Results 1 - 10 of 1473
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[en] It is the aim of this work to demonstrate the potential of this method for solving problems resulting from nuclear engineering applications. From this a concept is derived which allows the development of a program system. (orig./HP)
[de]Ziel ist es, das Potential dieser Methode im Kontext reaktorphysikalischer Anwendungen aufzuzeigen. Daraus ergeben sich Aussagen ueber die Anwendbarkeit der Methode zur Loesung reaktorphysikalischer Fragestellungen. Aufgrund dieser Aussagen wurde ein Konzept fuer die Umsetzung der Methode in ein Programmsystem entwickelt und realisiert. (orig./HP)
[en] The code ICM2D, developed for fast two-dimensional transport calculations in x, y-geometry, is based on a consistent interface current coupling method. The code has an option for treating heterogeneous substructures of complicated geometry by imbedded collision probability calculations. Applications to mixed-oxide fuel assembly calculations are presented and compared with Monte-Carlo solutions. The ability of the code for core calculations is demonstrated. (orig.)
[de]Das Programm ICM2D wurde fuer schnelle zweidimensionale Transportrechnungen in x, y-Geometrie entwickelt. Es basiert auf der Methode der konsistenten Oberflaechenstrom-Kopplung und kann heterogen unterteilte Maschen mit Hilfe von Stosswahrscheinlichkeitsrechnungen behandeln, die ebenfalls ueber Oberflaechenstroeme an die Nachbarmaschen gekoppelt werden. Die Anwendbarkeit des Programms fuer die Berechnung von Buendeln aus Uran- und Mischoxid-Brennelementen wird gezeigt und durch Vergleich mit Monte-Carlo-Loesungen verifiziert. Als weitere Anwendungsmoeglichkeit wird der Einsatz fuer Core-Rechnungen demonstriert. (orig.)
[en] Nuclear reaction data are needed for many applications to determine the production of neutron and gamma radiation. The present upper limit of evaluated data libraries is about 20 MeV. At incident energies above 10 MeV, preequilibrium emission of nucleons has to be considered because it gives a non-negligible contribution at the higher end of the emission spectra. To get the required information, we tested and updated the program Alice, which calculates differential cross-sections with the hybrid model of Blann, and we implemented the systematics of continuum angular distributions by Kalbach. For (n, n), (p, n) and (d, n) reactions, comparisons of the calculational results with published experimental data show satisfying agreement. (orig./HP)
[en] In this paper a process of determination of methodological component of power distribution engineering factors of the macrocode MOBY-DICK (MD) is presented. This process is based on a direct comparison of measured and calculated data. The directly measured data analyzed in this article are thermocouple (TC) and Rhodium self-powered neutron detector (SPND) readings. Also the normality of deviations between measured and calculated TC temperature rises is discussed and methodological uncertainties of the macrocode MD are then checked by 97.5th percentile. Absolute and relative uncertainties of volumetric (KV) and radial (Kq) power distributions are determined by the comparisons of power distributions calculated by MD code with either power distributions provided by the Dukovany NPP core monitoring system (SCORPIO) or with power distributions obtained from direct TC and SPND readings. Final determination of engineering factors supposes hyperbolic dependence of MD methodological uncertainties on power peaking.
[en] The paper reports about the development and verification of the new nodal methods to be used in the KIKO3DMG code. Two classes of the new methods are presented. The first class makes the treatment of the heterogeneities possible inside the assemblies while the extent of the crucial approximations applied on the node boundaries is more considerable in comparison to those in case of the second type. The nodal methods were validated by two VVER reference problems found in the AER benchmark book (aerbench.kfki.hu/aerbench). The AER-2 and FCM-101 benchmarks correspond to the VVER-440 and VVER-1000 geometry, respectively. It was found that the differences between the converged and the reference solutions are negligible from the practical point of view. The performance characteristics concerning the accuracy and the necessary CPU time - both depending on the mesh refinement - were also compared.
[en] The effect of gamma radiation on sorption of 85Sr2+ on trace manganese dioxide and on precipitation of 131I- with Ag+ ions has been studied. The transition of both radionuclides into the solid phase by sorption and precipitation was found by measurement of the self-diffusion coefficients of 85Sr2+ and of 131I-. The gamma radiation had negative effect on sorption and precipitation and also affected the ionic state of 85Sr2+ and of 131I-. (orig.)
[en] The health impacts and corresponding damage costs of radioactive emissions of Yenikoey and Kemerkoey lignite-fired power plants in Mugla have been assessed by using the simplified impact pathway approach. Radiation dose and risk calculations have been carried out by the code CAP88-PC around the power plants. Specific isotopes, 226Ra, 232Th, 40K and 238U in the flying ash samples are considered as radioactive sources. The estimated total collective doses around Yenikoey and Kemerkoey power plants are 3.15 x 10-4 man Sv/year and 3.77 x 10-4 man Sv/year. Health effects and the corresponding damage costs around the power plants due to radioactive emissions from the power plants are negligible. (orig.)
[en] In this study, the molten salt-heavy metal mixtures 93-85 % Li20Sn80 + 5 % SFG-PuO2 and 2-10 % UO2, 93-85 % Li20Sn80 + 5 % SFG-PuO2 and 2-10 % NpO2, 93-85 % Li20Sn80 + 5 % SFG-PuO2 and 2-10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion-fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library. (orig.)
[en] CANDU type reactors have some peculiarities in initiation and progression of severe accidents. In the present paper one of the specificities - the End Fitting Failure accident - is analysed from the point of view of source term formation. The accident is initiated by a failure of the re-fueling machine. Fuel bundles are ejected in the re-fuelling machine room and fuel elements suffer a significant fragmentation by mechanical impact and by the rapid increase of the temperature. A direct transfer of the fission products occurs directly to the containment. The source term in the containment and also the source term to the environment is calculated supposing an open venting communication to the external atmosphere. The simulation is performed by using the ASTEC code in coupled calculation CPA-IODE-ISODOPE-DOSE option. The evolution of the distributions for the most important released fission products is presented for different regions and for different hosts. The most important factors of influence on the source term formation are identified and discussed. (orig.)