated for the ten heavily-used nuclear codes by analyzing their dynamic behaviors run on a scalar machine. 3) Vector processing efficiency is measured for the other five nuclear codes by using the current vector processors, FACOM 230-75 APU and CRAY-1. 4) Effectiveness of applying a high-speed vector processor to nuclear codes is evaluated by taking account of the characteristics in JAERI jobs. Problems of vector processors are also discussed from the view points of code performance and ease of use. (author)

  20. A FACSIMILE code for calculating void swelling and creep, with vacancy loops present: version VS4

    International Nuclear Information System (INIS)

    Windsor, M.E.; Bullough, R.; Wood, M.H.

    1981-10-01

    This FACSIMILE code calculates void swelling and creep of irradiated materials, taking into account the effects of cavities, interstitial loops, vacancy loops, dislocation network and either grain boundaries or foil surfaces. The creep calculations are based on SIPA theory (stress induced preferred absorption), with no preferred nucleation. Either interactive or non-interactive options are available for the sink strength equations, but rate limitation is not incorporated. FACSIMILE is a computer program for solving simultaneous differential equations, and this VS4 code is one of a series of codes for calculating void swelling using increasingly complex theories. Other reports describing the VS1 and VS2 codes explain their use under control of the TSO system of the Harwell IBM 3033 computer, and explain the basic organization of the codes as required for use by FACSIMILE. The creep theory assumes that the material is under a constant uniaxial tensile stress during the irradiation. Three directions are considered for network parameters relative to the direction of the stress, and two directions for interstitial and vacancy loops. To give a full picture of these various contributions to the total creep, a large set of output parameter values are printed for each demanded dose value via a FORTRAN subroutine. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Adams, C. H.

    1976-07-01

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

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

    International Nuclear Information System (INIS)

    Adams, C.H.

    1976-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-01-01

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

  4. An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

    International Nuclear Information System (INIS)

    Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E.; Nikulshin, V.

    1996-09-01

    This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases

  5. SILENE and TDT: A code for collision probability calculations in XY geometries

    International Nuclear Information System (INIS)

    Sanchez, R.; Stankovski, Z.

    1993-01-01

    Collision probability methods are routinely used for cell and assembly multigroup transport calculations in core design tasks. Collision probability methods use a specialized tracking routine to compute neutron trajectories within a given geometric object. These trajectories are then used to generate the appropriate collision matrices in as many groups as required. Traditional tracking routines are based on open-quotes globalclose quotes geometric descriptions (such as regular meshes) and are not able to cope with the geometric detail required in actual core calculations. Therefore, users have to modify their geometry in order to match the geometric model accepted by the tracking routine, introducing thus a modeling error whose evaluation requires the use of a open-quotes referenceclose quotes method. Recently, an effort has been made to develop more flexible tracking routines either by directly adopting tracking Monte Carlo techniques or by coding of complicated geometries. Among these, the SILENE and TDT package is being developed at the Commissariat a l' Energie Atomique to provide routine as well as reference calculations in arbitrarily shaped XY geometries. This package combines a direct graphical acquisition system (SILENE) together with a node-based collision probability code for XY geometries (TDT)

  6. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  7. Development of NRESP98 Monte Carlo codes for the calculation of neutron response functions of neutron detectors. Calculation of the response function of spherical BF{sub 3} proportional counter

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, M.; Saito, K.; Ando, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-05-01

    The method to calculate the response function of spherical BF{sub 3} proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF{sub 3} proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within {+-}10%. (author)

  8. Improvement of group collapsing in TRANSX code

    International Nuclear Information System (INIS)

    Jeong, Hyun Tae; Kim, Young Cheol; Kim, Young In; Kim, Young Kyun

    1996-07-01

    A cross section generating and processing computer code TRANSX version 2.15 in the K-CORE system, being developed by the KAERI LMR core design technology development team produces various cross section input files appropriated for flux calculation options from the cross section library MATXS. In this report, a group collapsing function of TRANSX has been improved to utilize the zone averaged flux file RZFLUX written in double precision as flux weighting functions. As a result, an iterative calculation system using double precision RZFLUX consisting of the cross section data library file MATXS, the effective cross section producing and processing code TRANSX, and the transport theory calculation code TWODANT has been set up and verified through a sample model calculation. 4 refs. (Author)

  9. Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

    International Nuclear Information System (INIS)

    Moukhamadeev, R.; Suvorov, A.

    2000-01-01

    This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

  10. CALIPSO - a computer code for the calculation of fluiddynamics, thermohydraulics and changes of geometry in failing fuel elements of a fast breeder reactor

    International Nuclear Information System (INIS)

    Kedziur, F.

    1982-07-01

    The computer code CALIPSO was developed for the calculation of a hypothetical accident in an LMFBR (Liquid Metal Fast Breeder Reactor), where the failure of fuel pins is assumed. It calculates two-dimensionally the thermodynamics, fluiddynamics and changes in geometry of a single fuel pin and its coolant channel in a time period between failure of the pin and a state, at which the geometry is nearly destroyed. The determination of temperature profiles in the fuel pin cladding and the channel wall make it possible to take melting and freezing processes into account. Further features of CALIPSO are the variable channel cross section in order to model disturbances of the channel geometry as well as the calculation of two velocity fields including the consideration of virtual mass effects. The documented version of CALIPSO is especially suited for the calculation of the SIMBATH experiments carried out at the Kernforschungszentrum Karlsruhe, which simulate the above-mentioned accident. The report contains the complete documentation of the CALIPSO code: the modeling of the geometry, the equations used, the structure of the code and the solution procedure as well as the instructions for use with an application example. (orig.) [de

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

    International Nuclear Information System (INIS)

    VOOGD, J.A.

    1999-01-01

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

  12. A code for the calculation of self-absorption fractions of photons

    International Nuclear Information System (INIS)

    Jaegers, P.; Landsberger, S.

    1988-01-01

    Neutron activation analysis (NAA) is now a well-established technique used by many researchers and commercial companies. It is often wrongly assumed that these NAA methods are matrix independent over a wide variety of samples. Accuracy at the level of a few percent is often difficult to achieve, since components such as timing, pulse pile-up, high dead-time corrections, sample positioning, and chemical separations may severely compromise the results. One area that has received little attention is the calculation of the effect of self-absorption of gamma-rays (including low-energy ones) in samples, particularly those with major components of high-Z values. The analysis of trace components in lead samples is an obvious example, but other high-Z matrices such as various permutations and combinations of zinc, tin, lead, copper, silver, antimony, etc.; ore concentrates; and meteorites are also affected. The authors have developed a simple but effective personal-computer-compatible user-friendly code, however, which can calculate the amount of energy signal that is lost due to the presence of any amount of one or more Z components. The program is based on Dixon's paper of 1951 for the calculation of self-absorption corrections for linear, cylindrical, and spherical sources. To determine the self-absorption fraction of a photon in a source, the FORTRAN computer code SELFABS was written

  13. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)

  14. Testing of a Code for the Calculation of Spectra of Neutrons Produced in a Target of a Neutron Generator

    Science.gov (United States)

    Gaganov, V. V.

    2017-12-01

    The correctness of calculations performed with the SRIANG code for modeling the spectra of DT neutrons is estimated by comparing the obtained spectra to the results of calculations carried out with five different codes based on the Monte Carlo method.

  15. Improved response function calculations for scintillation detectors using an extended version of the MCNP code

    CERN Document Server

    Schweda, K

    2002-01-01

    The analysis of (e,e'n) experiments at the Darmstadt superconducting electron linear accelerator S-DALINAC required the calculation of neutron response functions for the NE213 liquid scintillation detectors used. In an open geometry, these response functions can be obtained using the Monte Carlo codes NRESP7 and NEFF7. However, for more complex geometries, an extended version of the Monte Carlo code MCNP exists. This extended version of the MCNP code was improved upon by adding individual light-output functions for charged particles. In addition, more than one volume can be defined as a scintillator, thus allowing the simultaneous calculation of the response for multiple detector setups. With the implementation of sup 1 sup 2 C(n,n'3 alpha) reactions, all relevant reactions for neutron energies E sub n <20 MeV are now taken into consideration. The results of these calculations were compared to experimental data using monoenergetic neutrons in an open geometry and a sup 2 sup 5 sup 2 Cf neutron source in th...

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

    International Nuclear Information System (INIS)

    Nikityuk, N.M.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Xhonneux, Andre; Allelein, Hans-Josef

    2014-01-01

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

  18. KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

    International Nuclear Information System (INIS)

    Kim, Young Gyun; Kim, Young Il

    2006-12-01

    Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006

  19. UNR. A code for processing unresolved resonance data for MCNP

    International Nuclear Information System (INIS)

    Hogenbirk, A.

    1994-09-01

    In neutron transport problems the correct treatment of self-shielding is important for those nuclei present in large concentrations. Monte Carlo calculations using continuous-energy cross section data, such as calculations with the code MCNP, offer the advantage that neutron transport is calculated in a very accurate way. Self-shielding in the resolved resonance region is taken into account exactly in MCNP. However, self-shielding in the unresolved resonance region can not be taken into account by MCNP, although the effect of it may be important in many applications. In this report a description is given of the computer code UNR. With this code problem-dependent cross section libraries can be produced for MCNP. In these libraries self-shielded cross section data in the unresolved resonance range are given, which are produced by NJOY-module UNRESR. It is noted, that the treatment for resonance self-shielding presented in this report is approximate. However, the current version of MCNP does not allow the use of probability tables, which would be a general solution. (orig.)

  20. Description of a heat transfer model suitable to calculate transient processes of turbocharged diesel engines with one-dimensional gas-dynamic codes

    Energy Technology Data Exchange (ETDEWEB)

    Galindo, J.; Lujan, J.M.; Serrano, J.R.; Dolz, V. [CMT-Motores Termicos, Universidad Politecnica de Valencia, Valencia (Spain); Guilain, S. [Renault s.a.s., Lardy (France)

    2006-01-15

    This paper describes a heat transfer model to be implemented in a global engine 1-D gas-dynamic code to calculate reciprocating internal combustion engine performance in steady and transient operations. A trade off between simplicity and accuracy has been looked for, in order to fit with the stated objective. To validate the model, the temperature of the exhaust manifold wall in a high-speed direct injection (HSDI) turbocharged diesel engine has been measured during a full load transient. In addition, an indirect assessment of the exhaust gas temperature during this transient process has been carried out. The results show good agreement between the measured and modelled data with good accuracy to predict the engine performance. A dual-walled air gap exhaust manifold has been tested in order to quantify the potential of exhaust gas thermal energy saving on engine transient performance. The experimental results together with the heat transfer model have been used to analyse the influence of thermal energy saving on dynamic performance during the load transient of an HSDI turbocharged diesel engine. (author)

  1. MADNIX a code to calculate prompt fission neutron spectra and average prompt neutron multiplicities

    International Nuclear Information System (INIS)

    Merchant, A.C.

    1986-03-01

    A code has been written and tested on the CDC Cyber-170 to calculate the prompt fission neutron spectrum, N(E), as a function of both the fissioning nucleus and its excitation energy. In this note a brief description of the underlying physical principles involved and a detailed explanation of the required input data (together with a sample output for the fission of 235 U induced by 14 MeV neutrons) are presented. Weisskopf's standard nuclear evaporation theory provides the basis for the calculation. Two important refinements are that the distribution of fission-fragment residual nuclear temperature and the cooling of the fragments as neutrons are emitted approximately taken into account, and also the energy dependence of the cross section for the inverse process of compound nucleus formation is included. This approach is then used to calculate the average number of prompt neutrons emitted per fission, v-bar p . At high excitation energies, where fission is still possible after neutron emission, the consequences of the competition between first, second and third chance fission on N(E) and v-bar p are calculated. Excellent agreement with all the examples given in the original work of Madland and Nix is obtained. (author) [pt

  2. The PHREEQE Geochemical equilibrium code data base and calculations

    International Nuclear Information System (INIS)

    Andersoon, K.

    1987-01-01

    Compilation of a thermodynamic data base for actinides and fission products for use with PHREEQE has begun and a preliminary set of actinide data has been tested for the PHREEQE code in a version run on an IBM XT computer. The work until now has shown that the PHREEQE code mostly gives satisfying results for specification of actinides in natural water environment. For U and Np under oxidizing conditions, however, the code has difficulties to converge with pH and Eh conserved when a solubility limit is applied. For further calculations of actinide and fission product specification and solubility in a waste repository and in the surrounding geosphere, more data are needed. It is necessary to evaluate the influence of the large uncertainties of some data. A quality assurance and a check on the consistency of the data base is also needed. Further work with data bases should include: an extension to fission products, an extension to engineering materials, an extension to other ligands than hydroxide and carbonate, inclusion of more mineral phases, inclusion of enthalpy data, a control of primary references in order to decide if values from different compilations are taken from the same primary reference and contacts and discussions with other groups, working with actinide data bases, e.g. at the OECD/NEA and at the IAEA. (author)

  3. Comparative calculations on selected two-phase flow phenomena using major PWR system codes

    International Nuclear Information System (INIS)

    1990-01-01

    In 1988 a comparative study on important features and models in six major best estimate thermal hydraulic codes for PWR systems was implemented (Comparison of thermal hydraulic safety codes for PWR Graham, Trotman, London, EUR 11522). It was a limitation of that study that the source codes themselves were not available but the comparison had to be based on the available documentation. In the present study, the source codes were available and the capability of four system codes to predict complex two-phase flow phenomena has been assessed. Two areas of investigation were selected: (a) pressurized spray phenomena; (b) boil-up phenomena in rod bundles. As regards the first area, experimental data obtained in 1972 on the Neptunus Facility (Delft University of Technology) were compared with the results of the calculations using Athlet, Cathare, Relap 5 and TRAC-PT1 and, concerning the second area, the results of two experimental facilities obtained in 1980 and 1985 on Thetis (UKEA) and Pericles (CEA-Grenoble) were considered

  4. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  5. SIMMER-II code and its applications

    International Nuclear Information System (INIS)

    Smith, L.L.

    1979-01-01

    The significant features of SIMMER-II, a disrupted-core analysis code, are described. The code has the capabalities to begin space-time neutronics calculations from nonstationary reactor states, to track the intermixing of fuel of different enrichments, and to model the complicated heat- and mass-transfer processes that occur in the transition phase. Example calculations are presented for analyses of whole-core accidents and for analyses of experiments supporting the code models

  6. BEAVRS full core burnup calculation in hot full power condition by RMC code

    International Nuclear Information System (INIS)

    Liu, Shichang; Liang, Jingang; Wu, Qu; Guo, JuanJuan; Huang, Shanfang; Tang, Xiao; Li, Zeguang; Wang, Kan

    2017-01-01

    Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.

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

    Directory of Open Access Journals (Sweden)

    Mohammadnia Meysam

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Zmijarevic, I.; Petrovic, I.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Aggery, A.

    1999-12-01

    The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

  10. Code for calculation of spreading of radioactivity in reactor containment systems

    International Nuclear Information System (INIS)

    Vertes, P.

    1992-09-01

    A detailed description of the new version of TIBSO code is given, with applications for accident analysis in a reactor containment system. The TIBSO code can follow the nuclear transition and the spatial migration of radioactive materials. The modelling of such processes is established in a very flexible way enabling the user to investigate a wide range of problems. The TIBSO code system is described in detail, taking into account the new developments since 1983. Most changes improve the capabilities of the code. The new version of TIBSO system is written in FORTRAN-77 and can be operated both under VAX VMS and PC DOS. (author) 5 refs.; 3 figs.; 21 tabs

  11. A Sample Calculation of Tritium Production and Distribution at VHTR by using TRITGO Code

    International Nuclear Information System (INIS)

    Park, Ik Kyu; Kim, D. H.; Lee, W. J.

    2007-03-01

    TRITGO code was developed for estimating the tritium production and distribution of high temperature gas cooled reactor(HTGR), especially GTMHR350 by General Atomics. In this study, the tritium production and distribution of NHDD was analyzed by using TRITGO Code. The TRITGO code was improved by a simple method to calculate the tritium amount in IS Loop. The improved TRITGO input for the sample calculation was prepared based on GTMHR600 because the NHDD has been designed referring GTMHR600. The GTMHR350 input with related to the tritium distribution was directly used. The calculated tritium activity among the hydrogen produced in IS-Loop is 0.56 Bq/g- H2. This is a very satisfying result considering that the limited tritium activity of Japanese Regulation Guide is 5.6 Bq/g-H2. The basic system to analyze the tritium production and the distribution by using TRITGO was successfully constructed. However, there exists some uncertainties in tritium distribution models, the suggested method for IS-Loop, and the current input was not for NHDD but for GTMHR600. The qualitative analysis for the distribution model and the IS-Loop model and the quantitative analysis for the input should be done in the future

  12. Review and comparison of effective delayed neutron fraction calculation methods with Monte Carlo codes

    International Nuclear Information System (INIS)

    Bécares, V.; Pérez-Martín, S.; Vázquez-Antolín, M.; Villamarín, D.; Martín-Fuertes, F.; González-Romero, E.M.; Merino, I.

    2014-01-01

    Highlights: • Review of several Monte Carlo effective delayed neutron fraction calculation methods. • These methods have been implemented with the Monte Carlo code MCNPX. • They have been benchmarked against against some critical and subcritical systems. • Several nuclear data libraries have been used. - Abstract: The calculation of the effective delayed neutron fraction, β eff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for β eff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of β eff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of β eff

  13. A Sample Calculation of Tritium Production and Distribution at VHTR by using TRITGO Code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ik Kyu; Kim, D. H.; Lee, W. J

    2007-03-15

    TRITGO code was developed for estimating the tritium production and distribution of high temperature gas cooled reactor(HTGR), especially GTMHR350 by General Atomics. In this study, the tritium production and distribution of NHDD was analyzed by using TRITGO Code. The TRITGO code was improved by a simple method to calculate the tritium amount in IS Loop. The improved TRITGO input for the sample calculation was prepared based on GTMHR600 because the NHDD has been designed referring GTMHR600. The GTMHR350 input with related to the tritium distribution was directly used. The calculated tritium activity among the hydrogen produced in IS-Loop is 0.56 Bq/g- H2. This is a very satisfying result considering that the limited tritium activity of Japanese Regulation Guide is 5.6 Bq/g-H2. The basic system to analyze the tritium production and the distribution by using TRITGO was successfully constructed. However, there exists some uncertainties in tritium distribution models, the suggested method for IS-Loop, and the current input was not for NHDD but for GTMHR600. The qualitative analysis for the distribution model and the IS-Loop model and the quantitative analysis for the input should be done in the future.

  14. The neutrons flux density calculations by Monte Carlo code for the double heterogeneity fuel

    International Nuclear Information System (INIS)

    Gurevich, M.I.; Brizgalov, V.I.

    1994-01-01

    This document provides the calculation technique for the fuel elements which consists of the one substance as a matrix and the other substance as the corn embedded in it. This technique can be used in the neutron flux density calculation by the universal Monte Carlo code. The estimation of accuracy is presented too. (authors). 6 refs., 1 fig

  15. Improvements to the nuclear model code GNASH for cross section calculations at higher energies

    International Nuclear Information System (INIS)

    Young, P.G.; Chadwick, M.B.

    1994-01-01

    The nuclear model code GNASH, which in the past has been used predominantly for incident particle energies below 20 MeV, has been modified extensively for calculations at higher energies. The model extensions and improvements are described in this paper, and their significance is illustrated by comparing calculations with experimental data for incident energies up to 160 MeV

  16. Development of EASYQAD version β. A visualization code system for gamma and neutron shielding calculations

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Kim, Soon Young; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Jong Kyung

    2008-01-01

    EASYQAD version β was developed by MATLAB GUI (Graphical User Interface) as a visualization code system based on QAD-CGGP-A point-kernel code for convenient shielding calculations of gammas and neutrons. It consists of four graphic interface modules including GEOMETRY, INPUT, OUTPUT, and SHIELD. These modules were compiled in C++ programming language by using the MATLAB Compiler Toolbox to form a stand-along code system that can be run on the Windows XP operating system without MATLAB installation. In addition, EASYQAD version β has user-friendly graphical interfaces and, additionally, many useful functions in comparison with QAD- CGGP-A such as common material library, line and grid detectors, and multi-group energy calculations so as to increase its applicability in the field of radiation shielding analysis. It is a powerful tool for non-experts to analyze easily the shielding problems without special training. Therefore, EASYOAD version β is expected to contribute effectively to the development of radiation shielding analysis by providing users in medical and industrial fields with an efficient radiation shielding code. (author)

  17. Simulation of the burnup in cell calculation using the WIMSD-5B Code considering different nuclear data libraries

    International Nuclear Information System (INIS)

    Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de

    2017-01-01

    This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)

  18. Simulation of the burnup in cell calculation using the WIMSD-5B Code considering different nuclear data libraries

    Energy Technology Data Exchange (ETDEWEB)

    Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de, E-mail: zelmolima@yahoo.com.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)

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

    Science.gov (United States)

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

    2018-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-01

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

  1. Computer code for double beta decay QRPA based calculations

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-11

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

  2. A Deep Penetration Problem Calculation Using AETIUS:An Easy Modeling Discrete Ordinates Transport Code UsIng Unstructured Tetrahedral Mesh, Shared Memory Parallel

    Science.gov (United States)

    KIM, Jong Woon; LEE, Young-Ouk

    2017-09-01

    As computing power gets better and better, computer codes that use a deterministic method seem to be less useful than those using the Monte Carlo method. In addition, users do not like to think about space, angles, and energy discretization for deterministic codes. However, a deterministic method is still powerful in that we can obtain a solution of the flux throughout the problem, particularly as when particles can barely penetrate, such as in a deep penetration problem with small detection volumes. Recently, a new state-of-the-art discrete-ordinates code, ATTILA, was developed and has been widely used in several applications. ATTILA provides the capabilities to solve geometrically complex 3-D transport problems by using an unstructured tetrahedral mesh. Since 2009, we have been developing our own code by benchmarking ATTILA. AETIUS is a discrete ordinates code that uses an unstructured tetrahedral mesh such as ATTILA. For pre- and post- processing, Gmsh is used to generate an unstructured tetrahedral mesh by importing a CAD file (*.step) and visualizing the calculation results of AETIUS. Using a CAD tool, the geometry can be modeled very easily. In this paper, we describe a brief overview of AETIUS and provide numerical results from both AETIUS and a Monte Carlo code, MCNP5, in a deep penetration problem with small detection volumes. The results demonstrate the effectiveness and efficiency of AETIUS for such calculations.

  3. NULIF: neutron spectrum generator, few-group constant calculator, and fuel depletion code

    International Nuclear Information System (INIS)

    Wittkopf, W.A.; Tilford, J.M.; Andrews, J.B. II; Kirschner, G.; Hassan, N.M.; Colpo, P.N.

    1977-02-01

    The NULIF code generates a microgroup neutron spectrum and calculates spectrum-weighted few-group parameters for use in a spatial diffusion code. A wide variety of fuel cells, non-fuel cells, and fuel lattices, typical of PWR (or BWR) lattices, are treated. A fuel depletion routine and change card capability allow a broad range of problems to be studied. Coefficient variation with fuel burnup, fuel temperature change, moderator temperature change, soluble boron concentration change, burnable poison variation, and control rod insertion are readily obtained. Heterogeneous effects, including resonance shielding and thermal flux depressions, are treated. Coefficients are obtained for one thermal group and up to three epithermal groups. A special output routine writes the few-group coefficient data in specified format on an output tape for automated fitting in the PDQ07-HARMONY system of spatial diffusion-depletion codes

  4. Calculation of the effective delayed neutron fraction by TRIPOLI-4 code for IPEN/MB-01 research reactor

    International Nuclear Information System (INIS)

    Lee, Y.K.; Hugot, F.X.

    2011-01-01

    The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)

  5. New Three-Dimensional Neutron Transport Calculation Capability in STREAM Code

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Youqi [Xi' an Jiaotong University, Xi' an (China); Choi, Sooyoung; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    The method of characteristics (MOC) is one of the best choices for its powerful capability in the geometry modeling. To reduce the large computational burden in 3D MOC, the 2D/1D schemes were proposed and have achieved great success in the past 10 years. However, such methods have some instability problems during the iterations when the neutron leakage for axial direction is large. Therefore, full 3D MOC methods were developed. A lot of efforts have been devoted to reduce the computational costs. However, it still requires too much memory storage and computational time for the practical modeling of a commercial size reactor core. Recently, a new approach for the 3D MOC calculation without transverse integration has been implemented in the STREAM code. In this approach, the angular flux is expressed as a basis function expansion form of only axial variable z. A new approach based on the axial expansion and 2D MOC sweeping to solve the 3D neutron transport equation is implemented in the STREAM code. This approach avoids using the transverse integration in the traditional 2D/1D scheme of MOC calculation. By converting the 3D equation into the 2D form of angular flux expansion coefficients, it also avoids the complex 3D ray tracing. Current numerical tests using two benchmarks show good accuracy of the new method.

  6. Methods and codes for neutronic calculations of the MARIA research reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.; Bretscher, M.M.; Hanan, N.A.; Matos, J.E.

    1998-01-01

    The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6x8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminium. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization. (author)

  7. Calculation of the effective dose from natural radioactivity sources in soil using MCNP code

    International Nuclear Information System (INIS)

    Krstic, D.; Nikezic, D.

    2008-01-01

    Full text: Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this report. Calculations have been done for the most common natural radionuclides in soil as 238 U, 232 Th series and 40 K. A ORNL age-dependent phantom and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs of phantom.The effective dose was calculated according to ICRP74 recommendations. Conversion coefficients of effective dose per air kerma were determined. Results obtained here were compared with other authors

  8. Benchmark evaluation of the RELAP code to calculate boiling in narrow channels

    International Nuclear Information System (INIS)

    Kunze, J.F.; Loyalka, S.K.; McKibben, J.C.; Hultsch, R.; Oladiran, O.

    1990-01-01

    The RELAP code has been tested with benchmark experiments (such as the loss-of-fluid test experiments at the Idaho National Engineering Laboratory) at high pressures and temperatures characteristic of those encountered in loss-of-coolant accidents (LOCAs) in commercial light water power reactors. Application of RELAP to the LOCA analysis of a low pressure (< 7 atm) and low temperature (< 100 degree C), plate-type research reactor, such as the University of Missouri Research Reactor (MURR), the high-flux breeder reactor, high-flux isotope reactor, and Advanced Test Reactor, requires resolution of questions involving overextrapolation to very low pressures and low temperatures, and calculations of the pulsed boiling/reflood conditions in the narrow rectangular cross-section channels (typically 2 mm thick) of the plate fuel elements. The practical concern of this problem is that plate fuel temperatures predicted by RELAP5 (MOD2, version 3) during the pulsed boiling period can reach high enough temperatures to cause plate (clad) weakening, though not melting. Since an experimental benchmark of RELAP under such LOCA conditions is not available and since such conditions present substantial challenges to the code, it is important to verify the code predictions. The comparison of the pulsed boiling experiments with the RELAP calculations involves both visual observations of void fraction versus time and measurements of temperatures near the fuel plate surface

  9. Aerosol behaviour calculations with the code NAUA-Mod5M

    International Nuclear Information System (INIS)

    Bunz, H.; Koyro, M.

    1995-03-01

    This report presents the aerosol behaviour calculations within the framework of SEAFP task A8 'Radioactivity confinement analysis'. The retention capability for the aerosol-type activity of the containment has been evaluated for a number of different accident scenarios with the code NAUA-Mod5M. This code is designed to simulate the aerosol behaviour for an arbitrary multi-compartment containment originally for applications in LWR containments after severe accidents. Altogether six different scenarios have been evaluated, two for the He-cooled RPM and four for the watercooled APM. These scenarios differ mainly in the primary source taken into account, if e.g. the armour of the first wall consists of Be or W or if the divertor cooling loop or a primary cooling loop fails. The results show the positive influence of the system of step by step barriers already proved to be successful for other applications. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Mochamad-Imron

    2003-01-01

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

  12. SPARC-90: A code for calculating fission product capture in suppression pools

    International Nuclear Information System (INIS)

    Owczarski, P.C.; Burk, K.W.

    1991-10-01

    This report describes the technical bases and use of two updated versions of a computer code initially developed to serve as a tool for calculating aerosol particle retention in boiling water reactor (BWR) pressure suppression pools during severe accidents, SPARC-87 and SPARC-90. The most recent version is SPARC-90. The initial or prototype version (Owczarski, Postma, and Schreck 1985) was improved to include the following: rigorous treatment of local particle deposition velocities on the surface of oblate spherical bubbles, new correlations for hydrodynamic behavior of bubble swarms, models for aerosol particle growth, both mechanistic and empirical models for vent exit region scrubbing, specific models for hydrodynamics of bubble breakup at various vent types, and models for capture of vapor iodine species. A complete user's guide is provided for SPARC-90 (along with SPARC-87). A code description, code operating instructions, partial code listing, examples of the use of SPARC-90, and summaries of experimental data comparison studies also support the use of SPARC-90. 29 refs., 4 figs., 11 tabs

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

    International Nuclear Information System (INIS)

    Cintra, Felipe Belonsi de

    2010-01-01

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

  14. French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation-Part II: Cracked plates

    International Nuclear Information System (INIS)

    Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.

    2007-01-01

    French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for the development and evaluation of the accuracy of the proposed analytical methods. This second article in the series presents all details for the stress intensity factor and J calculations for cracked plates. General data applicable for all defect geometries are first presented, and then, available defect geometries where compendia for K I and σ ref calculation are provided are given

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

    Science.gov (United States)

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

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

  18. MRPP: multiregion processing plant code

    International Nuclear Information System (INIS)

    Kee, C.W.; McNeese, L.E.

    1976-09-01

    The report describes the machine solution of a large number (approximately 52,000) of simultaneous linear algebraic equations in which the unknowns are the concentrations of nuclides in the fuel salt of a fluid-fueled reactor (MSBR) having a continuous fuel processing plant. Most of the equations define concentrations at various points in the processing plant. The code allows as input a generalized description of a processing plant flowsheet; it also performs the iterative adjustment of flowsheet parameters for determination of concentrations throughout the flowsheet, and the associated effect of the specified processing mode on the overall reactor operation

  19. Procedures of grasp92 code to calculate accurate Dirac-Coulomb energy for the ground sate of helium atom

    International Nuclear Information System (INIS)

    Utsumi, Takayuki; Sasaki, Akira

    2000-02-01

    The procedures of grasp92 code to calculate accurate (relative error nearly equal 10 -7 ) eigenvalue for the ground sate of helium atom of the Dirac-Coulomb Hamiltonian are presented. The grasp92 code, based on the multi-configuration Dirac-Fock method, is widely used to calculate the atomic properties. However, the main part of the accurate calculations, extended optimal level calculation (EOL), suffer frequently numerical instabilities due to the lack of the confident procedures. The purpose of this report is to illustrate the guideline for stable EOL calculations by calculating the most fundamental atomic system, i.e. the ground sate of helium atom ls 2 1 S 2 . This procedure could be extended for the high-precise eigenfunction calculation of more complex atomic systems, for example highly ionized atoms and high-Z atoms. (author)

  20. A group of neutronics calculations in the MNSR using the MCNP-4C code

    International Nuclear Information System (INIS)

    Khattab, K.; Sulieman, I.

    2009-11-01

    The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)

  1. Improvements of MCOR: A Monte Carlo depletion code system for fuel assembly reference calculations

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, C.; Ivanov, K. [Pennsylvania State Univ., Univ. Park (United States); Misu, S. [AREVA NP GmbH, An AREVA and SIEMENS Company, Erlangen (Germany)

    2006-07-01

    This paper presents the improvements of MCOR, a Monte Carlo depletion code system for fuel assembly reference calculations. The improvements of MCOR were initiated by the cooperation between the Penn State Univ. and AREVA NP to enhance the original Penn State Univ. MCOR version in order to be used as a new Monte Carlo depletion analysis tool. Essentially, a new depletion module using KORIGEN is utilized to replace the existing ORIGEN-S depletion module in MCOR. Furthermore, the online burnup cross section generation by the Monte Carlo calculation is implemented in the improved version instead of using the burnup cross section library pre-generated by a transport code. Other code features have also been added to make the new MCOR version easier to use. This paper, in addition, presents the result comparisons of the original and the improved MCOR versions against CASMO-4 and OCTOPUS. It was observed in the comparisons that there were quite significant improvements of the results in terms of k{sub inf}, fission rate distributions and isotopic contents. (authors)

  2. Calculation of Single Cell and Fuel Assembly IRIS Benchmarks Using WIMSD5B and GNOMER Codes

    International Nuclear Information System (INIS)

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

    2002-01-01

    IRIS reactor (an acronym for International Reactor Innovative and Secure) is a modular, integral, light water cooled, small to medium power (100-335 MWe/module) reactor, which addresses the requirements defined by the United States Department of Energy for Generation IV nuclear energy systems, i.e., proliferation resistance, enhanced safety, improved economics, and waste reduction. An international consortium led by Westinghouse/BNFL was created for development of IRIS reactor; it includes universities, institutes, commercial companies, and utilities. Faculty of Electrical Engineering and Computing, University of Zagreb joined the consortium in year 2001, with the aim to take part in IRIS neutronics design and safety analyses of IRIS transients. A set of neutronic benchmarks for IRIS reactor was defined with the objective to compare results of all participants with exactly the same assumptions. In this paper a calculation of Benchmark 44 for IRIS reactor is described. Benchmark 44 is defined as a core depletion benchmark problem for specified IRIS reactor operating conditions (e.g., temperatures, moderator density) without feedback. Enriched boron, inhomogeneously distributed in axial direction, is used as an integral fuel burnable absorber (IFBA). The aim of this benchmark was to enable a more direct comparison of results of different code systems. Calculations of Benchmark 44 were performed using the modified CORD-2 code package. The CORD-2 code package consists of WIMSD and GNOMER codes. WIMSD is a well-known lattice spectrum calculation code. GNOMER solves the neutron diffusion equation in three-dimensional Cartesian geometry by the Green's function nodal method. The following parameters were obtained in Benchmark 44 analysis: effective multiplication factor as a function of burnup, nuclear peaking factor as a function of burnup, axial offset as a function of burnup, core-average axial power profile, core radial power profile, axial power profile for selected

  3. use of the RESRAD-BUILD code to calculate building surface contamination limits

    International Nuclear Information System (INIS)

    Faillace, E.R.; LePoire, D.; Yu, C.

    1996-01-01

    Surface contamination limits in buildings were calculated for 226 Ra, 230 Th, 232 Th, and natural uranium on the basis of 1 mSv y -1 (100 mrem y -1 ) dose limit. The RESRAD-BUILD computer code was used to calculate these limits for two scenarios: building occupancy and building renovation. RESRAD-BUILD is a pathway analysis model designed to evaluate the potential radiological dose incurred by individuals working or living inside a building contaminated with radioactive material. Six exposure pathways are considered in the RESRAD-BUILD code: (1) external exposure directly from the source; (2) external exposure from materials deposited on the floor; (3) external exposure due to air submersion; (4) inhalation of airborne radioactive particles; (5) inhalation of aerosol indoor radon progeny; and (6) inadvertent ingestion of radioactive material, either directly from the sources or from materials deposited on the surfaces. The code models point, line, area, and volume sources and calculates the effects of radiation shielding, building ventilation, and ingrowth of radioactive decay products. A sensitivity analysis was performed to determine how variations in input parameters would affect the surface contamination limits. In most cases considered, inhalation of airborne radioactive particles was the primary exposure pathway. However, the direct external exposure contribution from surfaces contaminated with 226 Ra was in some cases the dominant pathway for building occupancy depending on the room size, ventilation rates, and surface release fractions. The surface contamination limits are most restrictive for 232 Th, followed by 230 Th, natural uranium, and 226 Ra. The results are compared with the surface contamination limits in the Nuclear Regulatory Commission's Regulatory Guide 1.86, which are most restrictive for 226 Ra and 230 Th, followed by 232 Th, and are least restrictive for natural uranium

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  5. Development of FEMAG. Calculation code of magnetic field generated by ferritic plates in the tokamak devices

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Kazuhiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2003-03-01

    In design of the future fusion devises in which low activation ferritic steel is planned to use as the plasma facing material and/or the inserts for ripple reduction, the appreciation of the error field effect against the plasma as well as the optimization of ferritic plate arrangement to reduce the toroidal field ripple require calculation of magnetic field generated by ferritic steel. However iterative calculations concerning the non-linearity in B-H curve of ferritic steel disturbs high-speed calculation required as the design tool. In the strong toroidal magnetic field that is characteristic in the tokamak fusion devices, fully magnetic saturation of ferritic steel occurs. Hence a distribution of magnetic charges as magnetic field source is determined straightforward and any iteration calculation are unnecessary. Additionally objective ferritic steel geometry is limited to the thin plate and ferritic plates are installed along the toroidal magnetic field. Taking these special conditions into account, high-speed calculation code ''FEMAG'' has been developed. In this report, the formalization of 'FEMAG' code, how to use 'FEMAG', and the validity check of 'FEMAG' in comparison with a 3D FEM code, with the measurements of the magnetic field in JFT-2M are described. The presented examples are numerical results of design studies for JT-60 modification. (author)

  6. Repairing business process models as retrieved from source code

    NARCIS (Netherlands)

    Fernández-Ropero, M.; Reijers, H.A.; Pérez-Castillo, R.; Piattini, M.; Nurcan, S.; Proper, H.A.; Soffer, P.; Krogstie, J.; Schmidt, R.; Halpin, T.; Bider, I.

    2013-01-01

    The static analysis of source code has become a feasible solution to obtain underlying business process models from existing information systems. Due to the fact that not all information can be automatically derived from source code (e.g., consider manual activities), such business process models

  7. Electronic data processing codes for California wildland plants

    Science.gov (United States)

    Merton J. Reed; W. Robert Powell; Bur S. Bal

    1963-01-01

    Systematized codes for plant names are helpful to a wide variety of workers who must record the identity of plants in the field. We have developed such codes for a majority of the vascular plants encountered on California wildlands and have published the codes in pocket size, using photo-reductions of the output from data processing machines. A limited number of the...

  8. About the use of the CATHARE code for best estimate Large Break LOCA calculations: benefits for Safety and constraints

    International Nuclear Information System (INIS)

    Vacher, J.L.

    1994-01-01

    Since 1979, EDF has participated to the development of CATHARE, a best estimate accidental thermalhydraulic code, in collaboration with CEA and FRAMATOME. EDF is now investigating the use of this code for licensing studies and particularly for Large Break LOCA calculations. Until now, the work done at EDF, in relation with FRAMATOME and CEA, has mainly focused on the physical analysis of the transient and on the identification of the key phenomena. This task is a necessary step before uncertainty evaluation. To illustrate this point of view, a peculiar example of calculated Large Break transients for a 900 MW three loop plant is presented. In one of these calculations, a high value of Peak Cladding Temperature was obtained. This peculiar scenario was initiated by a large entrainment of water to the steam generators at the very beginning of the reflooding stage, followed by a strong pressurization which led to a lasting draining of the reactor vessel. The physical phenomena which determine the existence and amplitude of this scenario were identified and their influence was explained: condensation at the accumulator injection, heat exchange in the core, entrainment process to the steam generators. It appeared obvious that the large observed uncertainty was associated to only a few parameters. Although this peculiar system behaviour was obtained for only a particular combination of parameters and a narrow range of thermalhydraulic conditions, the capability of the code to simulate these phenomena was investigated in regard to experimental data. It was concluded that this scenario was definitely unrealistic on a reactor. Nevertheless, this peculiar example tends to demonstrate, firstly, that the use of a best-estimate code improves Safety as it makes possible to point out physical phenomena that could not be considered when using non mechanistic codes, secondly, that the uncertainty evaluation must be guided by a pertinent physical analysis of the transient, focusing

  9. Theoretical calculations of the reaction cross-sections for proton-induced reactions on natural copper using ALICE-IPPE code

    International Nuclear Information System (INIS)

    Alharbi, A.A.; Azzam, A.

    2012-01-01

    A theoretical study of the nuclear-reaction cross sections for proton-induced reactions on 63 Cu and 65 Cu was performed in the proton energy range from threshold values up to 50 MeV. The produced nuclei were different isotopes of Zn, Cu, Ni, Co and Mn, some of which have important applications. The reaction cross-section calculations were performed using the ALICE-IPPE code, which depends on the pre-equilibrium compound nucleus model. This code is suitable for the studied energy and isotopic mass ranges. Approximately 14 excitation functions for the different reactions have been constructed from the calculated cross-section values. The excitation function curves for the proton reactions with natural copper targets have been constructed from those for enriched targets using the natural abundance of the copper isotopes. Comparisons between the calculated excitation functions with those previously experimentally measured are given whenever the experimental values were available. Some statistical parameters were introduced to control the quality of the fitting between both the experimental and the theoretical calculated cross-section values. - Highlights: ► We performed reaction cross section calculations using ALICE-IPPE code. ► We constructed 14 excitation functions for nat Cu(p,xn)Zn,Cu,Ni,Co,Mn reactions. ► The available experimental data were fitted to the performed ALICE-IPPE calculations. ► Statistical parameters were introduced to control the quality of the fitting. ► The code failed to fit the experimental data for reactions with large nucleon emissions.

  10. Uncertainty and sensitivity analysis applied to coupled code calculations for a VVER plant transient

    International Nuclear Information System (INIS)

    Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K. D.

    2004-01-01

    The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics, is an important step to perform best-estimate plant transient calculations. It is generally agreed that the application of best-estimate methods should be supplemented by an uncertainty and sensitivity analysis to quantify the uncertainty of the results. The paper presents results from the application of the GRS uncertainty and sensitivity method for a VVER-440 plant transient, which was already studied earlier for the validation of coupled codes. For this application, the main steps of the uncertainty method are described. Typical results of the method applied to the analysis of the plant transient by several working groups using different coupled codes are presented and discussed The results demonstrate the capability of an uncertainty and sensitivity analysis. (authors)

  11. A 3D coarse-mesh time dependent code for nuclear reactor kinetic calculations

    International Nuclear Information System (INIS)

    Montagnini, B.; Raffaelli, P.; Sumini, M.; Zardini, D.M.

    1996-01-01

    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)

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

    International Nuclear Information System (INIS)

    Behling, S.R.

    1985-05-01

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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  14. Neutronic study of nuclear reactors. Complete calculation of TRIGA MARKII reactor and calculations of fuel temperature coefficients. (Qualification of WIMS code)

    International Nuclear Information System (INIS)

    Benmansour, L.

    1992-01-01

    The present work shows a group of results, obtained by a neutronic study, concerning the TRIGA MARK II reactor and LIGHT WATER reactors. These studies aim to make cell and diffusion calculations. WIMS D-4 with extended library and DIXY programs are used and tested for those purposes. We also have proceeded to a qualification of WIMS code based on the fuel temperature coefficient calculations. 33 refs.; 23 figs.; 30 tabs. (author)

  15. FOOD II: an interactive code for calculating concentrations of radionuclides in food products

    International Nuclear Information System (INIS)

    Zach, R.

    1978-11-01

    An interactive code, FOOD II, has been written in FORTRAN IV for the PDP 10 to allow calculation of concentrations of radionuclides in food products and internal doses to man under chronic release conditions. FOOD II uses models unchanged from a previous code, FOOD, developed at Battelle, Pacific Northwest Laboratories. The new code has different input and output features than FOOD and a number of options have been added to increase flexibility. Data files have also been updated. FOOD II takes into account contamination of vegetation by air and irrigation water containing radionuclides. Contamination can occur simultaneously by air and water. Both direct deposition of radionuclides on leaves, and their uptake from soil are possible. Also, animals may be contaminated by ingestion of vegetation and drinking water containing radionuclides. At present, FOOD II provides selection of 14 food types, 13 diets and numerous radionuclides. Provisions have been made to expand all of these categories. Six additional contaminated food products can also be entered directly into the dose model. Doses may be calculated for the total body and six internal organs. Summaries of concentrations in food products and internal doses to man can be displayed at a local terminal or at an auxiliary high-speed printer. (author)

  16. SCATLAW: a code of scattering law and cross sections calculation for liquids and solids

    International Nuclear Information System (INIS)

    Padureanu, I.; Rapeanu, S.; Rotarascu, G.; Craciun, C.

    1978-11-01

    A code for calculation of the scattering law S(Q,ω), differential and double differential cross sections and scattering kernels in the energy range E(0 - 683 meV) and wave-vector transfer Q(0 - 40 A -1 ) is presented. The code can be used both for solids and liquids which are coherent or incoherent scatterer. For liquids the calculations are based on the most recent theoretical models involving the correlation functions and generalized field approach. The phonon expansion model and the free gas model are also analysed in term of frequency spectra obtained from inelastic neutron scattering using time-of-flight technique. Several results on liquid sodium at T = 233 deg C and on liquid bismuth at T = 286 deg C and T = 402 deg C are presented. (author)

  17. MIF-SCD computer code for thermal hydraulic calculation of supercritical water cooled reactor core

    International Nuclear Information System (INIS)

    Galina P Bogoslovskaia; Alexander A Karpenko; Pavel L Kirillov; Alexander P Sorokin

    2005-01-01

    Full text of publication follows: Supercritical pressure power plants constitute the basis of heat power engineering in many countries to day. Starting from a long-standing experience of their operation, it is proposed to develop a new type of fast breeder reactor cooled by supercritical water, which enables the economical indices of NPP to be substantially improved. In the Thermophysical Department of SSC RF-IPPE, an attempt is made to provide thermal-hydraulic validation of the reactor under discussion. The paper presents the results of analysis of the thermal-hydraulic characteristics of fuel subassemblies cooled by supercritical water based on subchannel analysis. Modification of subchannel code MIF - MIF-SCD Code - developed in the SSC RF IPPE is designed as block code and permits one to calculate the coolant temperature and velocity distributions in fuel subassembly channels, the temperature of fuel pin claddings and fuel subassembly wrapper under conditions of irregular geometry and non-uniform axial and radial power generation. The thermal hydraulics under supercritical pressure of water exhibits such peculiarities as abrupt variation of the thermal physical properties in the range of pseudo-critical temperature, the absence of such phenomenon as the critical heat flux which can lead to fuel element burnout in WWERs. As compared with subchannel code for light water, in order to take account of the variation of the coolant properties versus temperature in more detail, a block for evaluating the thermal physical properties of supercritical water versus the local coolant temperature in the fuel subassembly channels was added. The peculiarities of the geometry and power generation in the fuel subassembly of the supercritical reactor are considered as well in special blocks. The results of calculations have shown that considerable preheating of supercritical coolant (several hundreds degrees) can occur in the fuel subassembly. The test calculations according to

  18. FINEDAN - an explicit finite-element calculation code for two-dimensional analyses of fast dynamic transients in nuclear reactor technology

    International Nuclear Information System (INIS)

    Adamik, V.; Matejovic, P.

    1989-01-01

    The problems are discussed of nonstationary, nonlinear dynamics of the continuum. A survey is presented of calculation methods in the given area with emphasis on the area of impact problems. A description is presented of the explicit finite elements method and its application to two-dimensional Cartesian and cylindrical configurations. Using the method the explicit calculation code FINEDAN was written which was tested in a series of verification calculations for different configurations and different types of continuum. The main characteristics are presented of the code and of some, of its practical applications. Envisaged trends of the development of the code and its possible applications in the technology of nuclear reactors are given. (author). 9 figs., 4 tabs., 10 refs

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

    International Nuclear Information System (INIS)

    Hardie, R.W.

    1982-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

  1. Estimation of small perturbation effects in multiversion calculations by the PRIZMA-D code

    International Nuclear Information System (INIS)

    Kandiev, Ya.Z.; Malakhov, A.A.; Serova, E.V.; Spirina, S.G.

    2005-01-01

    The PRIZMA-D code is intended for solving by the Monte Carlo method of the problems, connected with calculations of nuclear reactors and critical assemblies. Taking into account the effect of the perturbation on the distribution of the source division points is carried out by means of the method of small iterations for the division points. This method is described in the paper. Possibilities of its application are shown by the examples of calculations of some problems. The comparative results are presented [ru

  2. Study of multiplication factor sensitivity to the spread of WWER spent fuel isotopics calculated by different codes

    International Nuclear Information System (INIS)

    Markova, L.

    2001-01-01

    As a sensitivity study the impact on the system reactivity was studied in the case that different calculational methodologies of spent fuel isotopic concentrations were used for WWER spent fuel inventory computations. The sets of isotopic concentrations obtained by calculations with different codes and libraries as a result of the CB2 international benchmark focused on WWER-440 burnup credit were used to show the spread of the calculated spent fuel system reactivity. Using the MCNP 4B code and changing the isotopics input data, the multiplication factor of an infinite array of the WWER-440 fuel pin cells was calculated. The evaluation of the results shows the sensitivity of the calculated reactivity to different calculational methodologies used for the spent fuel inventory computation. In the studied cases of the CB2 benchmark, the spread of the reference k-results relative to the mean was found less or about ±1% in spite of the fact that the data of isotopic concentrations were spread much more. (author)

  3. BetaShape: A new code for improved analytical calculations of beta spectra

    Directory of Open Access Journals (Sweden)

    Mougeot Xavier

    2017-01-01

    Full Text Available The new code BetaShape has been developed in order to improve the nuclear data related to beta decays. An analytical model was considered, except for the relativistic electron wave functions, for ensuring fast calculations. Output quantities are mean energies, log ft values and beta and neutrino spectra for single and multiple transitions. The uncertainties from the input parameters, read from an ENSDF file, are propagated. A database of experimental shape factors is included. A comparison over the entire ENSDF database with the standard code currently used in nuclear data evaluations shows consistent results for the vast majority of the transitions and highlights the improvements that can be expected with the use of BetaShape.

  4. User effects on the thermal-hydraulic transient system code calculations

    International Nuclear Information System (INIS)

    Aksan, S.N.; D'Auria, F.; Staedtke, H.

    1993-01-01

    In the 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 that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects will be discussed in detail and general recommendations and conclusions will be presented to control and limit them. (orig.)

  5. Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent codes

    International Nuclear Information System (INIS)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

    2014-01-01

    The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

  6. Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent codes

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)

    2014-07-01

    The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

  7. Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, A.K.

    1978-02-01

    Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation.

  8. Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)

    International Nuclear Information System (INIS)

    Agrawal, A.K.

    1978-02-01

    Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation

  9. Code ACTIVE for calculation of the transmutation, induced activity and decay heat in neutron irradiation

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Harada, Yuhei; Asami, Naoto.

    1976-03-01

    The computer code ACTIVE has been prepared for calculation of the transmutation rate, induced activity and decay heat. Calculations are carried out with activation chain and spatial distribution of neutron energy spectrum. The spatial distribution of secondary gamma-ray source due to the unstable nuclides is also obtainable. Special attension is paid to the short life decays. (auth.)

  10. Accelerated radiotherapy planners calculated by parallelization with GPUs

    International Nuclear Information System (INIS)

    Reinado, D.; Cozar, J.; Alonso, S.; Chinillach, N.; Cortina, T.; Ricos, B.; Diez, S.

    2011-01-01

    In this paper we have developed and tested by a subroutine parallelization architectures graphics processing units (GPUs) to apply to calculations with standard algorithms known code. The experience acquired during these tests shall also apply to the MC calculations in radiotherapy if you have the code.

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

    International Nuclear Information System (INIS)

    Hugot, F.X.; Lee, Y.K.; Malvagi, F.

    2008-01-01

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

  12. Nuclide Inventory Calculation Using MCNPX for Wolsung Unit 1 Reactor Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie; Noh, Kyoung Ho; Hah, Chang Joo [KEPCO International Nuclear Graduate School, Daejeon (Korea, Republic of)

    2014-05-15

    The CINDER90 computation process involves utilizing linear Markovian chains to determine the time dependent nuclide densities. The CINDER90 depletion algorithm is implemented the MCNPX code package. The coupled depletion process involves a Monte-Carlo steady-state reaction rate calculation linked to a deterministic depletion calculation. The process is shown in Fig.1. MCNPX runs a steady state calculation to determine the system eigenvalue collision densities, recoverable energies from fission and neutrons per fission events. In order to generate number densities for the next time step, the CINDER90 code takes the MCNPX generated values and performs a depletion calculation. MCNPX then takes the new number densities and caries out a new steady-stated calculation. The process repeats itself until the final time step. This paper describe the preliminary source term and nuclide inventory calculation of Candu single fuel channel using MCNPX, as a part of the activities to support the equilibrium core model development and decommissioning evaluation process of a Candu reactor. The aim of this study was to apply the MCNPX code for source term and nuclide inventory calculation of Candu single fuel channel. Nuclide inventories as a function of burnup will be used to model an equilibrium core for Candu reactor. The core lifetime neutron fluence obtained from the model is used to estimate radioactivity at the stage of decommisioning. In general, as expected, the actinides and fission products build up increase with increasing burnup. Despite the fact that the MCNPX code is still in development we can conclude that the code is capable of obtaining relevant results in burnup and source term calculation. It is recommended that in the future work, the calculation has to be verified on the basis of experimental data or comparison with other codes.

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

    International Nuclear Information System (INIS)

    Suzuki, Motoye; Kawasaki, Satoru; Furuta, Teruo

    1978-06-01

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

  14. User's manual for EXALPHA (a code for calculating electronic properties of molecules). [Muscatel code, multiply scattered electron approximation

    Energy Technology Data Exchange (ETDEWEB)

    Jones, H.D.

    1976-06-01

    The EXALPHA procedures provide a simplified method for running the MUSCATEL computer code, which in turn is used for calculating electronic properties of simple molecules and atomic clusters, based on the multiply scattered electron approximation for the wave equations. The use of the EXALPHA procedures to set up a run of MUSCATEL is described.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  16. Improvement and test calculation on basic code or sodium-water reaction jet

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Yoshinori; Itooka, Satoshi [Advanced Reactor Engineering Center, Hitachi Works, Hitachi Ltd., Hitachi, Ibaraki (Japan); Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo [Consulting Engineering Dept., Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan)

    1999-03-01

    In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3{center_dot}Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

  17. Improvement and test calculation on basic code or sodium-water reaction jet

    International Nuclear Information System (INIS)

    Saito, Yoshinori; Itooka, Satoshi; Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo

    1999-03-01

    In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

  18. Calculations for the intermediate-spectrum cells of Zebra 8 using the MONK Monte-Carlo Code

    International Nuclear Information System (INIS)

    Hanlon, D.; Franklin, B.M.; Stevenson, J.M.

    1987-10-01

    The Monte-Carlo Code MONK 6A and its associated point-energy cross-section data have been used to analyse seven, zero-leakage, plate-geometry cells from the ZEBRA 8 assemblies. The convergence of the calculations was such that the uncertainties in k-infinity and the more important reaction-rate ratios were generally less than the experimental uncertainties. The MONK 6A predictions have been compared with experiment and with predictions from the MURAL collision-probability code. This uses FGL5 data which has been adjusted on the basis of ZEBRA 8 and other integral experiments. The poor predictions from the MONK calculations with errors of up to 10% in k-infinity, are attributed to deficiencies in the database for intermediate to fast spectrum systems. (author)

  19. Calculation of age-dependent effective doses for external exposure using the MCNP code

    International Nuclear Information System (INIS)

    Hung, Tran Van

    2013-01-01

    Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV. (orig.)

  20. Calculation of age-dependent effective doses for external exposure using the MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Hung, Tran Van [Research and Development Center for Radiation Technology, ThuDuc, HoChiMinh City (VT)

    2013-07-15

    Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV. (orig.)

  1. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

    International Nuclear Information System (INIS)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

    2013-10-01

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  2. Current status of high energy nucleon-meson transport code

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)

  3. French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation-Part I: General overview

    International Nuclear Information System (INIS)

    Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.

    2007-01-01

    Two French nuclear codes include flaw assessment procedures: the RSE-M code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of five parts: this first one presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components: plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for their development and on evaluation of the accuracy of the proposed analytical methods. This first article of the series presents an overview of the calculation of K I and J in these two codes and describes briefly the defect assessment analyses. Specific details in the Appendix A16 of RCC-MR (LBB procedure and creep analyses) are also introduced in this article

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

    International Nuclear Information System (INIS)

    Mill, A.J.

    1979-08-01

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

  5. Test of Effective Solid Angle code for the efficiency calculation of volume source

    Energy Technology Data Exchange (ETDEWEB)

    Kang, M. Y.; Kim, J. H.; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of); Sun, G. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is hard to determine a full energy (FE) absorption peak efficiency curve for an arbitrary volume source by experiment. That's why the simulation and semi-empirical methods have been preferred so far, and many works have progressed in various ways. Moens et al. determined the concept of effective solid angle by considering an attenuation effect of γ-rays in source, media and detector. This concept is based on a semi-empirical method. An Effective Solid Angle code (ESA code) has been developed for years by the Applied Nuclear Physics Group in Seoul National University. ESA code converts an experimental FE efficiency curve determined by using a standard point source to that for a volume source. To test the performance of ESA Code, we measured the point standard sources and voluminous certified reference material (CRM) sources of γ-ray, and compared with efficiency curves obtained in this study. 200∼1500 KeV energy region is fitted well. NIST X-ray mass attenuation coefficient data is used currently to check for the effect of linear attenuation only. We will use the interaction cross-section data obtained from XCOM code to check the each contributing factor like photoelectric effect, incoherent scattering and coherent scattering in the future. In order to minimize the calculation time and code simplification, optimization of algorithm is needed.

  6. Description of the CAREM Reactor Neutronic Calculation Codes; Descripcion de la Linea de Calculo Neutronica del CAREM

    Energy Technology Data Exchange (ETDEWEB)

    Villarino, Eduardo; Hergenreder, Daniel [Investigacion Aplicada, INVAP, San Carlos de Bariloche (Argentina)

    2000-07-01

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

  7. Improvements in practical applicability of NSHEX: nodal transport calculation code for three-dimensional hexagonal-Z geometry

    International Nuclear Information System (INIS)

    Sugino, Kazuteru

    1998-07-01

    As a tool to perform a fast reactor core calculations with high accuracy, NSHEX the nodal transport calculation code for three-dimensional hexagonal-Z geometry is under development. To improve the practical applicability of NSHEX, for instance, in its application to safety analysis and commercial reactor core design studies, we investigated the basic theory used in it, improved the program performance, and evaluated its applicability to the analysis of commercial reactor cores. The current studies show the following: (1) An improvement in the treatment of radial leakage in the radial nodal coupling equation bettered calculational convergence for safety analysis calculation, so the applicability of NSHEX to safety analysis was improved. (2) As a result of comparison of results from NSHEX and the standard core calculation code, it was confirmed that there was consistency between them. (3) According to the evaluation of the effect due to the difference of calculational condition, it was found that the calculation under appropriate nodal expansion orders and Sn orders correspond to the one under most detailed condition. However further investigation is required to reduce the uncertainty in calculational results due to the treatment of high order flux moments. (4) A whole core version of NSHEX enabling calculation for any FBR core geometry has been developed, this improved general applicability of NSHEX. (5) An investigation of the applicability of the rebalance method to acceleration clarified that this improved calculational convergence and it was effective. (J.P.N.)

  8. Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

    Directory of Open Access Journals (Sweden)

    Ahmed Amin El Said Abd El Hameed

    2015-08-01

    Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code.  We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon

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

    International Nuclear Information System (INIS)

    Young, P.G.; Chadwick, M.B.

    1994-01-01

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

  10. Validation of dose calculation programmes for recycling

    International Nuclear Information System (INIS)

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

    2002-12-01

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

  11. Validation of dose calculation programmes for recycling

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  12. Applying the universal neutron transport codes to the calculation of well-logging probe response at different rock porosities

    International Nuclear Information System (INIS)

    Bogacz, J.; Loskiewicz, J.; Zazula, J.M.

    1991-01-01

    The use of universal neutron transport codes in order to calculate the parameters of well-logging probes presents a new approach first tried in U.S.A. and UK in the eighties. This paper deals with first such an attempt in Poland. The work is based on the use of MORSE code developed in Oak Ridge National Laboratory in U.S.A.. Using CG MORSE code we calculated neutron detector response when surrounded with sandstone of porosities 19% and 38%. During the work it come out that it was necessary to investigate different methods of estimation of the neutron flux. The stochastic estimation method as used currently in the original MORSE code (next collision approximation) can not be used because of slow convergence of its variance. Using the analog type of estimation (calculation of the sum of track lengths inside detector) we obtained results of acceptable variance (∼ 20%) for source-detector spacing smaller than 40 cm. The influence of porosity on detector response is correctly described for detector positioned at 27 cm from the source. At the moment the variances are quite large. (author). 33 refs, 8 figs, 8 tabs

  13. Research on calculation of mixing fraction for natural uranium equivalent fuel

    International Nuclear Information System (INIS)

    Huang Shien; Wang Lianjie; Wei Yanqin; Li Qing; Zheng Jiye

    2013-01-01

    Based on the first-order perturbation theory and reasonable approximations, the calculation method of recycled uranium (RU) and depleted uranium (DU) mixing fraction for natural uranium equivalent (NUE) fuel was studied, so the equivalence between NUE fuel and natural uranium (NU) fuel was assured. The adopted calculation method accurately takes the variation of micro cross sections alone with fuel depletion into account. A computer code named ALPHA was programmed to execute the calculation procedure. Then the ALPHA code and the WIMS-AECL code compose a processing system, which is applicable to the mixing fraction calculation for heavy water reactor NUE fuel. The validation shows that the processing system can accurately calculate the mixing fraction for NUE fuel. (authors)

  14. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for PKA energy spectra and heating number under neutron irradiation

    International Nuclear Information System (INIS)

    Iwamoto, Y.; Ogawa, T.

    2016-01-01

    The modelling of the damage in materials irradiated by neutrons is needed for understanding the mechanism of radiation damage in fission and fusion reactor facilities. The molecular dynamics simulations of damage cascades with full atomic interactions require information about the energy distribution of the Primary Knock on Atoms (PKAs). The most common process to calculate PKA energy spectra under low-energy neutron irradiation is to use the nuclear data processing code NJOY2012. It calculates group-to-group recoil cross section matrices using nuclear data libraries in ENDF data format, which is energy and angular recoil distributions for many reactions. After the NJOY2012 process, SPKA6C is employed to produce PKA energy spectra combining recoil cross section matrices with an incident neutron energy spectrum. However, intercomparison with different processes and nuclear data libraries has not been studied yet. Especially, the higher energy (~5 MeV) of the incident neutrons, compared to fission, leads to many reaction channels, which produces a complex distribution of PKAs in energy and type. Recently, we have developed the event generator mode (EGM) in the Particle and Heavy Ion Transport code System PHITS for neutron incident reactions in the energy region below 20 MeV. The main feature of EGM is to produce PKA with keeping energy and momentum conservation in a reaction. It is used for event-by-event analysis in application fields such as soft error analysis in semiconductors, micro dosimetry in human body, and estimation of Displacement per Atoms (DPA) value in metals and so on. The purpose of this work is to specify differences of PKA spectra and heating number related with kerma between different calculation method using PHITS-EGM and NJOY2012+SPKA6C with different libraries TENDL-2015, ENDF/B-VII.1 and JENDL-4.0 for fusion relevant materials

  15. Comparative calculations with the HETC/MCNP and HETC/TWODAN codes

    International Nuclear Information System (INIS)

    Broeders, C.; Broeders, I.

    1995-01-01

    Transmutations of actinides and fission products can be achieved also by proton accelerators. For a theoretical study of this process the HETC code has been developed. A special procedure has been developed for dealing with spallation neutrons whose kinetic energy is below 10 to 20 MeV. (orig.)

  16. Calculations of the giant-dipole-resonance photoneutrons using a coupled EGS4-morse code

    International Nuclear Information System (INIS)

    Liu, J.C.; Nelson, W.R.; Kase, K.R.; Mao, X.S.

    1995-10-01

    The production and transport of the photoneutrons from the giant-dipoleresonance reaction have been implemented in a coupled EGS4-MORSE code. The total neutron yield (including both the direct neutron and evaporation neutron components) is calculated by folding the photoneutron yield cross sections with the photon track length distribution in the target. Empirical algorithms based on the measurements have been developed to estimate the fraction and energy of the direct neutron component for each photon. The statistical theory in the EVAP4 code, incorporated as a MORSE subroutine, is used to determine the energies of the evaporation neutrons. These represent major improvements over other calculations that assumed no direct neutrons, a constant fraction of direct neutrons, monoenergetic direct neutron, or a constant nuclear temperature for the evaporation neutrons. It was also assumed that the slow neutrons ( 2 θ, which have a peak emission at 900. Comparisons between the calculated and the measured photoneutron results (spectra of the direct, evaporation and total neutrons; nuclear temperatures; direct neutron fractions) for materials of lead, tungsten, tantalum and copper have been made. The results show that the empirical algorithms, albeit simple, can produce reasonable results over the interested photon energy range

  17. ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

    1976-07-01

    The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

  18. Development of SCINFUL-CG code to calculate response functions of scintillators in various shapes used for neutron measurement

    Energy Technology Data Exchange (ETDEWEB)

    Endo, Akira; Kim, Eunjoo; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-10-01

    A Monte Carlo code SCINFUL has been utilized for calculating response functions of organic scintillators for high-energy neutron spectroscopy. However, the applicability of SCINFUL is limited to the calculations for cylindrical NE213 and NE110 scintillators. In the present study, SCINFUL-CG was developed by introducing a geometry specifying function and high-energy neutron cross section data into SCINFUL. The geometry package MARS-CG, the extended version of the CG (Combinatorial Geometry), was programmed into SCINFUL-CG to express various geometries of detectors. Neutron spectra in the regions specified by the CG can be evaluated by the track length estimator. The cross section data of silicon, oxygen and aluminum for neutron transport calculation were incorporated up to 100 MeV using the data of LA150 library. Validity of SCINFUL-CG was examined by comparing calculated results with those by SCINFUL and MCNP and experimental data measured using high-energy neutron fields. SCINFUL-CG can be used for the calculations of the response functions and neutron spectra in the organic scintillators in various shapes. The computer code will be applicable to the designs of high-energy neutron spectrometers and neutron monitors using the organic scintillators. The present report describes the new features of SCINFUL-CG and explains how to use the code. (author)

  19. Validation of VHTRC calculation benchmark of critical experiment using the MCB code

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2016-01-01

    Full Text Available The calculation benchmark problem Very High Temperature Reactor Critical (VHTR a pin-in-block type core critical assembly has been investigated with the Monte Carlo Burnup (MCB code in order to validate the latest version of Nuclear Data Library based on ENDF format. Executed benchmark has been made on the basis of VHTR benchmark available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments. This benchmark is useful for verifying the discrepancies in keff values between various libraries and experimental values. This allows to improve accuracy of the neutron transport calculations that may help in designing the high performance commercial VHTRs. Almost all safety parameters depend on the accuracy of neutron transport calculation results that, in turn depend on the accuracy of nuclear data libraries. Thus, evaluation of the libraries applicability to VHTR modelling is one of the important subjects. We compared the numerical experiment results with experimental measurements using two versions of available nuclear data (ENDF-B-VII.1 and JEFF-3.2 prepared for required temperatures. Calculations have been performed with the MCB code which allows to obtain very precise representation of complex VHTR geometry, including the double heterogeneity of a fuel element. In this paper, together with impact of nuclear data, we discuss also the impact of different lattice modelling inside the fuel pins. The discrepancies of keff have been successfully observed and show good agreement with each other and with the experimental data within the 1 σ range of the experimental uncertainty. Because some propagated discrepancies observed, we proposed appropriate corrections in experimental constants which can improve the reactivity coefficient dependency. Obtained results confirm the accuracy of the new Nuclear Data Libraries.

  20. Calculs de doses générées par les rayonnements ionisants principes physiques et codes de calcul

    CERN Document Server

    Vivier, Alain

    2016-01-01

    Cet ouvrage et les codes associés s’adressent aux utilisateurs de sources de rayonnements ionisants : techniciens, ingénieurs de sécurité, personnes compétentes en radioprotection, mais aussi médecins, chercheurs, concepteurs, décideurs… Les contraintes croissantes liées à la radioprotection rendent indispensables l’utilisation de codes de calcul permettant d’évaluer les débits de doses générées par ces sources et la façon dont on peut s’en protéger au mieux. De nombreux codes existent, dont certains restent des références incontournables, mais ils sont relativement complexes à mettre en oeuvre et restent en général réservés aux bureaux d’études. En outre, ces codes sont souvent des « boîtes noires » qui ne permettent pas de comprendre la physique sous-jacente. L’objectif de cet ouvrage est double : - Exposer les principes physiques permettant de comprendre les phénomènes à l’oeuvre lorsque la matière est irradiée par des rayonnements ionisants. Il devient al...

  1. Running codes through the web

    International Nuclear Information System (INIS)

    Clark, R.E.H.

    2001-01-01

    Dr. Clark presented a report and demonstration of running atomic physics codes through the WWW. The atomic physics data is generated from Los Alamos National Laboratory (LANL) codes that calculate electron impact excitation, ionization, photoionization, and autoionization, and inversed processes through detailed balance. Samples of Web interfaces, input and output are given in the report

  2. Development of DUST: A computer code that calculates release rates from a LLW disposal unit

    International Nuclear Information System (INIS)

    Sullivan, T.M.

    1992-01-01

    Performance assessment of a Low-Level Waste (LLW) disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the disposal unit source term). The major physical processes that influence the source term are water flow, container degradation, waste form leaching, and radionuclide transport. A computer code, DUST (Disposal Unit Source Term) has been developed which incorporates these processes in a unified manner. The DUST code improves upon existing codes as it has the capability to model multiple container failure times, multiple waste form release properties, and radionuclide specific transport properties. Verification studies performed on the code are discussed

  3. Contribution of the Nea data bank in the field of calculation codes in radiation protection, radio physics and dosimetry; Role de la banque de donnees de l'AEN dans le domaine des codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    Kodeli, I; Sartori, E [Organization for Econimic Co-Operation and Development (OECD NEA DB), 91 - Issy les Moulineaux (France)

    2003-07-01

    The Nuclear energy agency is a specialised agency of OECD (organization economic co-operation and development). These missions are to help its members to keep and improve by international cooperation, the scientific, technological and legal bases necessary to a peaceful use of nuclear energy. Nea includes twenty eight countries. Nea works in collaboration with IAEA. The field of activities concerns the acquisition, validation and distribution of nuclear data, calculation codes and experiments. To help users, it organises conferences and training about the calculation codes that it shares out. (N.C.)

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

    International Nuclear Information System (INIS)

    Keney, G.S.

    1981-08-01

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

  5. Summary of ENDF/B pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1981-12-01

    This document contains the summary documentation for the ENDF/B pre-processing codes: LINEAR, RECENT, SIGMA1, GROUPIE, EVALPLOT, MERGER, DICTION, CONVERT. This summary documentation is merely a copy of the comment cards that appear at the beginning of each programme; these comment cards always reflect the latest status of input options, etc. For the latest published documentation on the methods used in these codes see UCRL-50400, Vol.17 parts A-E, Lawrence Livermore Laboratory (1979)

  6. Packing simulation code to calculate distribution function of hard spheres by Monte Carlo method : MCRDF

    International Nuclear Information System (INIS)

    Murata, Isao; Mori, Takamasa; Nakagawa, Masayuki; Shirai, Hiroshi.

    1996-03-01

    High Temperature Gas-cooled Reactors (HTGRs) employ spherical fuels named coated fuel particles (CFPs) consisting of a microsphere of low enriched UO 2 with coating layers in order to prevent FP release. There exist many spherical fuels distributed randomly in the cores. Therefore, the nuclear design of HTGRs is generally performed on the basis of the multigroup approximation using a diffusion code, S N transport code or group-wise Monte Carlo code. This report summarizes a Monte Carlo hard sphere packing simulation code to simulate the packing of equal hard spheres and evaluate the necessary probability distribution of them, which is used for the application of the new Monte Carlo calculation method developed to treat randomly distributed spherical fuels with the continuous energy Monte Carlo method. By using this code, obtained are the various statistical values, namely Radial Distribution Function (RDF), Nearest Neighbor Distribution (NND), 2-dimensional RDF and so on, for random packing as well as ordered close packing of FCC and BCC. (author)

  7. A benchmark test of computer codes for calculating average resonance parameters

    International Nuclear Information System (INIS)

    Ribon, P.; Thompson, A.

    1983-01-01

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

  8. Calculations of multi-particle processes at the one-loop level: precise predictions for the LHC

    Energy Technology Data Exchange (ETDEWEB)

    Karg, Stefan

    2007-07-01

    We present a calculation of the loop-induced processes gg{yields}HH and gg{yields}HHH, and investigate the observability of multi-Higgs boson production at the LHC in the Standard Model and beyond. While the SM cross sections are too small to allow observation at the LHC, we demonstrate that physics beyond the SM can lead to amplified, observable cross sections. Furthermore, the applicability of the heavy top quark approximation in two- and three-Higgs boson production is investigated. The calculation of the process PP{yields}VV+jet at NLO is an important background process to Higgs production in association with a jet at the LHC. We compute the virtual corrections to this process which form the ''bottleneck'' for obtaining a complete NLO prediction. The resulting analytic expressions are generated with highly automated computer routines and translated into a flexible Fortran code, which can be employed in the computation of differential cross sections of phenomenological interest. (orig.)

  9. Calculations of multi-particle processes at the one-loop level: precise predictions for the LHC

    International Nuclear Information System (INIS)

    Karg, Stefan

    2007-01-01

    We present a calculation of the loop-induced processes gg→HH and gg→HHH, and investigate the observability of multi-Higgs boson production at the LHC in the Standard Model and beyond. While the SM cross sections are too small to allow observation at the LHC, we demonstrate that physics beyond the SM can lead to amplified, observable cross sections. Furthermore, the applicability of the heavy top quark approximation in two- and three-Higgs boson production is investigated. The calculation of the process PP→VV+jet at NLO is an important background process to Higgs production in association with a jet at the LHC. We compute the virtual corrections to this process which form the ''bottleneck'' for obtaining a complete NLO prediction. The resulting analytic expressions are generated with highly automated computer routines and translated into a flexible Fortran code, which can be employed in the computation of differential cross sections of phenomenological interest. (orig.)

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

    Directory of Open Access Journals (Sweden)

    Skrzypek Maciej

    2015-09-01

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

  11. The use of the SRIM code for calculation of radiation damage induced by neutrons

    Science.gov (United States)

    Mohammadi, A.; Hamidi, S.; Asadabad, Mohsen Asadi

    2017-12-01

    Materials subjected to neutron irradiation will being evolve to structural changes by the displacement cascades initiated by nuclear reaction. This study discusses a methodology to compute primary knock-on atoms or PKAs information that lead to radiation damage. A program AMTRACK has been developed for assessing of the PKAs information. This software determines the specifications of recoil atoms (using PTRAC card of MCNPX code) and also the kinematics of interactions. The deterministic method was used for verification of the results of (MCNPX+AMTRACK). The SRIM (formely TRIM) code is capable to compute neutron radiation damage. The PKAs information was extracted by AMTRACK program, which can be used as an input of SRIM codes for systematic analysis of primary radiation damage. Then the Bushehr Nuclear Power Plant (BNPP) radiation damage on reactor pressure vessel is calculated.

  12. Procedure and code for calculating black control rods taking into account epithermal absorption, code CAS-1; Postupak i program za proracun crnih kontrolnih sipki, uzimajuci u obzir i epitermalnu apsorpciju, CAS-1

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Trivunac, N; Zivkovic, Z [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1964-12-15

    This report describes the computer code CAS-1, calculation method and procedure applied for calculating the black control rods taking into account the epithermal neutron absorption. Results obtained for supercell method applied for regular lattice reflected in the multiplication medium is part of this report in addition to the computer code manual.

  13. Code package for calculation of damage effects of medium-energy protons in metal targets

    International Nuclear Information System (INIS)

    Coulter, C.A.

    1976-12-01

    A program package was developed to calculate radiation damage effects produced in a metal target by protons in the 100-MeV to 3.5-GeV energy range. A detailed description is given of the control cards and data cards required to use the code package

  14. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

    International Nuclear Information System (INIS)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-01-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  15. The status of simulation codes for extraction process using mixer-settler

    Energy Technology Data Exchange (ETDEWEB)

    Byeon, Kee Hoh; Lee, Eil Hee; Kwon, Seong Gil; Kim, Kwang Wook; Yang, Han Beom; Chung, Dong Yong; Lim, Jae Kwan; Shin, Hyun Kyoo; Kim, Soo Ho

    1999-10-01

    We have studied and analyzed the mixer-settler simulation codes such as three kinds of SEPHIS series, PUBG, and EXTRA.M, which is the most recently developed code. All of these are sufficiently satisfactory codes in the fields of process/device modeling, but it is necessary to formulate the accurate distribution data and chemical reaction mechanism for the aspect of accuracy and reliability. In the aspect of application to be the group separation process, the mixer-settler model of these codes have no problems, but the accumulation and formulation of partitioning and reaction equilibrium data of chemical elements used in group separation process is very important. (author)

  16. Spread-out Bragg peak and monitor units calculation with the Monte Carlo Code MCNPX

    International Nuclear Information System (INIS)

    Herault, J.; Iborra, N.; Serrano, B.; Chauvel, P.

    2007-01-01

    The aim of this work was to study the dosimetric potential of the Monte Carlo code MCNPX applied to the protontherapy field. For series of clinical configurations a comparison between simulated and experimental data was carried out, using the proton beam line of the MEDICYC isochronous cyclotron installed in the Centre Antoine Lacassagne in Nice. The dosimetric quantities tested were depth-dose distributions, output factors, and monitor units. For each parameter, the simulation reproduced accurately the experiment, which attests the quality of the choices made both in the geometrical description and in the physics parameters for beam definition. These encouraging results enable us today to consider a simplification of quality control measurements in the future. Monitor Units calculation is planned to be carried out with preestablished Monte Carlo simulation data. The measurement, which was until now our main patient dose calibration system, will be progressively replaced by computation based on the MCNPX code. This determination of Monitor Units will be controlled by an independent semi-empirical calculation

  17. Experimental validation of decay heat calculation codes and associated nuclear data libraries for fusion energy

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Wada, Masayuki; Ikeda, Yujiro

    2001-01-01

    Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within ±10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the 92 Mo(n, 2n) 91g Mo reaction in FENDL, and lack of activation cross section data, e.g., the 138 Ba(n, 2n) 137m Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)

  18. Experimental validation of decay heat calculation codes and associated nuclear data libraries for fusion energy

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-01-01

    Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within {+-}10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the {sup 92}Mo(n, 2n){sup 91g}Mo reaction in FENDL, and lack of activation cross section data, e.g., the {sup 138}Ba(n, 2n){sup 137m}Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)

  19. Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

    2017-03-15

    In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  20. Verification of calculational models of heat exchange crisis and overcrisis heat transfer, used in the KORSAR code

    International Nuclear Information System (INIS)

    Bezrukov, Yu.A.; Shchekoldin, V.I.

    2002-01-01

    On the basis of the Gidropress OKB (Special Design Bureau) experimental data bank one verified the KORSAR code design models and correlations as to heat exchange crisis and overcrisis heat transfer as applied to the WWER reactor normal and emergency conditions. The VI.006.000 version of KORSAR code base calculations is shown to describe adequately the conducted experiments and to deviate insignificantly towards the conservative approach. So it may be considered as one of the codes ensuring more precise estimation [ru

  1. Validation of the VTT's reactor physics code system

    International Nuclear Information System (INIS)

    Tanskanen, A.

    1998-01-01

    At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)

  2. Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system

    International Nuclear Information System (INIS)

    Arigane, Kenji

    1987-04-01

    The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)

  3. The enhanced variance propagation code for the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Kern, E.A.; Zack, N.R.; Britschgi, J.J.

    1992-01-01

    The Variance Propagation (VP) Code was developed by the Los Alamos National Laboratory's Safeguard's Systems Group to provide off-line variance propagation and systems analysis for nuclear material processing facilities. The code can also be used as a tool in the design and evaluation of material accounting systems. In this regard , the VP code was enhanced to incorporate a model of the material accountability measurements used in the Idaho Chemical Processing Plant operated by the Westinghouse Idaho Nuclear Company. Inputs to the code were structured to account for the dissolves/headend process, the waste streams, process performed to determine the sensitivity of measurement and sampling errors to the overall material balance error. We determined that the material balance error is very sensitive to changes in the sampling errors. 3 refs

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

    International Nuclear Information System (INIS)

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

    1976-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-06-01

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

  6. Advanced burnup calculation code system in a subcritical state with continuous-energy Monte Carlo code for fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Matsunaka, Masayuki; Ohta, Masayuki; Miyamaru, Hiroyuki; Murata, Isao

    2009-01-01

    The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before. (author)

  7. Analysis of quantum error-correcting codes: Symplectic lattice codes and toric codes

    Science.gov (United States)

    Harrington, James William

    Quantum information theory is concerned with identifying how quantum mechanical resources (such as entangled quantum states) can be utilized for a number of information processing tasks, including data storage, computation, communication, and cryptography. Efficient quantum algorithms and protocols have been developed for performing some tasks (e.g. , factoring large numbers, securely communicating over a public channel, and simulating quantum mechanical systems) that appear to be very difficult with just classical resources. In addition to identifying the separation between classical and quantum computational power, much of the theoretical focus in this field over the last decade has been concerned with finding novel ways of encoding quantum information that are robust against errors, which is an important step toward building practical quantum information processing devices. In this thesis I present some results on the quantum error-correcting properties of oscillator codes (also described as symplectic lattice codes) and toric codes. Any harmonic oscillator system (such as a mode of light) can be encoded with quantum information via symplectic lattice codes that are robust against shifts in the system's continuous quantum variables. I show the existence of lattice codes whose achievable rates match the one-shot coherent information over the Gaussian quantum channel. Also, I construct a family of symplectic self-dual lattices and search for optimal encodings of quantum information distributed between several oscillators. Toric codes provide encodings of quantum information into two-dimensional spin lattices that are robust against local clusters of errors and which require only local quantum operations for error correction. Numerical simulations of this system under various error models provide a calculation of the accuracy threshold for quantum memory using toric codes, which can be related to phase transitions in certain condensed matter models. I also present

  8. Calculation of the spallation product distribution in the evaporation process

    International Nuclear Information System (INIS)

    Nishida, T.; Kanno, I.; Nakahara, Y.; Takada, H.

    1989-01-01

    Some investigations are performed for the calculational model of nuclear spallation reaction in the evaporation process. A new version of a spallation reaction simulation code NUCLEUS has been developed by incorporating the newly revised Uno ampersand Yamada's mass formula and extending the counting region of produced nuclei. The differences between the new and original mass formulas are shown in the comparisons of mass excess values. The distributions of spallation products of a uranium target nucleus bombarded by energy (0.38 - 2.9 GeV) protons have been calculated with the new and original versions of NUCLEUS. In the fission component Uno ampersand Yamada's mass formula reproduces the measured data obtained from thin foil experiments significantly better, especially in the neutron excess side, than the combination of the Cameron's mass formula and the mass table compiled by Wapstra, et al., in the original version of NUCLEUS. Discussions are also made on how the mass-yield distribution of products varies dependent on the level density parameter a characterizing the particle evaporation. 16 refs., 7 figs., 1 tab

  9. Calculation of the spallation product distribution in the evaporation process

    International Nuclear Information System (INIS)

    Nishida, T.; Kanno, I.; Nakahara, Y.; Takada, H.

    1989-01-01

    Some investigations are performed for the calculational model of nuclear spallation reaction in the evaporation process. A new version of a spallation reaction simulation code NUCLEUS has been developed by incorporating the newly revised Uno and Yamada's mass formula and extending the counting region of produced nuclei. The differences between the new and original mass formulas are shown in the comparisons of mass excess values. The distributions of spallation products of a uranium target nucleus bombarded by energy (0.38 - 2.9 GeV) protons have been calculated with the new and original versions of NUCLEUS. In the fission component Uno and Yamada's mass formula reproduces the measured data obtained from thin foil experiments significantly better, especially in the neutron excess side, than the combination of the Cameron's mass formula and the mass table compiled by Wapstra, et al., in the original version of NUCLEUS. Discussions are also made on how the mass-yield distribution of products varies dependent on the level density parameter α characterizing the particle evaporation. (author)

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  11. Comparison of calculated integral values using measured and calculated neutron spectra for fusion neutronics analyses

    International Nuclear Information System (INIS)

    Sekimoto, H.

    1987-01-01

    The kerma heat production density, tritum production density, and dose in a lithium-fluoride pile with a deuterium-tritum neutron source were calculated with a data processing code, UFO, from the pulse height distribution of a miniature NE213 neutron spectrometer, and compared with the values calculated with a Monte Carlo code, MORSE-CV. Both the UFO and MORSE-CV values agreed with the statistical error (less than 6%) of the MORSE-CV calculations, except for the outer-most point in the pile. The MORSE-CV values were slightly smaller than the UFO values for almost all cases, and this tendency increased with increasing distance from the neutron source

  12. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko

    2017-04-01

    Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.

  13. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Klippel, H.T.; Hogenbirk, A.; Oppe, J.; Sciolla, C.M.; Stad, R.C.L. van der; Zhang, B.C.

    1997-06-01

    As part of the activities within the framework of the development of INCOGEN, a 'Dutch' conceptual design of a smaller HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRs, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (orig.)

  14. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Hass, J.B.M. De; Klippel, H.Th.; Hogenbirk, A.; Oppe, J.; Sciolla, C.; Stad, R.C.L. Van Der; Zhang, B.C.

    1997-01-01

    As part of the activities within the framework of the development of INCOGEN, a ''Dutch'' conceptual design of a small HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRS, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (author)

  15. Calculation of fuel and moderator temperature coefficients in APR1400 nuclear reactor by MVP code

    International Nuclear Information System (INIS)

    Pham Tuan Nam; Le Thi Thu; Nguyen Huu Tiep; Tran Viet Phu

    2014-01-01

    In this project, these fuel and moderator temperature coefficients were calculated in APR1400 nuclear reactor by MVP code. APR1400 is an advanced water pressurized reactor, that was researched and developed by Korea Experts, its electric power is 1400 MW. The neutronics calculations of full core is very important to analysis and assess a reactor. Results of these calculation is input data for thermal-hydraulics calculations, such as fuel and moderator temperature coefficients. These factors describe the self-safety characteristics of nuclear reactor. After obtaining these reactivity parameters, they were used to re-run the thermal hydraulics calculations in LOCA and RIA accidents. These thermal-hydraulics results were used to analysis effects of reactor physics parameters to thermal hydraulics situation in nuclear reactors. (author)

  16. NRSC, Neutron Resonance Spectrum Calculation System

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2004-01-01

    1 - Description of program or function: The NRSC system is a package of four programs for calculating detailed neutron spectra and related quantities, for homogeneous mixtures of isotopes and cylindrical reactor pin cells, in the energy resonance region, using ENDF/B evaluated nuclear data pre-processed with NJOY or Cullen's codes up to the Doppler Broadening and unresolved resonance level. 2 - Methods: NRSC consists of four programs: GEXSCO, RMET21, ALAMBDA and WLUTIL. GEXSCO prepares the nuclear data from ENDF/B evaluated nuclear data pre-processed with NJOY or Cullen's codes up to the Doppler Broadening or unresolved resonance level for RMET21 input. RMET21 calculates spectra and related quantities for homogeneous mixtures of isotopes and cylindrical reactor pin cells, in the energy resonance region, using slowing-down algorithms and, in the case of pin cells, the collision probability method. ALAMBDA obtains lambda factors (Goldstein-Cohen intermediate resonance factors in the formalism of WIMSD code) of different isotopes for including on WIMSD-type multigroup libraries for WIMSD or other cell-codes, from output of RMET21 program. WLUTIL is an auxiliary program for extracting tabulated parameters related with RMET21 program calculations from WIMSD libraries for comparisons, and for producing new WIMSD libraries with parameters calculated with RMET21 and ALAMBDA programs. 3 - Restrictions on the complexity of the problem: GEXSCO program has fixed array dimensions that are suitable for processing all reasonable outputs from nuclear data pre-processing programs. RMET21 program uses variable dimension method from a fixed general array. ALAMBDA and WLUTIL programs have fixed arrays that are adapted to standard WIMSD libraries. All programs can be easily modified to adapt to special requirements

  17. Development of thermal-hydraulic safety codes for HTGRs with gas-turbine and hydrogen process cycles

    International Nuclear Information System (INIS)

    No, Hee Cheon; Yoon, Ho Joon; Lee, Byung Jin; Kim, Yong Soo; Jin, Hyeng Gon; Kim, Ji Hwan; Kim, Hyeun Min; Lim, Hong Sik

    2008-01-01

    We present three nuclear/hydrogen-related R and D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA in which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed SANA code to predict the characteristics of HTGR helium turbines based on the throughflow calculation with a Newton-Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found out that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The GAMMA-SANA coupled code was assessed by comparing its results with the steady-state of the GTHTR300, and the load reduction transient was simulated for the 100% to 70% power operation. The calculation results confirm that two-dimensional throughflow modeling can be successfully used to describe the gas turbine behavior. The dynamic equations for the distillation column of the HI process in the I-S cycle are described with 4 material components involved in the HI process: H2O, HI, I2, and H2. For the VLE prediction in the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. Relative to the experimental data, the improved Neumann model shows deviations of 8.6% in maximum and 2.5% in average for the total pressure, and 9.5% in maximum for the liquid-liquid separation composition. Through a parametric analysis using the published experimental data related to the Bunsen reaction and liquid-liquid separation, an optimized operating condition for the

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

    International Nuclear Information System (INIS)

    Notari, Carla; Grant, Carlos R.

    2000-01-01

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

  19. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1999-01-01

    The present lecture has a main goal to show how the transport lattice calculations are realised in a standard computer code. This is illustrated on the example of the WIMSD code, belonging to the most popular tools for reactor calculations. Most of the approaches discussed here can be easily modified to any other lattice code. The description of the code assumes the basic knowledge of reactor lattice, on the level given in the lecture on 'Reactor lattice transport calculations'. For more advanced explanation of the WIMSD code the reader is directed to the detailed descriptions of the code cited in References. The discussion of the methods and models included in the code is followed by the generally used homogenisation procedure and several numerical examples of discrepancies in calculated multiplication factors based on different sources of library data. (author)

  20. Relative efficiency calculation of a HPGe detector using MCNPX code

    International Nuclear Information System (INIS)

    Medeiros, Marcos P.C.; Rebello, Wilson F.; Lopes, Jose M.; Silva, Ademir X.

    2015-01-01

    High-purity germanium detectors (HPGe) are mandatory tools for spectrometry because of their excellent energy resolution. The efficiency of such detectors, quoted in the list of specifications by the manufacturer, frequently refers to the relative full-energy peak efficiency, related to the absolute full-energy peak efficiency of a 7.6 cm x 7.6 cm (diameter x height) NaI(Tl) crystal, based on the 1.33 MeV peak of a 60 Co source positioned 25 cm from the detector. In this study, we used MCNPX code to simulate a HPGe detector (Canberra GC3020), from Real-Time Neutrongraphy Laboratory of UFRJ, to survey the spectrum of a 60 Co source located 25 cm from the detector in order to calculate and confirm the efficiency declared by the manufacturer. Agreement between experimental and simulated data was achieved. The model under development will be used for calculating and comparison purposes with the detector calibration curve from software Genie2000™, also serving as a reference for future studies. (author)

  1. User's manual of MANYCASK code for calculation of spatial distributions of radiation dose rates in a system composed of many spent-fuel-shipping casks

    International Nuclear Information System (INIS)

    Yamakoshi, Hisao

    1986-01-01

    A calculation code MANYCASK is designed for evaluation of spatial distributions of radiation dose rates in ships loaded with a lot of spent fuel shipping casks. Principle of the calculation method adopted in this code is different from that of ordinary codes, and is advantageous for calculating highly reliable dose rate distributions with a very short calculation time. Basic concept of the principle has been described in other reports in detail. A brief description of the principle will be included in the present report along with a technique named Shadow Technique in this report, in addition to format descriptions of output data as well as input data. Results of sample calculations are compared with measured results in figures so as to show how the calculation method adopted is valid. For the purpose of making this code popular among many people, the author writes the user's manual in the present report in Japanese for domestic users, and in English in another report for people in abroad. (author)

  2. Criticality calculation method for mixer-settlers

    International Nuclear Information System (INIS)

    Gonda, Kozo; Aoyagi, Haruki; Nakano, Ko; Kamikawa, Hiroshi.

    1980-01-01

    A new criticality calculation code MACPEX has been developed to evaluate and manage the criticality of the process in the extractor of mixer-settler type. MACPEX can perform the combined calculation with the PUREX process calculation code MIXSET, to get the neutron flux and the effective multiplication constant in the mixer-settlers. MACPEX solves one-dimensional diffusion equation by the explicit difference method and the standard source-iteration technique. The characteristics of MACPEX are as follows. 1) Group constants of 4 energy groups for the 239 Pu-H 2 O solution, water, polyethylene and SUS 28 are provided. 2) The group constants of the 239 Pu-H 2 O solution are given by the functional formulae of the plutonium concentration, which is less than 50 g/l. 3) Two boundary conditions of the vacuum condition and the reflective condition are available in this code. 4) The geometrical bucklings can be calculated for a certain energy group and/or region by using the three dimentional neutron flux profiles obtained by CITATION. 5) The buckling correction search can be carried out in order to get a desired k sub(eff). (author)

  3. KENOREST - A new coupled code system based on KENO and OREST for criticality and burnup inventory calculations

    International Nuclear Information System (INIS)

    Hesse, U.; Gmal, B.; Voggenberger, Th.; Baleanu, M.; Langenbuch, S.

    2001-01-01

    The program system KENOREST version 1998 will be presented, which is a useful tool for burnup and reactivity calculations for LWR fuel. The three-dimensional Monte Carlo code KENO-V.a is coupled with the one-dimensional GRS burnup program system OREST-98. The objective is to achieve a better modelling of plutonium and actinide build-up or burnout for advanced heterogeneous fuel assembly designs. Further objectives are directed to reliable calculations of the pin power distributions and of reactor safety parameters including axial and radial rod temperatures for fuel assemblies of modern design. The stand-alone-code KENO-V.a version is used without any changes in the program source. The OREST-98 system was developed to handle multirod problems and additional burnup dependent moderator conditions which can be applied to stretch-out simulations in the reactor. A new interface module RESPEFF between KENO and OREST transforms the 2-d or 3-d KENO flux results to the one-dimensional lattice code OREST in a fully automated manner to maintain reaction rate balance between the codes. First results for assembly multiplication factors, isotope inventories are compared with OECD results. (author)

  4. Utilization of the Nelkin model in a Hammer computer code for calculation the reactor parameters

    International Nuclear Information System (INIS)

    Leal, L.C.

    1980-07-01

    The possibility of modifying the HAMMER code, in the thermal part, by changing the thermal neutron scattering Kernel of its library for another one calculated in a subprogramm which can be incorporated to the code, is studied. This subprogramm uses the original version of the Nelkin model instead of its approximation which is used in the HAMMER. It has the advantage of giving the values of the Kernel for any temperature of the reactor for the approximations P 0 , P 1 , P 2 and P 3 . (Author) [pt

  5. Finite volume thermal-hydraulics and neutronics coupled calculations - 15300

    International Nuclear Information System (INIS)

    Araujo Silva, V.; Campagnole dos Santos, A.A.; Mesquit, A.Z.; Bernal, A.; Miro, R.; Verdu, G.; Pereira, C.

    2015-01-01

    The computational power available nowadays allows the coupling of neutronics and thermal-hydraulics codes for reactor studies. The present methodology foresees at least one constraint to the separated codes in order to perform coupled calculations: both codes must use the same geometry, however, meshes can be different for each code as long as the internal surfaces stays the same. Using the finite volume technique, a 3D diffusion nodal code was implemented to deal with neutron transport. This code can handle non-structured meshes which allows for complicated geometries calculations and therefore more flexibility. A computational fluid dynamics (CFD) code was used in order to obtain the same level of details for the thermal hydraulics calculations. The chosen code is OpenFOAM, an open-source CFD tool. Changes in OpenFOAM allow simple coupled calculations of a PWR fuel rod with neutron transport code. OpenFOAM sends coolant density information and fuel temperature to the neutron transport code that sends back power information. A mapping function is used to average values when one node in one side corresponds to many nodes in the other side. Data is exchanged between codes by library calls. As the results of a fuel rod calculations progress, more complicated and processing demanding geometries will be simulated, aiming to the simulation of a real scale PWR fuel assembly

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

    International Nuclear Information System (INIS)

    Daverio, Hernando J.

    2003-01-01

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

  7. Composition calculations by the KARATE code system for the spent-fuel samples from the Novovoronezh reactor

    International Nuclear Information System (INIS)

    Hordosy, G.

    2006-01-01

    KARATE is a code system developed in KFKI AERI. It is routinely used for core calculation. Its depletion module are now tested against the radiochemical measurements of spent fuel samples from the Novovoronezh Unit IV, performed in RIAR, Dimitrovgrad. Due to the insufficient knowledge of operational history of the unit, the irradiation history of the samples was taken from formerly published Russian calculations. The calculation of isotopic composition was performed by the MULTICEL module of program system. The agreement between the calculated and measured values of the concentration of the most important actinides and fission products is investigated (Authors)

  8. Software quality and process improvement in scientific simulation codes

    Energy Technology Data Exchange (ETDEWEB)

    Ambrosiano, J.; Webster, R. [Los Alamos National Lab., NM (United States)

    1997-11-01

    This report contains viewgraphs on the quest to develope better simulation code quality through process modeling and improvement. This study is based on the experience of the authors and interviews with ten subjects chosen from simulation code development teams at LANL. This study is descriptive rather than scientific.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-10-01

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

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

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki

    1999-10-01

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

  11. Feasibility study on embedded transport core calculations

    International Nuclear Information System (INIS)

    Ivanov, B.; Zikatanov, L.; Ivanov, K.

    2007-01-01

    The main objective of this study is to develop an advanced core calculation methodology based on embedded diffusion and transport calculations. The scheme proposed in this work is based on embedded diffusion or SP 3 pin-by-pin local fuel assembly calculation within the framework of the Nodal Expansion Method (NEM) diffusion core calculation. The SP 3 method has gained popularity in the last 10 years as an advanced method for neutronics calculation. NEM is a multi-group nodal diffusion code developed, maintained and continuously improved at the Pennsylvania State University. The developed calculation scheme is a non-linear iteration process, which involves cross-section homogenization, on-line discontinuity factors generation, and boundary conditions evaluation by the global solution passed to the local calculation. In order to accomplish the local calculation, a new code has been developed based on the Finite Elements Method (FEM), which is capable of performing both diffusion and SP 3 calculations. The new code will be used in the framework of the NEM code in order to perform embedded pin-by-pin diffusion and SP 3 calculations on fuel assembly basis. The development of the diffusion and SP 3 FEM code is presented first following by its application to several problems. Description of the proposed embedded scheme is provided next as well as the obtained preliminary results of the C3 MOX benchmark. The results from the embedded calculations are compared with direct pin-by-pin whole core calculations in terms of accuracy and efficiency followed by conclusions made about the feasibility of the proposed embedded approach. (authors)

  12. Conversion coefficients for individual monitoring calculated with integrated tiger series, ITS3, Monte Carlo code

    International Nuclear Information System (INIS)

    Devine, R.T.; Hsu, Hsiao-Hua

    1994-01-01

    The current basis for conversion coefficients for calibrating individual photon dosimeters in terms of dose equivalents is found in the series of papers by Grosswent. In his calculation the collision kerma inside the phantom is determined by calculation of the energy fluence at the point of interest and the use of the mass energy absorption coefficient. This approximates the local absorbed dose. Other Monte Carlo methods can be sued to provide calculations of the conversion coefficients. Rogers has calculated fluence-to-dose equivalent conversion factors with the Electron-Gamma Shower Version 3, EGS3, Monte Carlo program and produced results similar to Grosswent's calculations. This paper will report on calculations using the Integrated TIGER Series Version 3, ITS3, code to calculate the conversion coefficients in ICRU Tissue and in PMMA. A complete description of the input parameters to the program is given and comparison to previous results is included

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

    International Nuclear Information System (INIS)

    Shepherd, I.M.

    1982-01-01

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

  14. Coded Ultrasound for Blood Flow Estimation Using Subband Processing

    DEFF Research Database (Denmark)

    Gran, Fredrik; Udesen, Jesper; Nielsen, Michael Bachamnn

    2008-01-01

    the excitation signal is broadband and has good spatial resolution after pulse compression. This means that time can be saved by using the same data for B-mode imaging and blood flow estimation. Two different coding schemes are used in this paper, Barker codes and Golay codes. The performance of the codes......This paper investigates the use of coded excitation for blood flow estimation in medical ultrasound. Traditional autocorrelation estimators use narrow-band excitation signals to provide sufficient signal-to-noise-ratio (SNR) and velocity estimation performance. In this paper, broadband coded...... signals are used to increase SNR, followed by subband processing. The received broadband signal is filtered using a set of narrow-band filters. Estimating the velocity in each of the bands and averaging the results yields better performance compared with what would be possible when transmitting a narrow...

  15. The local skin dose conversion coefficients of electrons, protons and alpha particles calculated using the Geant4 code.

    Science.gov (United States)

    Zhang, Bintuan; Dang, Bingrong; Wang, Zhuanzi; Wei, Wei; Li, Wenjian

    2013-10-01

    The skin tissue-equivalent slab reported in the International Commission on Radiological Protection (ICRP) Publication 116 to calculate the localised skin dose conversion coefficients (LSDCCs) was adopted into the Monte Carlo transport code Geant4. The Geant4 code was then utilised for computation of LSDCCs due to a circular parallel beam of monoenergetic electrons, protons and alpha particles electrons and alpha particles are found to be in good agreement with the results using the MCNPX code of ICRP 116 data. The present work thus validates the LSDCC values for both electrons and alpha particles using the Geant4 code.

  16. One-, two- and three-dimensional transport codes using multi-group double-differential form cross sections

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.

    1988-11-01

    We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)

  17. Computation of the Genetic Code

    Science.gov (United States)

    Kozlov, Nicolay N.; Kozlova, Olga N.

    2018-03-01

    One of the problems in the development of mathematical theory of the genetic code (summary is presented in [1], the detailed -to [2]) is the problem of the calculation of the genetic code. Similar problems in the world is unknown and could be delivered only in the 21st century. One approach to solving this problem is devoted to this work. For the first time provides a detailed description of the method of calculation of the genetic code, the idea of which was first published earlier [3]), and the choice of one of the most important sets for the calculation was based on an article [4]. Such a set of amino acid corresponds to a complete set of representations of the plurality of overlapping triple gene belonging to the same DNA strand. A separate issue was the initial point, triggering an iterative search process all codes submitted by the initial data. Mathematical analysis has shown that the said set contains some ambiguities, which have been founded because of our proposed compressed representation of the set. As a result, the developed method of calculation was limited to the two main stages of research, where the first stage only the of the area were used in the calculations. The proposed approach will significantly reduce the amount of computations at each step in this complex discrete structure.

  18. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

  19. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

  20. Adaptation of penelope Monte Carlo code system to the absorbed dose metrology: characterization of high energy photon beams and calculations of reference dosimeter correction factors

    International Nuclear Information System (INIS)

    Mazurier, J.

    1999-01-01

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

  1. Calculation of conversion coefficients for clinical photon spectra using the MCNP code.

    Science.gov (United States)

    Lima, M A F; Silva, A X; Crispim, V R

    2004-01-01

    In this work, the MCNP4B code has been employed to calculate conversion coefficients from air kerma to the ambient dose equivalent, H*(10)/Ka, for monoenergetic photon energies from 10 keV to 50 MeV, assuming the kerma approximation. Also estimated are the H*(10)/Ka for photon beams produced by linear accelerators, such as Clinac-4 and Clinac-2500, after transmission through primary barriers of radiotherapy treatment rooms. The results for the conversion coefficients for monoenergetic photon energies, with statistical uncertainty <2%, are compared with those in ICRP publication 74 and good agreements were obtained. The conversion coefficients calculated for real clinic spectra transmitted through walls of concrete of 1, 1.5 and 2 m thick, are in the range of 1.06-1.12 Sv Gy(-1).

  2. Civil engineering: calculations of pre-stressed concrete structures using CodeAster

    International Nuclear Information System (INIS)

    Gerard, B.; Ulm, F.

    1997-11-01

    This document presents an analysis of the different calculation methods for pre-stressed concrete structure which can be performed by using finite element methods. Two methods of calculating the pre-stressing of concrete structures with finite elements have been determined. The equivalent method which consists of replacing the action of pre-stressing the concrete by equivalent forces. These method is well suited to dimensioning and studying the overall stability of a structure. It is not an easy matter to take into account the coupled or time-varying phenomena. This approach ignores the evolution of the interaction between the pre-stressing and the concrete. The explicit method which consists of including the mechanical resolution of the pre-stressed cables in that of a concrete structure. Not only does this allow a local study of the pre-stressed to be made, it also allows the coupling which developed over time to be determined, e.g. slip, deferred deformation and coupling between the steel and concrete behaviours. This method enables non-linear phenomena with varying degrees of complexity, such as fracture or yielding of the steels, drying out of the concrete, creep, etc to be described. The two methods are complementary. This document presents the mathematical and computer developments relating to each of this method. In the case of the explicit method, certain of the Code-Aster functions already make it possible to meet several EDF application requirements. Several couplings can be taken into account, such as thermomechanical, shrinkage in drying, creep, relaxation and injection of the cables. Three immediate developments of Code-Aster are proposed for the following applications: - a procedure for calculating the pre-stress losses along the pre-stressing cables; - a command to allocate these forces in the form of an initial force field in the bar elements associated with the cables; - a procedure for linking elements whose nodes do not coincide with each other

  3. Transient electromagnetic analysis in tokamaks using TYPHOON code

    International Nuclear Information System (INIS)

    Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.

    1996-01-01

    The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)

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

    International Nuclear Information System (INIS)

    Kim, Yong Ho

    1991-02-01

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

  5. Calculation of local flow conditions in the lower core of a PWR with code-Saturne

    International Nuclear Information System (INIS)

    Fournier, Y.

    2003-01-01

    In order to better understand the stresses to which fuel rods are subjected, we need to improve our knowledge of the fluid flow inside the core. A code specialized for calculations in tube bundles is used to calculate the flow inside the whole of the core, with a resolution at the assembly level. Still, it is necessary to obtain realistic entry conditions, and these depend on the flow in the downcomer and lower plenum. Also, the flow in the first stages of the core features 4 incoming jets per assembly, and requires a resolution much finer than that used for the whole core calculation. A series of calculations are thus run with our incompressible Navier-Stokes solver, Code-Saturne, using a classical Ranse turbulence model. The first calculations involve a detailed geometry, including part of the cold legs, downcomer, lower plenum, and lower core of a pressurized water reactor. The level of detail includes most obstacles below the core. The lower core plate, being pierced with close to 800 holes, cannot be realistically represented within a practical mesh size, so that a head loss model is used. The lower core itself requiring even more detail is also represented with head losses. We make full use of Code-Saturne's non conforming mesh possibilities to represent a complex geometry, being careful to retain a good mesh quality. Starting just under the lower core, the mesh is aligned with fuel rod assemblies, so that different types of assemblies can be represented through different head loss coefficients. These calculations yield steady-state or near steady-state results, which are compared to experimental data, and should be sufficient to yield realistic entry conditions for full core calculations at assembly width resolution, and beyond those mechanical strain calculations. We are also interested in more detailed flow conditions and fluctuations in the lower core area, so as to better quantify vibrational input. This requires a much higher resolution, which we limit

  6. Automatic calculations of electroweak processes

    International Nuclear Information System (INIS)

    Ishikawa, T.; Kawabata, S.; Kurihara, Y.; Shimizu, Y.; Kaneko, T.; Kato, K.; Tanaka, H.

    1996-01-01

    GRACE system is an excellent tool for calculating the cross section and for generating event of the elementary process automatically. However it is not always easy for beginners to use. An interactive version of GRACE is being developed so as to be a user friendly system. Since it works exactly in the same environment as PAW, all functions of PAW are available for handling any histogram information produced by GRACE. As its application the cross sections of all elementary processes with up to 5-body final states induced by e + e - interaction are going to be calculated and to be summarized as a catalogue. (author)

  7. Comparison calculation of a large sodium-cooled fast breeder reactor using the cell code MICROX-2 in connection with ENDF/B-VI and JEF-1.1 neutron data

    International Nuclear Information System (INIS)

    Pelloni, S.

    1992-02-01

    We have obtained results for a large sodium-cooled fast breeder reactor benchmark using data from the ENDF/B-VI and from Revision 1 of the JEF-1 (JEF-1.1) evaluation. The required cross sections were processed with the NJOY code system (Version 89.62) and homogenized with the spectrum cell code MICROX-2. Multigroup transport-theory calculations in 33 neutron groups (forward and adjoint) were performed using the two-dimensional code TWODANT and kinetic parameters were determined using the first-order perturbation-theory code PERT-V. We calculated eigenvalues, neutron balance data, global and regional breeding and conversion ratios, central rate ratios and reactivity worths with and without sodium, effective delayed neutron fraction and inhour reactivity, regional sodium void reactivity, and isothermal core fuel Doppler-reactivities. In particular, it is shown that good agreement (generally within one standard deviation) is achieved between these results and the average values over sixteen benchmark solutions obtained in the past. The eigenvalues predicted with ENDF/B-VI are up to 0.7% larger than those calculated with JEF-1.1 cross sections. This discrepancy is mainly due to different inelastic scattering cross sections for 23 Na and 238 U, and to different fast fission and nubar data for 239 Pu. (author) 5 figs., 30 tabs., 24 refs

  8. Development of process simulation code for reprocessing plant and process analysis for solvent degradation and solvent washing waste

    International Nuclear Information System (INIS)

    Tsukada, Tsuyoshi; Takahashi, Keiki

    1999-01-01

    We developed a process simulation code for an entire nuclear fuel reprocessing plant. The code can be used on a PC. Almost all of the equipment in the reprocessing plant is included in the code and the mass balance model of each item of equipment is based on the distribution factors of flow-out streams. All models are connected between the outlet flow and the inlet flow according to the process flow sheet. We estimated the amount of DBP from TBP degradation in the entire process by using the developed code. Most of the DBP is generated in the Pu refining process by the effect of α radiation from Pu, which is extracted in a solvent. On the other hand, very little of DBP is generated in the U refining process. We therefore propose simplification of the solvent washing process and volume reduction of the alkali washing waste in the U refining process. The first Japanese commercial reprocessing plant is currently under construction at Rokkasho Mura, Recently, for the sake of process simplification, the original process design has been changed. Using our code, we analyzed the original process and the simplified process. According our results, the volume of alkali waste solution in the low-level liquid treatment process will be reduced by half in the simplified process. (author)

  9. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  10. Evaluation of dose calculation models for inhabited areas applicable in nuclear accident consequence assessment codes

    International Nuclear Information System (INIS)

    Katalin Eged; Zoltan Kis; Natalia Semioschkina; Gabriele Voigt

    2004-01-01

    One of the objectives of the EC project EVANET-TERRA is to provide suitable inputs to the RODOS system. This study gives an overview on urban dose calculation models with special emphasis on the RECLAIM-EDEM2M and TEMAS-urban codes. The TEMAS-urban code is more complex compared to the RECLAIM-EDEM2M code although both models use similar and some times even same model parameters. The database and the way of its data collection as used in RECLAIM-EDEM2M is recommended as a preferred option because it contains many data from local and regional measurements. However in a decision situation the outputs of the TEMASurban model may better help stake holders by providing a ranking of the surfaces to be decontaminated. (author)

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

    CERN Document Server

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-03-01

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

  13. Contribution of the Nea data bank in the field of calculation codes in radiation protection, radio physics and dosimetry

    International Nuclear Information System (INIS)

    Kodeli, I.; Sartori, E.

    2003-01-01

    The Nuclear energy agency is a specialised agency of OECD (organization economic co-operation and development). These missions are to help its members to keep and improve by international cooperation, the scientific, technological and legal bases necessary to a peaceful use of nuclear energy. Nea includes twenty eight countries. Nea works in collaboration with IAEA. The field of activities concerns the acquisition, validation and distribution of nuclear data, calculation codes and experiments. To help users, it organises conferences and training about the calculation codes that it shares out. (N.C.)

  14. DFT calculation for elastic constants of orthorhombic structure within WIEN2K code: A new package (ortho-elastic)

    International Nuclear Information System (INIS)

    Reshak, Ali H.; Jamal, Morteza

    2012-01-01

    Highlights: ► A new package for calculating elastic constants of orthorhombic structure is released. ► The package called ortho-elastic. ► It is compatible with [FP-(L)APW+lo] method implemented in WIEN2k code. ► Several orthorhombic structure compounds were used to test the new package. ► Elastic constants calculated using this package show good agreement with experiment. - Abstract: A new package for calculating the elastic constants of orthorhombic structure is released. The package called ortho-elastic. The formalism of calculating the ortho-elastic constants is described in details. The package is compatible with the highly accurate all-electron full-potential (linearized) augmented plane-wave plus local orbital [FP-(L)APW+lo] method implemented in WIEN2k code. Several orthorhombic structure compounds were used to test the new package. We found that the calculated elastic constants using the new package show better agreement with the available experimental data than the previous theoretical results used different methods. In this package the second-order derivative E ″ (ε) of polynomial fit E=E(ε) of energy vs strains at zero strain (ε=0), used to calculate the orthorhombic elastic constants.

  15. One-dimensional thermohydraulic code THESEUS and its application to chilldown process simulation in two-phase hydrogen flows

    Science.gov (United States)

    Papadimitriou, P.; Skorek, T.

    THESUS is a thermohydraulic code for the calculation of steady state and transient processes of two-phase cryogenic flows. The physical model is based on four conservation equations with separate liquid and gas phase mass conservation equations. The thermohydraulic non-equilibrium is calculated by means of evaporation and condensation models. The mechanical non-equilibrium is modeled by a full-range drift-flux model. Also heat conduction in solid structures and heat exchange for the full spectrum of heat transfer regimes can be simulated. Test analyses of two-channel chilldown experiments and comparisons with the measured data have been performed.

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

    International Nuclear Information System (INIS)

    Okumura, Keisuke

    2015-10-01

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

  17. Design, experiments and Relap5 code calculations for the perseo facility

    International Nuclear Information System (INIS)

    Ferri, Roberta; Achilli, Andrea; Cattadori, Gustavo; Bianchi, Fosco; Meloni, Paride

    2005-01-01

    Research on innovative safety systems for light water reactors addressed to heat removal by in-pool immersed heat exchangers, led to design, build-up and test the PERSEO facility at SIET laboratories. The research started with the CEA-ENEA proposal of improving the GE-SBWR isolation condenser system, by moving the triggering valve from the high pressure primary side of the reactor to the low pressure pool side. A new configuration of the system was defined with the heat exchanger contained in a small pool, connected at bottom and top to a large water reservoir pool, the triggering valve being located on the pool bottom connecting pipe. ENEA funded the whole activity that included the definition and build-up of a new heat exchanger pool, on the basis of the already existing PANTHERS IC-PCC facility, at SIET laboratories, and the new plant requirements. The heat exchanger connections to the pressure vessel were maintained. An experimental campaign was executed at full scale and full thermal-hydraulic conditions for investigating the behaviour and performance of the plant in steady and unsteady conditions. The Relap5 code was utilised during all phases of the research: for the heat exchanger pool dimension definition and from pre-test and post-test analyses. The Cathare code was applied too from pre-test and post-test analyses. This paper deals with the experimental and calculated results limited to the Relap5 code

  18. Processes of code status transitions in hospitalized patients with advanced cancer.

    Science.gov (United States)

    El-Jawahri, Areej; Lau-Min, Kelsey; Nipp, Ryan D; Greer, Joseph A; Traeger, Lara N; Moran, Samantha M; D'Arpino, Sara M; Hochberg, Ephraim P; Jackson, Vicki A; Cashavelly, Barbara J; Martinson, Holly S; Ryan, David P; Temel, Jennifer S

    2017-12-15

    Although hospitalized patients with advanced cancer have a low chance of surviving cardiopulmonary resuscitation (CPR), the processes by which they change their code status from full code to do not resuscitate (DNR) are unknown. We conducted a mixed-methods study on a prospective cohort of hospitalized patients with advanced cancer. Two physicians used a consensus-driven medical record review to characterize processes that led to code status order transitions from full code to DNR. In total, 1047 hospitalizations were reviewed among 728 patients. Admitting clinicians did not address code status in 53% of hospitalizations, resulting in code status orders of "presumed full." In total, 275 patients (26.3%) transitioned from full code to DNR, and 48.7% (134 of 275 patients) of those had an order of "presumed full" at admission; however, upon further clarification, the patients expressed that they had wished to be DNR before the hospitalization. We identified 3 additional processes leading to order transition from full code to DNR acute clinical deterioration (15.3%), discontinuation of cancer-directed therapy (17.1%), and education about the potential harms/futility of CPR (15.3%). Compared with discontinuing therapy and education, transitions because of acute clinical deterioration were associated with less patient involvement (P = .002), a shorter time to death (P cancer were because of full code orders in patients who had a preference for DNR before hospitalization. Transitions due of acute clinical deterioration were associated with less patient engagement and a higher likelihood of inpatient death. Cancer 2017;123:4895-902. © 2017 American Cancer Society. © 2017 American Cancer Society.

  19. Streamline processing of discrete nuclear spectra by means of authoregularized iteration process (the KOLOBOK code)

    International Nuclear Information System (INIS)

    Gadzhokov, V.; Penev, I.; Aleksandrov, L.

    1979-01-01

    A brief description of the KOLOBOK computer code designed for streamline processing of discrete nuclear spectra with symmetric Gaussian shape of the single line on computers of the ES series, models 1020 and above, is given. The program solves the stream of discrete-spectrometry generated nonlinear problems by means of authoregularized iteration process. The Fortran-4 text of the code is reported in an Appendix

  20. Colors and geometric forms in the work process information coding

    Directory of Open Access Journals (Sweden)

    Čizmić Svetlana

    2006-01-01

    Full Text Available The aim of the research was to establish the meaning of the colors and geometric shapes in transmitting information in the work process. The sample of 100 students connected 50 situations which could be associated with regular tasks in the work process with 12 colors and 4 geometric forms in previously chosen color. Based on chosen color-geometric shape-situation regulation, the idea of the research was to find out regularities in coding of information and to examine if those regularities can provide meaningful data assigned to each individual code and to explain which codes are better and applicable represents of examined situations.