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Conference Article: Corrosion behavior of sintered pellet of graphite and boron carbide in helium containing water vapor

Fujii, K.; Nomura, S.; Shindo, M. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Dept. of Fuels and Materials Research); Imai, H. (Research Association for Nuclear Facilities Decommissioning, Tokai, Ibaraki (Japan))

Abstract

The corrosion test of the sintered pellet of graphite and boron carbide, which has been adopted as a neutron absorber material in the control rod system, the reserved shutdown system, etc. of the High Temperature Engineering Test Reactor (HTTR), was carried out in helium containing water vapor at up to 1000 deg. C from a viewpoint of oxidation of boron carbide (B4C). Moreover, from the results obtained of the corrosion test, the integrity of the reserved shutdown system was also evaluated. It is shown that the oxidation rates of the sintered pellet simply do not increase with temperature caused by the formation of B2O3. The oxidation reaction of the sintered pellet consists of three competing reactions: B4C+6H2O=2B2O3+6H2+C(F) first step. C(F)+H2O=CO+H2 and C(G)+H2O=CO+H2 second step. For the integrity of the reserved shutdown system, the mutual adhesion of the sintered pellets is one of the key factors because the sintered pellets mutually adhered may not be inserted into the core in an emergency. For the sintered pellet containing 30% B, which is the same B concentration as the specification of the sintered pellet adopted in the reserved shutdown system, mutual adhesion yields in such a case about 2% if the contained B is oxidized and the temperature exceeds 577 deg. C, which is the melting point of B2O3. There is no possibility of both conditions yielding in the HTTR, therefore it can be concluded that the integrity of the reserved shutdown system is not damaged against the corrosion.

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key words: boron carbides; carbon; corrosion; graphite; helium; hot pressing; httr reactor; moderator pellets; reaction kinetics; shutdown; sintered materials; water vapor; boron compounds; carbides; carbon compounds; enriched uranium reactors; experimental reactors; gas cooled reactors; graphite moderated reactors; helium cooled reactors; htgr type reactors; kinetics; materials working
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:169-176