Experimental opportunities and main results of the impulse graphite reactor use for research in safety area

Batyrbekov, Erlan; Vityuk, Vladimir; Vurim, Alexander; Vityuk, Galina

doi:10.1016/j.anucene.2022.109582

Cited by 10 publications

(1 citation statement)

References 9 publications

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“…For this purpose, experiments are carried out both in reactor and non-reactor installations. For example, thanks to the design features of the research impulse graphite reactor IGR [1], it is possible to conduct large-scale experiments with a model fuel assembly containing up to 10 kg of uranium dioxide [2]. There is a need to transport nuclear materials to the assembly site of model fuel assemblies to conduct experiments on time.…”

Section: Introductionmentioning

confidence: 99%

Safety Justification for Transportation of Nuclear Materials

Kelsingazina

2024

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The Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan conducts a large number of studies to justify the safety of designed reactors and already operating reactors. Experiments are carried out both at in-pile and out-of-pile research installations. To implement these studies, there is a need to transport nuclear materials to the place where they are performed. Thus, it is necessary to justify the safety of nuclear material transportation. This paper presents neutronic calculations of the effective neutron multiplication factor to justify the safety of nuclear material transportation in accordance with the legislation of the Republic of Kazakhstan. The calculations used the most commonly transported type of nuclear materials – uranium dioxide (UO2) fuel pellets of various enrichments. Transportation was conducted using a transport packaging set IP-1, in accordance with the rules for the transportation of nuclear materials. Both normal and emergency conditions for the transportation of nuclear materials were modeled, taking into account the possibility of water leaking into the packaging and water getting into canisters with uranium dioxide fuel pellets. To determine the effective neutron multiplication factor (keff) of the transport and packaging set, the Monte Carlo calculation method was used in the MCNP5 program, with the ENDF/B-VII nuclear data library.

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Section: Introductionmentioning

confidence: 99%