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    Wyświetlanie 1-5 z 5
    Tytuł:
    Neutronic analysis of nanofluids as a coolant in the Bushehr VVER - 1000 reactor
    Autorzy:
    Zarifi, E.
    Jahanfarnia, G.
    Veisy, F.
    Powiązania:
    https://bibliotekanauki.pl/articles/146618.pdf
    Data publikacji:
    2012
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    nanofluids
    neutronic analysis
    VVER-1000
    WIMS-D5
    CITATION
    Opis:
    The main goal of this study was to perform the neutronic analysis of nanofluids as a coolant in reactor simulation. The variation of multiplication factor and thermal power have been investigated in the Bushehr VVER-1000 reactor core with using different nanofluids as coolant. In the applied analysis, water-based nanofluids containing various volume fractions of Al2O3, TiO2, CuO and Cu nanoparticles were used. The addition of different types and volume fractions of nanoparticles were found to have various effects on reactor neutronic characteristics. By using WIMS-D5 and CITATION code, the appropriate nanofluid with optimum volume percentage of nanoparticles was achieved. The results show that at low concentration (0.1% volume fraction) alumina is the optimum nanoparticle for normal reactor operation.
    Źródło:
    Nukleonika; 2012, 57, 3; 375-381
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Neutronic analysis of LBE-uranium spallation target accelerator driven system loaded with uranium dioxide in TRISO particles
    Autorzy:
    Bakir, G.
    Selçuklu, S.
    Genç, G.
    Yapici, H.
    Powiązania:
    https://bibliotekanauki.pl/articles/1068556.pdf
    Data publikacji:
    2016-07
    Wydawca:
    Polska Akademia Nauk. Instytut Fizyki PAN
    Tematy:
    28.20.-v
    28.20.Np
    28.65.+a
    29.20.dg
    29.85.Fj
    Opis:
    This study presents the neutronic performances of fissile breeding and energy production of a gas cooled accelerator-driven system with LBE-uranium dioxide (UO₂) spallation target. The accelerator-driven system is designed and optimized by considering various target materials, in terms of neutronic. Two different materials, LBE + natural UO₂ and LBE + 15% enrichment UO₂ are selected as target materials. The target zone is divided into two parts, one within the other; the outer part is pure LBE target part, and the inner part is UO₂ target part cooled with the helium gas. Tristructural-isotropic (TRISO)-coated fuel particles, containing UO₂ fuel, are embedded in a carbon matrix pebble with the packing fraction of a 29%, and the pebbles are placed in the UO₂ target part and in the fuel core with the packing fraction of a 60%. The fuel core is cooled with helium that is a high-temperature coolant. The target is bombarded with the continuous beams of a 1 GeV protons to produce high-flux neutrons that enter the fuel core. The fuel core is surrounded with a graphite reflector zone serving as both effective moderation and reflection of these neutrons. Furthermore, the whole system is enclosed by boron carbide, B₄C (shielding zone), to prevent the neutrons leakage out of the accelerator-driven system. The high-energy Monte Carlo code MCNPX along with the LA150 library is used for neutronic calculations. The numerical results bring out that the investigated accelerator-driven system has a high neutronic performance, from the energy production and fissile breeding points of view. Namely, it can be obtained over the thermal power of a 350 MW and produced over the fissile breeding of a 300 g/day.
    Źródło:
    Acta Physica Polonica A; 2016, 130, 1; 30-32
    0587-4246
    1898-794X
    Pojawia się w:
    Acta Physica Polonica A
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Production of fission product 99Mo using high-enriched uranium plates in Polish nuclear research reactor MARIA: Technology and neutronic analysis
    Autorzy:
    Jaroszewicz, J.
    Marcinkowska, Z.
    Pytel, K.
    Powiązania:
    https://bibliotekanauki.pl/articles/147625.pdf
    Data publikacji:
    2014
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    fission products
    99Mo production
    neutronic calculations
    research reactor
    Opis:
    The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.
    Źródło:
    Nukleonika; 2014, 59, 2; 43-52
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code
    Autorzy:
    Aldawahra, S.
    Khattab, K.
    Gorge, S.
    Powiązania:
    https://bibliotekanauki.pl/articles/971506.pdf
    Data publikacji:
    2015
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    MNSR reactor
    HEU fuel
    LEU fuel
    MCNP4C code
    safety parameters
    Opis:
    Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (κ eff), excess reactivity (ρ ex), control rod worth (CRW), shutdown margin (SDM), safety reactivity factor (SRF), delayed neutron fraction (β eff) and the neutron fl uxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fl uxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.
    Źródło:
    Nukleonika; 2015, 60, 2; 367-371
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    RELAP5/MOD3 model and transient analyses for the MARIA research reactor in Poland
    Autorzy:
    Szczurek, J.
    Czerski, P.
    Bykowski, W.
    Powiązania:
    https://bibliotekanauki.pl/articles/147320.pdf
    Data publikacji:
    2004
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    research reactor
    transient analysis
    neutronic parameters
    qualification
    Opis:
    The RELAP5/MOD3 input data model of the MARIA research reactor has been developed to provide the capability for the analysis of the reactor core under loss of flow and reactivity insertion transients. The model was qualified against the reactor data at steady state conditions and, additionally, against the existing reliable experimental data for a transient initiated by the reactor scram. The results obtained with the code agree well with the experimental data. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. The presented input data model should be treated as the first step for developing of the model including the whole primary cooling circuit of the reactor.
    Źródło:
    Nukleonika; 2004, 49, 4; 149-157
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
      Wyświetlanie 1-5 z 5

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