HTGR Knowledge Base
Conference Article: Fission product diffusion in fuel element materials for HTGR
Chernikov, A.S.; Khromov, Yu.F.; Lyutikov, R.A.; Gusev, A.A. (I.V. Kurchatov Institute of Atomic Energy, Moscow (Russian Federation))Abstract
Diffusion characteristics of radioactive noble gases and of some solid fission products in fuel element materials of HTGR were researched. Values of diffusion coefficients of fission gas products in UO2 kernel at 1450-2000 K in pyrocarbon with density of 1.6-1.8 g/cm3 (1300-1900 K) and in matrix graphite with density of 1.82-1.84 g/cm3 (1200-1700 K) and also diffusion coefficients of silver, barium, cerium in pyrocarbon, silicon carbide, zirconium carbide and promethium in zirconium carbide and uranium dioxide in a wide temperature range were obtained. It was shown that before loss of fuel particle, sealing radioactive noble gas release from a fuel element is determined by technological contamination of matrix graphite with fission material. It's noted that while estimating total fission product release from fuel particles, it is necessary to take into consideration their transport along short-circuit diffusion paths.
view the full text of this article (12 pages, format: PDF, size= 395kB)
key words: Gas Cooled Reactor, Nuclear Technology
- Reference:
- Specialists' meeting on fission product release and
transport in gas-cooled reactors Berkeley (United
Kingdom) 22-25 Oct 1985
- International Atomic Energy Agency, International Working Group on Gas-Cooled Reactors, Vienna (Austria)
- IWGGCR--13, pp:170-181
- International Atomic Energy Agency, International Working Group on Gas-Cooled Reactors, Vienna (Austria)