This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MWth power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MWth critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinde fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations.
- Nuclear Engineering Division
Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors
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Carlsson, J, Tucek, K, & Wider, H. "Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors." Proceedings of the 14th International Conference on Nuclear Engineering. Volume 3: Structural Integrity; Nuclear Engineering Advances; Next Generation Systems; Near Term Deployment and Promotion of Nuclear Energy. Miami, Florida, USA. July 17–20, 2006. pp. 557-564. ASME. https://doi.org/10.1115/ICONE14-89316
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