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Conference Article: Development of graphite for fuel element sleeves in advanced gas cooled reactors

Burridge, D.P.; Naylor, J.E. (British Nuclear Fuels plc, Salwick (United Kingdom). Fuel Engineering Dept.)

Abstract

The graphite sleeve is the prime structural member of the advanced gas cooled reactor (AGR) fuel element in that each sleeve supports the pin cluster and all of the fuel assembly above it. It must have a low oxidation rate, a low permeability to maintain coolant flow in the channel, must be dimensionally stable within certain limits and have sufficient strength at the end of life to sustain refuelling operations. The paper describes a development programme covering design and improved graphite. Data is presented which is related to that anticipated in reactor behaviour of the material. It then proceeds to identify the quality assurance requirements of the sleeve material. Graphite sleeves produced from this programme of work are currently being irradiated as pilot loadings in UK reactors.

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key words: agr type reactors; carbon; carbonates; fuel assemblies; fuel elements; graphite; materials testing; mechanical structures; physical radiation effects; quality assurance; enriched uranium reactors; gas cooled reactors; graphite moderated reactors; radiation effects; reactor components; reactors; testing
Reference:
Specialists' meeting on the status of graphite development for gas cooled reactors. Tokai, Ibaraki (Japan). 9-12 Sep 1991
International Atomic Energy Agency, Vienna (Austria)
IAEA-TECDOC--690, pp:134-139