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Wyszukujesz frazę "Cetnar, J." wg kryterium: Autor

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    Wyświetlanie 1-8 z 8
    Tytuł:
    Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR
    Autorzy:
    Oettingen, M.
    Cetnar, J.
    Powiązania:
    https://bibliotekanauki.pl/articles/148636.pdf
    Data publikacji:
    2015
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    comparative analysis
    major actinides
    MCB
    Monte Carlo
    ressurized water reactor (PWR)
    Opis:
    In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB) in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238) and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242). The main results were presented as a calculated-to-experimental ratio (C/E) for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55). The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.
    Źródło:
    Nukleonika; 2015, 60, No. 3, part 2; 571-580
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Assesment of advanced step models for steady state Monte Carlo burnup calculations in application to prismatic HTGR
    Autorzy:
    Kępisty, G.
    Cetnar, J.
    Powiązania:
    https://bibliotekanauki.pl/articles/146581.pdf
    Data publikacji:
    2015
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    burnup
    high temperature gas-cooled reactor (HTGR)
    Monte Carlo
    stability
    step model
    Opis:
    In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5). The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR) system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.
    Źródło:
    Nukleonika; 2015, 60, No. 3, part 2; 523-529
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent
    Autorzy:
    Królikowski, I. P.
    Cetnar, J.
    Powiązania:
    https://bibliotekanauki.pl/articles/147457.pdf
    Data publikacji:
    2015
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    code coupling
    modeling
    Monte Carlo
    neutronics
    nuclear reactor
    thermal hydraulics
    Opis:
    Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD) code. The main purpose of the coupling is to exchange data such as temperature and power profi le between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profi le is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant flow is generated only by natural convection.
    Źródło:
    Nukleonika; 2015, 60, No. 3, part 2; 531-536
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste
    Autorzy:
    Stanisz, P.
    Cetnar, J.
    Domańska, G.
    Powiązania:
    https://bibliotekanauki.pl/articles/146516.pdf
    Data publikacji:
    2015
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    adiabatic reactor
    closed nuclear fuel cycle
    lead-cooled fast reactor (LFR)
    nuclear reactor core design
    Opis:
    The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR) was defi ned and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System) and LEADER (Lead-cooled European Advanced Demonstration Reactor) projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA) are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs), and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB) code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.
    Źródło:
    Nukleonika; 2015, 60, No. 3, part 2; 581-590
    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:
    The MCB code for numerical modeling of Fourth Generation nuclear reactors
    Autorzy:
    Oettingen, M.
    Cetnar, J.
    Mirowski, T.
    Powiązania:
    https://bibliotekanauki.pl/articles/305705.pdf
    Data publikacji:
    2015
    Wydawca:
    Akademia Górniczo-Hutnicza im. Stanisława Staszica w Krakowie. Wydawnictwo AGH
    Tematy:
    Monte Carlo
    nuclear reactors
    radiation transport
    MCB
    VHTR
    LFR
    Opis:
    R&D in the nuclear reactor physics demands state-of-the-art numerical tools that are able to characterize investigated nuclear systems with high accuracy. In this paper, we present the Monte Carlo Continuous Energy Burnup Code (MCB) developed at AGH University’s Department of Nuclear Energy. The code is a versatile numerical tool dedicated to simulations of radiation transport and radiation-induced changes in matter in advanced nuclear systems like Fourth Generation nuclear reactors.We present the general characteristics of the code and its application for modeling of Very-High-Temperature Reactors and Lead-Cooled Fast Rectors. Currently, the code is being implemented on the supercomputers of the Academic Computer Center (CYFRONET) of AGH University and will soon be available to the international scientific community.
    Źródło:
    Computer Science; 2015, 16 (4); 329-350
    1508-2806
    2300-7036
    Pojawia się w:
    Computer Science
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Assessment of the control rods shadow effect in the VENUS-F core
    Autorzy:
    Cetnar, J.
    Domańska, G.
    Gajda, P.
    Janczyszyn, J.
    Powiązania:
    https://bibliotekanauki.pl/articles/147908.pdf
    Data publikacji:
    2014
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    accelerator driven systems (ADS)
    control rod
    FREYA
    GUINEVERE
    reactivity
    Opis:
    The partitioning and transmutation (P&T) of spent nuclear fuel is an important fi eld of present development of nuclear energy technologies. One of the possible ways to carry out the P&T process is to use the accelerator driven systems (ADS). This technology has been developed within the EURATOM Framework Programmes for several years now. Current research in this fi eld is carried out within the scope of 7th FP project FREYA. Important parts of the project are experiments performed in the GUINEVERE facility devoted to characterising the subcritical core kinetics and development of reactivity monitoring techniques. The present paper considers the effects of control rods use on the core reactivity. In order to carry out the evaluation of the experimental results, it is important to have detailed core characteristics at hand and to take into consideration the differences in the effect of control rods acting separately or together (the so-called shadow effect) on both the reactivity value and the measured neutron fl ux. Also any core asymmetry should be revealed. This goal was achieved by both MCNP simulations and the experimental results. However, in the case of experimental results, the need for calculating respective correction factors was unavoidable.
    Źródło:
    Nukleonika; 2014, 59, 4; 137-143
    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
      Wyświetlanie 1-8 z 8

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