We found a match
Your institution may have access to this item. Find your institution then sign in to continue.
- Title
Tests of fuel elements with uranium-plutonium nitride fuel in an IGR pulsed reactor.
- Authors
Kaplienko, A. V.; Lemekhov, V. V.; Cherepnin, Yu. S.; Moiseyev, A. V.; Zhirnov, A. P.; Ivanyuta, A. N.; Rozhdestvenskiy, I. M.; Loginov, D. Yu.; Mezhina, Ye. R.; Izhutov, A. L.; Zvir, Ye. A.; Shevlyakov, G. V.; Volkova, I. N.; Batyrbekov, Ye. G.; Baklanov, V. V.; Korovikov, A. G.; Kotlyar, A. N.; Miller, A. A.; Irkimbekov, R. A.; Vurim, A. D.
- Abstract
Fuel elements with mixed uranium-plutonium nitride fuel were tested in an IGR reactor to justify their application in a BREST-OD-300 reactor. The mid-radial enthalpy limit for the fresh mixed nitride fuel as experimentally determined during IGR launching with fast reactivity input amounted to 167 cal/g. The maximum temperature of 1000 °C and its maintenance for 100 s, comprising one of the design limits for the fuel cladding temperature, was experimentally confirmed. The main results of the performed experiments and post-irradiation studies are analyzed.
- Subjects
NITRIDES; FAST reactors; URANIUM; ENTHALPY
- Publication
Atomic Energy, 2023, Vol 134, Issue 5/6, p275
- ISSN
1063-4258
- Publication type
Article
- DOI
10.1007/s10512-024-01055-1