Predicted and measured response of the EBR-II plant to large steam pressure changes [electronic resource].
- Argonne, Ill. : Argonne National Laboratory, 1987. and Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy.
- Physical Description:
- Pages: 4 : digital, PDF file
- Additional Creators:
- Argonne National Laboratory and United States. Department of Energy. Office of Scientific and Technical Information
- Restrictions on Access:
- Free-to-read Unrestricted online access
- The Experimental Breeder Reactor II (EBR-II) is a liquid metal reactor (LMR) whose sodium-bonded metallic fuel core has substantial negative reactivity feedback. It has been demonstrated that this feedback enables a loss-of-flow without scram (LOFWS) to shut the reactor down without operator action. This inherent effect also causes a reactor shutdown following a loss-of-heat-sink without scram (LOHSWS). On April 3, 1986, an LOFWS and an LOHSWS were each performed from full power. In the latter test the secondary sodium flow was reduced to less than 1% about three minutes. This caused the reactor inlet temperature to increase only about 45/sup 0/C and essentially stopped the fission process and shut the reactor down. These tests demonstrated that an LMR plant can be designed in which natural phenomena, rather than electromechanical systems (such as those which move control rods), are effective in protecting the reactor against the potentially adverse consequences of loss-of-primary-flow and loss-of-heat-sink accidents. Moreover, the same phenomena which shut the reactor down during these two tests could be exploited to allow primary flow or inlet temperature, rather than control rods, to be used to maneuver plant output within a considerable power range. This capability would enable an LMR to be designed in which far less reactivity is invested in the control rods than is currently the practice. The ultimate goal is to design an ''inherently safe'' LMR in which reactor safety does not depend upon control rods and also where severe rod withdrawal accidents do not need to be considered.
- Published through SciTech Connect., 01/01/1987., "conf-870418-1", "DE87001580", Topical meeting on anticipated and abnormal transients in nuclear power plants, Atlanta, GA, USA, 12 Apr 1987., and Chang, L.K.; Feldman, E.E.; Betten, P.R.; Mohr, D.; Planchon, H.P.; Messick, N.C.
- Funding Information:
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