Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage
EPJ Web of Conferences, 2017 Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded ...
Ternovykh Mikhail +3 more
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A simple model of reactor cores for reactor neutrino flux calculations for the KamLAND experiment [PDF]
, 2006 KamLAND is a reactor neutrino oscillation experiment with a very long baseline. This experiment successfully measured oscillation phenomena of reactor antineutrinos coming mainly from 53 reactors in Japan. In order to extract the results, it is necessary
A. Suzuki +10 more
core +2 more sources
Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle
EPJ Web of Conferences, 2017 The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development.
Tikhomirov Georgy +4 more
doaj +1 more source
Characteristics of iron and nickel diffusion in molten lead-bismuth eutectic
Mechanical Engineering Journal, 2015 The characteristics of iron (Fe) and nickel (Ni) diffusion in molten lead-bismuth eutectic (LBE) were investigated experimentally by using capillary method.
Yun GAO, Minoru TAKAHASHI, Masao NOMURA
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Consistent neutron-physical and thermal-physical calculations of fuel rods of VVER type reactors
EPJ Web of Conferences, 2017 For modeling the isotopic composition of fuel, and maximum temperatures at different moments of time, one can use different algorithms and codes. In connection with the development of new types of fuel assemblies and progress in computer technology, the ...
Tikhomirov Georgy +3 more
doaj +1 more source
Acta Polytechnica CTU Proceedings, 2022 Current efforts to develop high-fidelity multi-physics tools for research reactors with IRT-4M fuel leads to the involvement of Monte-Carlo transport and subchannel thermo-hydraulic codes in the process of safety assessment.
Pavel Jíška, Jan Frýbort
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UNCERTAINTY PROPAGATION FOR LWR BURNUP BENCHMARK USING SAMPLING BASED CODE SCALE/SAMPLER
Acta Polytechnica CTU Proceedings, 2018 Sampling based method is adopted in many fields of engineering and it is currently used to propagate uncertainties from physical parameters and from nuclear data, to integral indicators of nuclear systems.
Daniel Campolina, Jan Frybort
doaj +1 more source
Dynamical approach to isotopic-distribution of fission fragments from actinide nuclei
EPJ Web of Conferences, 2016 Measurements of the isotope distribution of fission fragments, often denoted as the primary fission yield (pre-neutron yield) or independent fission yield (post-neutron yield) are still challenging at low excitation energies, so that it is important to ...
Ishizuka Chikako +3 more
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POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION
Acta Polytechnica CTU Proceedings, 2016 The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF) is not reached, accidents where ...
Vojtěch Caha, Jakub Krejčí
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Analysis of Heat Release Processes inside Storage Facilities Containing Irradiated Nuclear Graphite
Science and Technology of Nuclear Installations, 2022 The article is dedicated to the safety assessment of mixed storage of irradiated graphite and other types of radioactive waste accumulated during the operation of uranium-graphite reactors.
Alexander O. Pavliuk +4 more
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