Institute for Neutron Physics and Reactor Technology (INR)

Subchannel Code SUBCHANFLOW

Reactor Safety: Development of 3D Thermal-hydraulic Codes

SUBCHANFLOW: Fast Running Subchannel Code for LWR and Innovative Reactors

Flow regime dependent heat transfer models

  • Slip model for boiling including sub-cooled boiling

  • Several “Critical Heat Flux” correlations

  • Pressure drop models including spacers and wire-wraps

  • Flow regime dependent turbulent cross-flow mixing models
    (equal mass or equal volume)

  • 2D (r-z) fuel pin heat conduction, axially for transients, only

  • Simplified correlation based fuel pin behaviour
    (cracking, swelling, fission gas release)

  • Gap conductance model for fuel-cladding gap

  • Water EOS and properties: IAPWS-97 formulation

  • Point kinetic model based transient power calculation

  • Optimized for whole core subchannel-wise simulation of PWR cores: OpenMP parallelization implemented

  • Fast running code: ¼ PWR core with 18144 subchannels and 20 axial nodes ~ 4 min. (8 threads)

  • Coupled with Monte Carlo Codes (MCNP and SERPENT) for reference solutions

  • Coupled with deterministic Diffusion (COBAYA3) and transport (DYN3D-SP3) for prediction of local parameters

Validation based on bundle tests and international Benchmarks such as PWR PSBT and BFBT

  • Steady state pressure drop and void fraction tests

    • Transient test representing main recirculation pump or turbine trip