Simulation of Steam Bubble Transport in Primary System of Pool Type Lead Cooled Fast Reactors
NURETH-14 - 2011 September 25-30

Presented at:
2011 September 25-30
Toronto, Canada
Session Title:
G1-2 - Sodium Cooled Fast Reactors Design and Safety and Lead/Lead-Bismuth Cooled Reactors

Marti Jeltsov (Royal Institute of Technology)
Pavel Kudinov (Royal Institute of Technology)


Pool-type design makes Lead cooled Fast Reactor (LFR) economically competitive with other advanced reactor designs considered under the Generation IV framework. However, close proximity of steam generator to the core increases the risks associated with Steam Generator Tube Rupture/Leakage (SGTR/SGTL) such as voiding of the core and resulting reactivity insertion and/or local damage (burnout) of fuel rod cladding. Analysis of consequences of SGTL provided in present paper suggests that small bubbles of steam can be dragged by the turbulent coolant flow into the core region. Trajectories of the bubbles are determined by location of the leak, bubbles size and turbulent flow field of lead coolant. The influence of epistemic uncertainty in drag coefficient on prediction of the fraction of bubbles that can reach the core and accumulate in the primary coolant system is discussed in the paper.

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