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Kizub P.A.,Blokhin A.I.,Blokhin P.A.,Mitenkova E.F.,Mosunova N.A.,Kovrov V.A.,Shishkin A.V.,Zaikov Yu.P.,Rakhmanova O.R. 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.3
A preliminary criticality analysis for novel pyrochemical apparatuses for the reprocessing of mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor was performed. Hightemperature processing apparatuses, “metallization” electrolyzer, refinery remelting apparatus, refining electrolyzer, and “soft” chlorination apparatus are considered in this work. Computational models of apparatuses for two neutron radiation transport codes (MCU-FR and MCNP) were developed and calculations for criticality were completed using the Monte Carlo method. The criticality analysis was performed for different loads of fissile material into the apparatuses including overloading conditions. Various emergency situations were considered, in particular, those associated with water ingress into the chamber of the refinery remelting apparatus. It was revealed that for all the considered computational models nuclear safety rules are satisfied.
B. Jansky,Z. Turzik,E. Novak,M. Svadlenkova,M. Barta,L. A. Trykov,A. I. Blokhin 한국물리학회 2011 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.59 No.23
The leakage neutron and gamma spectra measurements have been done on benchmark spherical assembly-nickel sphere with diameter of 50 cm. The Cf-252 neutron sources with different emissions were placed into the centre of nickel sphere. The proton recoil method was used for neutron spectra measurement using stilbene crystals and hydrogen proportional counters. The neutron energy range of spectrometer was from 0.02 to 17 MeV. The gamma pulse shape discrimination method has been applied in stilbene measurements. The gamma energy range of spectrometer was from 0.1 to 10 MeV. The fine structure of gamma spectrum was measured by HPGe spectrometer. The experimental data were compared to results of transport calculations based on different evaluated nuclear data libraries (ENDF/B-VII.0, JENDL-3.3, JEFF-3.1.1, TENDL-2009). The continuous energy Monte Carlo transport calculation code MCNP-4C was employed for the calculations. Main observed differences between experiments and transport calculations are discussed.