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    Wyświetlanie 1-8 z 8
    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
    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:
    Investigation of 99Mo potential production via UO2SO4 liquid target irradiation in a 5 MW nuclear research reactor
    Autorzy:
    Gholamzadeh, Z.
    Mirvakili, S. M.
    Davari, A.
    Alizadeh, M.
    Joz-Vaziri, A.
    Powiązania:
    https://bibliotekanauki.pl/articles/148499.pdf
    Data publikacji:
    2017
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    research reactor
    99Mo production
    uranyl sulphate
    liquid target
    MCNPX code
    Opis:
    The activation method for 99Mo production in comparison to fi ssionable target irradiation in research reactors is less preferable. Therefore, 99Mo yield using UO2SO4 samples was theoretically investigated. Computational results revealed admirable potential of the liquid samples for 99Mo production. Low-concentrated uranyl sulphate samples could easily be handled by the irradiation box. The sample geometry optimization improves thermal hydraulic conditions and production yield. The optimized geometry including only 0.12 g 235U produced 57Ci99Mo at end-of-irradiation (EOI) with a temperature peak of 72°C during the irradiation.
    Źródło:
    Nukleonika; 2017, 62, 1; 37-45
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Correction of the electric resistivity distribution of Si wafers using selective neutron transmutation doping (SNTD) in MARIA nuclear research reactor
    Autorzy:
    Tarchalski, M.
    Kordyasz, A. J.
    Pytel, K.
    Dorosz, D.
    Powiązania:
    https://bibliotekanauki.pl/articles/146646.pdf
    Data publikacji:
    2012
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    silicon wafer
    thermal neutron doping
    silicon resistivity homogeneity
    silicon resistivity heterogeneity
    neutron transmutation doping (NTD)
    selective neutron transmutation doping (SNTD)
    MARIA nuclear research reactor
    Opis:
    The result of the electric resistivity distribution modification in silicon wafers, by means of selective neutron transmutation doping (SNTD) method in the MARIA nuclear research reactor at Świerk/Otwock (Poland) is presented. Silicon wafer doping system has been fully designed for the MARIA reactor, where irradiation took place. The silicon wafer resistivity distribution after SNTD has been measured by the capacity voltage (C-V) method. In this article we show first results of this correction technique. The result of the present investigation is that the planar resolution of the correction process is about 4 mm. It is the full width at half maximum (FWHM) of the resistivity distribution produced by thermal neutrons irradiation of Si wafer through a 3 mm hole in the Cd-mask.
    Źródło:
    Nukleonika; 2012, 57, 3; 363-367
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    An application IRIDM the decision making process on fuel conversion of the MARIA reactor
    Zastosowanie IRIDM w procesie decyzyjnym dotyczącym konwersji paliwa w reaktorze MARIA
    Autorzy:
    Borysiewicz, M.
    Kowal, K.
    Prusiński, P. A.
    Dabrowski, M.
    Powiązania:
    https://bibliotekanauki.pl/articles/408038.pdf
    Data publikacji:
    2013
    Wydawca:
    Politechnika Lubelska. Wydawnictwo Politechniki Lubelskiej
    Tematy:
    nuclear safety
    Integrated Risk Informed Decision Making
    Research Reactor MARIA
    Global Threat Reduction Initiative
    nuclear fuel conversion
    bezpieczeństwo jądrowe
    zintegrowany proces decyzyjny
    reaktor badawczy MARIA
    Inicjatywa Redukcji Zagrożeń Globalnych
    konwersja paliwa
    Opis:
    Poland, when acceded to GTRI (Global Threat Reduction Initiative) in 2004, has committed to convert the nuclear fuel of the Research Reactor MARIA, operated by the National Centre for Nuclear Research (NCBJ) in Świerk. The conversion means giving up of high enriched uranium fuel containing 36% of U-235, which was used so far, and replacing it with the low enriched uranium fuel (19.7% U-235). This article describes the potential usability of the Integrated Risk Informed Decision Making (IRIDM) methodology in optimization of the fuel conversion procedure.
    Polska, przystępując w 2004 roku do programu GTRI (Inicjatywa Redukcji Zagrożeń Globalnych), zobowiązała się do konwersji paliwa jądrowego w reaktorze badawczym MARIA, eksploatowanym przez Narodowe Centrum Badań Jądrowych (NCBJ) w Świerku. Konwersja ta oznacza rezygnację z dotychczas użytkowanego paliwa, zawierającego 36% U-235 i zastąpienie go paliwem nisko wzbogaconym (19.7% U-235). Niniejszy artykuł opisuje potencjalne zastosowanie zintegrowanego procesu decyzyjnego (IRIDM) w optymalizacji procedury konwersji paliwa.
    Źródło:
    Informatyka, Automatyka, Pomiary w Gospodarce i Ochronie Środowiska; 2013, 3; 3-6
    2083-0157
    2391-6761
    Pojawia się w:
    Informatyka, Automatyka, Pomiary w Gospodarce i Ochronie Środowiska
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Analiz processa poluchenija biodizelnogo gorjuchego i obosnovanie osnovnykh parametrov reaktora-razdelitelja
    Analysis of the producing biodiesel fuel and justification of basic parameters reactor-separator
    Autorzy:
    Golub, G.
    Kuharets, S.
    Osypchuk, O.
    Kuharets, V.
    Powiązania:
    https://bibliotekanauki.pl/articles/76410.pdf
    Data publikacji:
    2015
    Wydawca:
    Komisja Motoryzacji i Energetyki Rolnictwa
    Tematy:
    biodiesel fuel
    biodiesel production
    vegetable oil
    energy consumption
    reactor
    theoretical research
    Źródło:
    Motrol. Motoryzacja i Energetyka Rolnictwa; 2015, 17, 4
    1730-8658
    Pojawia się w:
    Motrol. Motoryzacja i Energetyka Rolnictwa
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Modelling of thermal hydraulics in a KAROLINA calorimeter for its calibration methodology validation
    Autorzy:
    Luks, A.
    Pytel, K.
    Tarchalski, M.
    Uzunow, N.
    Krok, T.
    Powiązania:
    https://bibliotekanauki.pl/articles/147579.pdf
    Data publikacji:
    2016
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    computional fluid dynamics
    CFD
    differential calorimeter
    gamma heating
    gamma rays
    Jules Horowitz Reactor
    JHR
    modelling of thermal hydraulics
    nuclear heating
    nuclear reactor
    research
    reactor MARIA
    Opis:
    Results of numerical calculations of heat exchange in a nuclear heating detector for nuclear reactors are presented in this paper. The gamma radiation is generated in nuclear reactor during fission and radiative capture reactions as well as radioactive decay of its products. A single-cell calorimeter has been designed for application in the MARIA research reactor in the National Centre for Nuclear Research (NCBJ) in Świerk near Warsaw, Poland, and can also be used in the Jules Horowitz Reactor (JHR), which is under construction in the research centre in Cadarache, France. It consists of a cylindrical sample, which is surrounded by a gas layer, contained in a cylindrical housing. Additional calculations had to be performed before its insertion into the reactor. Within this analysis, modern computational fuid dynamics (CFD) methods have been used for assessing important parameters, for example, mean surface temperature, mean volume temperature, and maximum sample (calorimeter core) temperature. Results of an experiment performed at a dedicated out-of-pile calibration bench and results of numerical modelling validation are also included in this paper.
    Źródło:
    Nukleonika; 2016, 61, 4; 453-460
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Fusion 2050 : European and Polish Perspective
    Autorzy:
    Romaniuk, R. S.
    Powiązania:
    https://bibliotekanauki.pl/articles/226588.pdf
    Data publikacji:
    2014
    Wydawca:
    Polska Akademia Nauk. Czytelnia Czasopism PAN
    Tematy:
    European research programs
    7FP
    accelerator science and technology
    large research infrastructures
    JET
    ITER
    IFMIF
    DEMO
    nuclear fusion
    thermonuclear power engineering
    DT fusion
    Neutrons
    aneutronic technologies
    inertial fusion
    tokomak
    stellarator
    fusion reactor
    tritium
    electronics for fusion
    Opis:
    Fusion, in all its varieties, is a very current subject of science and technology. The results of strongly exothermic reaction of thermonuclear fusion between nuclei of deuterium and tritium are: helium nuclei and neutrons, plus considerable kinetic energy of neutrons of over 14 MeV. DT nuclides synthesis reaction is probably not the most favorable one for energy production, but is the most advanced technologically. More efficient would be possibly an aneutronic fusion. The EU by its EURATOM agenda prepared a Road Map for research and implementation of Fusion as a commercial method of thermonuclear energy generation in the time horizon of 2050. The milestones on this road are tokomak experiments JET, ITER and DEMO, and neutron experiment IFMIF. There is a hope, that by engagement of the national government, and all research and technical fusion and plasma communities, part of this Road Map may be realized in Poland. The infrastructure build for fusion experiments may be also used for material engineering research, chemistry, biomedical, associated with environment protection, power engineering, security, etc. Construction of such research and industrial accelerator and tokomak infrastructure may have potentially a profound meaning for the development of science and technology in Poland.
    Źródło:
    International Journal of Electronics and Telecommunications; 2014, 60, 1; 95-101
    2300-1933
    Pojawia się w:
    International Journal of Electronics and Telecommunications
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
      Wyświetlanie 1-8 z 8

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