Fast Reactors and Accelerator Driven Systems Knowledge Base

Conference Article: Analysis of primary pipe break for the Korean advanced liquid metal reactor (KALIMER)

Abstract

A postulated break in the primary pump discharge pipe is analyzed to assure the inherent safety of the Korean Advanced Liquid Metal Reactor (KALIMER), a pool-type liquid metal-cooled reactor generating 392 MWth of power in the core. The main concern of the analysis is the amount of increase in the fuel and the coolant temperatures. The stabilization of the transient due to reactivity feedback is also important. In the present analysis, it is assumed that one of the four pipes connecting the pump discharge to the core inlet plenum is broken. The break is located 3.7 m below the pump outlet and the diameter of the break is 0.4 m. It is also assumed that the reactor is not scrammed after the initiation of the break, therefore, the pumps keep on running during the accident. The analysis is performed with the SSC-K code, which was developed for the analysis of the transient system response of a pool-type reactor. As soon as the break occurs, the core flow decreases drastically to 65% of full flow in the base case. A more conservative case is also analyzed, in which the core flow is reduced artificially to 50% full flow. The reactor power stabilizes by the reactivity feedback effects in about 10 minutes. The increase of the fuel and coolant temperatures due to the sudden reduction of the core flow are also mitigated with a large margin to coolant saturation temperature. The gas expansion module plays an important role providing the dominant reactivity feedback when the core flow is reduced to less than 50% full power. It is evident from these results that both a sufficient subcooling margin of more than 400 K and a stable system response are maintained in the KALIMER design during the primary pipe break accident, which guarantees the inherent safety of KALIMER against a pipe break. (author)

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Reference:

Proceedings of a technical meeting held in Kalpakkam, India 13-17 January 2003

International Atomic Energy Agency, Vienna (Austria)

IAEA-TECDOC--1406, pp:105-123