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    Wyświetlanie 1-6 z 6
    Tytuł:
    Reconsideration of thermonuclear possibilities of Z-pinches
    Autorzy:
    Vikhrev, V. V.
    Powiązania:
    https://bibliotekanauki.pl/articles/147474.pdf
    Data publikacji:
    2001
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    fusion burn wave
    neutrons
    plasma focus
    Z-pinch
    Opis:
    The paper considers the Z-pinch as the basis for future thermonuclear fusion reactors. Experiments on Z-pinches always concern small and high temperature and a high density plasma regions that arise spontaneously in the Z-pinch neck. A burn wave might be initiated in the Z-pinch column if in this small plasma region a Lawson-like condition is fulfilled.
    Źródło:
    Nukleonika; 2001, 46, suppl. 1; 9-12
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    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:
    Depletion analysis of the HELIOS experiment using the MCB code
    Autorzy:
    Oettingen, M.
    D'Agata, E.
    Döderlein, C.
    Tuček, K.
    Cetnar, J.
    Powiązania:
    https://bibliotekanauki.pl/articles/146842.pdf
    Data publikacji:
    2012
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    depletion
    HELIOS
    Monte Carlo continuous energy burn-up code (MCB)
    minor actinides
    Opis:
    The focus of our studies is to present an advanced depletion analysis of the HELIOS experiment by means of the Monte Carlo continuous energy burn-up code (MCB). The MCB was used mainly to calculate nuclide density evolution in nuclear reactor cores. We present the capability of the MCB to investigate the depletion of nuclear fuel samples irradiated in the HELIOS experiment. In our studies we traced the behaviour of the main fissile isotopes, 242mAm and 239Pu, respectively. We also perform a sensitivity analysis to the choice of JEF2.2 and JEFF3.1 cross section libraries in terms of the released fission power and the evolution of actinide inventories. The amount of He produced at the end of irradiation, as well as Am and Pu depletion, were also considered.
    Źródło:
    Nukleonika; 2012, 57, 4; 435-441
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    On the neutronics of European lead-cooled fast reactor
    Autorzy:
    Cetnar, J.
    Oettingen, M.
    Domańska, G.
    Powiązania:
    https://bibliotekanauki.pl/articles/148566.pdf
    Data publikacji:
    2010
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    fuel burn-up
    plutonium
    minor actinides
    oxide fuel
    nitride fuel
    LFR
    ELSY
    MCB
    Opis:
    The perspective of nuclear energy development in the near future imposes a new challenge on a number of sciences over the world. For years, the European Commission (EC) has sponsored scientific activities through the framework programmes (FP). The lead-cooled fast reactor (LFR) development in the European Union (EU) has been carried out within European lead-cooled system (ELSY) project of the 6th FP of EURATOM. This paper concerns the reactor core neutronic and burn-up design studies. We discuss two different core configurations of ELSY reactor; one loaded with the reference – mixed oxide fuel (MOX), whereas the second one with an advanced fuel – uranium- -plutonium nitride. Both fuels consist of reactor grade plutonium, depleted uranium and additionally, a fraction of minor actinides (MA). The fuel burn-up and the time evolution of the reactor characteristics has been assessed using a Monte Carlo burn-up code (MCB). One of the important findings concerns the importance of power profile evolution with burn-up as a limiting factor of the refuelling interval.
    Źródło:
    Nukleonika; 2010, 55, 3; 317-322
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code
    Autorzy:
    Gholamzadeh, Z.
    Feghhi, S. A. H.
    Soltani, L.
    Rezazadeh, M
    Tenreiro, C.
    Joharifard, M.
    Powiązania:
    https://bibliotekanauki.pl/articles/148179.pdf
    Data publikacji:
    2014
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    ThO2
    neutronic parameters
    fuel burn-up
    233U
    235U
    239Pu fissile material
    Opis:
    Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fi ssile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fi ssile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th) O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view.
    Źródło:
    Nukleonika; 2014, 59, 4; 129-136
    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 1-6 z 6

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