Tests of Models of BREST-OD-300 Reactor Fuel Elements in an Autonomous Lead-Cooled Channel of a BOR-60 Reactor

Autor:	A. S. Korol'kov, Yu. M. Pevchikh, B. D. Rogozkin, V. N. Leonov, A. I. Filin, V. V. Brovko, A. G. Sila-Novitskii, A. D. Yurchenko, V. N. Efimov, A.A. Pikalov, P. N. Martynov, K. D. Ivanov, A. A. Ryzhov, A. G. Kordyukov, Yu. A. Kostyukov, Yu. I. Orlov, V. M. Troyanov, Yu. E. Stynda
Rok vydání:	2004
Předmět:	Materials science Nuclear fuel Nuclear engineering chemistry.chemical_element Uranium Coolant Plutonium Lead (geology) Nuclear Energy and Engineering chemistry Volume (thermodynamics) Square array Nuclear chemistry Communication channel
Zdroj:	Atomic Energy. 97:564-570
ISSN:	1063-4258
DOI:	10.1023/b:aten.0000047683.77555.98
Popis:	A special feature of the BREST-OD-300 reactor that is now being designed is that it employs a container-type heat-conducting fuel element with mixed uranium–plutonium mononitride fuel, a lead sublayer, and an expansion volume at the top to collect gaseous products. The fuel elements are arranged in a square array with a wide spacing and are spaced by laminated spacing lattices.
Databáze:	OpenAIRE
Externí odkaz:	https://explore.openaire.eu/search/publication?articleId=doi_________::7631cc00e04c1ce2a815818656c68e60 https://doi.org/10.1023/b:aten.0000047683.77555.98 Zobrazit plný text záznamu Full text from SpringerLink