Scaling approach and forced-circulation flow patterns in the liquid metal-cooled reactor-pool experiment E-SCAPE

Katrien Van Tichelen, Fabio Mirelli, Matteo Greco, Yann Bartosiewicz, William D'haeseleer

Research outputpeer-review


The validation of decay heat removal systems and the characterization of thermal hydraulic phenomena in the plena of the liquid metal-cooled, pool-type research reactor MYRRHA, under design at SCK CEN, the Belgian Nuclear Research Centre, is accomplished by experiments and numerical investigations. For this purpose, the E-SCAPE facility at SCK CEN is a thermal hydraulic 1/6-scale 3-D model of the primary system of MYRRHA, with an electrical core simulator, and cooled with Lead Bismuth Eutectic (LBE). Its scaling is based on the preservation of the overall system behavior and the reproduction of the major thermal hydraulic phenomena under decay heat removal conditions. Specific attention in the scaling approach is given to stratification phenomena in the upper plenum and the mixing effect associated with jets entering the plena. From steady-state isothermal operation in forced circulation, experimental information is obtained on the pressure losses in the system. The relation between the pressure loss in the core and above core structure, and the level difference between the lower and upper plenum is confirmed. From hot plug experiments, using temperature as a flow tracer, a large amount of information on the flow patterns in forced circulation mode is generated. For instance, a clear difference can be observed in the behavior of pump jets entering the lower plenum between the high and low flow rate cases. One can also see that the LBE that exits the core, is going upwards through the above core structure before distributing radially in the uppermost fluid layers. From these layers, the fluid goes downward in the upper plenum towards the inlet windows of the heat exchangers.

Original languageEnglish
Article number113361
Number of pages17
JournalNuclear Engineering and Design
StatePublished - Sep 2024

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • General Materials Science
  • Nuclear Energy and Engineering
  • Safety, Risk, Reliability and Quality
  • Waste Management and Disposal
  • Mechanical Engineering

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