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Conference Article: The study of metallic fission product release from the VHTR core under power operating condition

Okamoto, F.; Mikami, H. (Nuclear Promotion Div., Fuji Electric Co., Ltd., Tokyo (Japan)); Mitake, S.; Suzuki, K. (Department of Power Reactor Projects, Japan Atomic Energy Research Inst., Tokai, Ibaraki)

Abstract

The experimental VHTR (Very High Temperature Reactor), being developed in Japan Atomic Energy Research Institute, is to be designed to produce the reactor outlet gas temperature of 950 deg. C. It is necessary to investigate more in detail the fission product release characteristics in high temperature range, because the fuel temperature of the VHTR becomes higher than those of HTGRs for steam cycle application. FORNAX code has been developed to investigate the release characteristics of volatile metallic fission products from the core under power operating condition. This code calculates the diffusion of metallic fission products based on the Fick's law of diffusion and can evaluate fission product transport behaviour in the coated fuel particle(cfp), matrix and graphite sleeve. This code can also take into account the distribution of the power, temperature and coating failure in the core and their time history. Several calculations have been carried out to study the metallic fission product release characteristic from the cfp and the core for VHTR configuration (Detailed Design Stage I). In this study the following conclusions are obtained: (1) The contribution of the release by diffusion through an intact TRISO coating to the release from the core becomes larger at higher core temperature. (2) The release by diffusion through an intact TRISO coating depends on not only the diffusion coefficient in the SiC layer but also that in the kernel.

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key words: cesium 137; coated fuel particles; computer calculations; coolants; f codes; fission product release; radioactivity transport; temperature dependence; time dependence; very high temperature; vhtr reactor; beta decay radioisotopes; beta-minus decay radioisotopes; cesium isotopes; computer codes; enriched uranium reactors; experimental reactors; fuel particles; gas cooled reactors; graphite moderated reactors; helium cooled reactors; htgr type reactors; intermediate mass nuclei; isotopes; nuclei; odd-even nuclei; power reactors; radioisotopes; reactors; research and test reactors; thermal reactors
Reference:
Specialists' meeting on safety and accident analysis for gas-cooled reactors. Oak Ridge, TN (USA). 13-15 May 1985
International Atomic Energy Agency, Vienna (Austria)
IAEA-TECDOC--358, pp:203-212