HTGR Knowledge Base
Conference Article: Evaluation of graphite oxidation during water ingress accidents in HTTR
Maruyama, S.; Saikusa, A.; Ikoku, T.; Kunitomi, K.; Shiozawa, S. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Ohashi, K.; Hayakawa, H. (Fuji Electric Co. Ltd., Tokyo (Japan))Abstract
The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The graphite oxidation is one of the major concerns in the water ingress accident with respect to the structural integrity of the core. The water ingress accident is initiated by the rupture of a heat transfer tube in the primary pressurized water cooler. After the steam enters the core, it reacts with graphite structures. In order to evaluate the structural integrity of the graphite structures in the HTTR, the graphite oxidation analysis was made using the OXIDE-3F code. In the present study, it was confirmed that the graphite structures were not oxidized significantly because of rapid temperature decrease due to forced cooling by the auxiliary cooling system. It was also clarified that the water ingressed into the primary cooling system (PCS) did not lead significant pressure increment which could cause safety valve opening and consequent fission products release outside the PCS.
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key words: Gas Cooled Reactor, Nuclear Technology
- Reference:
- Technical committee meeting on response of fuel, fuel elements and gas cooled reactor
cores under accidental air or water ingress conditions. Beijing (China). 25-27 Oct 1993
- International Atomic Energy Agency, Vienna (Austria)
- IAEA-TECDOC--784, pp:97-103
- International Atomic Energy Agency, Vienna (Austria)