Experimental investigation and CFD analysis on cross flow in the core of PMR200

Jeong Hun Lee, Su Jong Yoon, Hyoung Kyu Cho, Moosung Jae, Goon Cherl Park

Research output: Contribution to journalArticle

4 Scopus citations


Abstract The Prismatic Modular Reactor (PMR) is one of the major Very High Temperature Reactor (VHTR) concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear grade graphite. However, the shape of the graphite blocks could be easily changed by neutron damage during the reactor operation and the shape change can create gaps between the blocks inducing the bypass flow. In the VHTR core, two types of gaps, a vertical gap and a horizontal gap which are called bypass gap and cross gap, respectively, can be formed. The cross gap complicates the flow field in the reactor core by connecting the coolant channel to the bypass gap and it could lead to a loss of effective coolant flow in the fuel blocks. Thus, a cross flow experimental facility was constructed to investigate the cross flow phenomena in the core of the VHTR and a series of experiments were carried out under varying flow rates and gap sizes. The results of the experiments were compared with CFD (Computational Fluid Dynamics) analysis results in order to verify its prediction capability for the cross flow phenomena. Fairly good agreement was seen between experimental results and CFD predictions and the local characteristics of the cross flow was discussed in detail. Based on the calculation results, pressure loss coefficient across the cross gap was evaluated, which is necessary for the thermo-fluid analysis of the VHTR core using a lumped parameter code.

Original languageEnglish
Article number4502
Pages (from-to)422-435
Number of pages14
JournalAnnals of Nuclear Energy
Publication statusPublished - 2015 Sep 1



  • Bypass flow
  • CFD
  • Cross flow
  • PMR200
  • Pressure loss coefficient
  • VHTR

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