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Wyszukujesz frazę "Rahgoshay, M." wg kryterium: Autor

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    Tytuł:
    Computation of concentration changes of heavy metals in the fuel assemblies with 1.6% enrichment by ORIGEN code for VVER-1000
    Autorzy:
    Rahgoshay, M.
    Powiązania:
    https://bibliotekanauki.pl/articles/146274.pdf
    Data publikacji:
    2006
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    ORIGEN code
    burn-up
    heavy metals
    BUSHEHR Nuclear Power Plant
    Opis:
    ORIGEN code is a widely used computer code for calculating the buildup, decay, and processing of radioactive materials. During the past few years, a sustained effort was undertaken by ORNL to update the original ORIGEN code [4] and its associated data bases. The results of this effort were updated on the reactor model, cross section, fission product yields, decay data, decay photon data and the ORIGEN computer code itself. In this paper we have obtained concentration changes of uranium and plutonium isotopes by ORIGEN code at different burn-up and then the results have been compared with VVER-1000 results in the first fuel cycle for fuel assemblies with 1.6% enrichment in the BUSHEHR Nuclear Power Plant.
    Źródło:
    Nukleonika; 2006, 51, 3; 161-167
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Study of temperature distribution of fuel, clad and coolant in the VVER-1000 reactor core during group-10 control rod scram by using diffusion and point kinetic methods
    Autorzy:
    Rahgoshay, M.
    Rahmani, Y.
    Powiązania:
    https://bibliotekanauki.pl/articles/146827.pdf
    Data publikacji:
    2007
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    point kinetic
    diffusion
    COSTANZA-R,Z
    VVER-1000
    control rod
    RELAP5
    Opis:
    In this paper, through the application of two different methods (point kinetic and diffusion), the temperature distribution of fuel, clad and coolant has been studied and calculated during group-10 control rod scram, in the Bushehr Nuclear Power Plant (Iran) with a VVER-1000 reactor core. In the reactor core of Bushehr NPP, 10 groups of control rods are used of which, group-10 control rods contain the highest amount of injected negative reactivity in terms of quantity as compared to other groups of control rods. In this paper we explain impacts of negative reactivity, caused by a complete or minor scram of group-10 control rods, on thermoneutronic parameters of the VVER-1000 nuclear reactor core. It should be noted that through these calculations and by using the results, we can develop a sound understanding of impacts of this controlling element in optimum control of the reactor core and, on this basis, with careful attention and by gaining access to a reliable simulation (on the basis of results of calculations made in this survey) we can monitor the VVER-1000 reactor core through a smart control system. In continuation, for a more accurate survey and for comparing results of different calculation systems (point kinetic and diffusion), by using COSTANZA-R,Z calculation code (in which neutronic calculations are based on diffusion model) and using WIMS code at different areas and temperatures (for calculation of constant physical coefficients and temperature coefficients needed in COSTANZAR, Z code) for the VVER-1000 reactor core of Bushehr NPP, calculation of temperature distribution of fuel elements and coolant by using diffusion model is made in the course of group-10 control rods scram and afterwards.
    Źródło:
    Nukleonika; 2007, 52, 4; 159-165
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    A study of the effects of changing burn-up and gap gaseous compound on the gap convection coefficient (in a hot fuel pin) in VVER-1000 reactor
    Autorzy:
    Rahgoshay, M.
    Rahmani, Y.
    Powiązania:
    https://bibliotekanauki.pl/articles/147623.pdf
    Data publikacji:
    2007
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    VVER-1000
    nuclear reactor
    burn-up
    Ross and Stoute model
    gap convection
    hot fuel pin
    thermal expansion
    Opis:
    In this article we worked on the result and process of calculation of the gap heat transfer coefficient for a hot fuel pin in accordance with burn-up changes in the VVER-1000 reactor at the Bushehr nuclear power plant (Iran). With regard to the fact that in calculating the fuel gap heat transfer coefficient, various parameters are effective and the need for designing a model is being felt, therefore, in this article we used Ross and Stoute gap model to study impacts of different effective parameters such as thermal expansion and gaseous fission products on the hgap change rate. Over time and with changes in fuel burn-up some gaseous fission products such as xenon, argon and krypton gases are released to the gas mixture in the gap, which originally contained helium. In this study, the composition of gaseous elements in the gap volume during different times of reactor operation was found using ORIGEN code [3]. Considering that the thermal conduction of these gases is lower than that of helium, and by using the Ross and Stoute gap model, we find first that the changes in gaseous compounds in the gap reduce the values of gap thermal conductivity coefficient, but considering thermal expansion (due to burn-up alterations) of fuel and clad resulting in the reduction of gap thickness we find that the gap heat transfer coefficient will augment in a broad range of burn-up changes. These changes result in a higher rate of gap thickness reduction than the low rate of decrease of heat conduction coefficient of the gas in the gap during burn-up. Once these changes have been defined, we can proceed with the analysis of the results of calculations based on the Ross and Stoute model and compare the results obtained with the experimental results for a hot fuel pin as presented in the final safety analysis report of the VVER-1000 reactor at Bushehr [2]. It is noteworthy that the results of accomplished calculations based on the Ross and Stoute model correspond well with the existing experimental results for this reactor.
    Źródło:
    Nukleonika; 2007, 52, 3; 93-95
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
      Wyświetlanie 1-3 z 3

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