Institute for Neutron Physics and Reactor Technology (INR)

The subchannel thermal hydraulic code SubChanFlow

SubChanFlow solves a system of mixture equations (mass, momentum, enthalpy) for stationary and transient single and two-phase upward flow in rod bundles or cores. The conservation of the momentum includes lateral flow between neighbouring sub-channels in a simplified manner. The latest version includes a new solver for downward and buoyancy driven flow. Properties for different working fluids are implemented e.g. the IAPWS-97 for water and steam and functions for liquid metals (sodium and lead), and gases (helium, air, etc.). The heat conduction in the fuel rod is calculated based on a finite volume method, where temperature dependent thermo-physical properties (density, heat conductivity, heat capacity) of fuel (UO2, UO2PuO2) and cladding (Zircaloy, stainless steel) materials are implemented.  The heat conduction solver can handle rod and plate as fuel geometry. 

Key-features and physical phenomena

  • Slip model for boiling including sub-cooled boiling
  • Several “Critical Heat Flux” correlations for the prediction of safety parameters e.g. MDNBR
  • Pressure drop models including spacer grids
  • Flow regime-dependent turbulent crossflow mixing models
  • (equal mass or equal volume)
  • Radial fuel pin heat conduction
  • Simplified correlation based fuel pin behaviour
  • (cracking, swelling, fission gas release)
  • Gap conductance model for the gap between the fuel and the cladding
  • Point kinetic model based transient power calculation
  • Boron transport model

Potential applications

  • Subchannel level analysis of different fuel assembly designs
  • Full core analysis at subchannel level  (if pin power is provided)
  • Fuel assembly level analysis of cores with different fuel assemblies (square, hexagonal, MTR-type)
  • Analysis of bundle experiments representing different core designs


Coupling options

SubChanFlow is coupled with neutronic solvers using different coupling methodologies e.g.

  • Coupling of MCNP5/SubChanFlow based on internal coupling optimized to solve full cores at pin /subchannel level
  • Coupling of Serpent2/SubChanFlow based on an internal master-sleeve approach very much appropriate for the simulation of transients developed in H2020 McSAFE project
  • Coupling of Serpent2/SubChanFlow/Transuranus based on the modular and standardized ICoCo-approach for the static and depletion problems of full cores with square (PWR) and hexagonal (VVER) fuel assemblies optimized for pin /subchannel level simulations of real cores (developed in H2020 McSAFE Project)
  • Coupling of SubChanFlow within the NURESIM Platform with different Diffusion solvers at nodal level e.g. COBAYA3, DYN3D, CRONOS2 based on the MED-MED coupling approach
  • Coupling of SubChanFlow with the SP3-transport solver (DYN3D-SP3) for the simulation of transient at pin /subchannel level in PWRs
  • Multiscale coupling of SubChanFlow with different thermal hydraulic codes e.g. TRACE (based on ECI and ICoCo), TrioCFD (based on ICoCo) for improved TH-simulation of reactor systems


The validation of SubChanFlow is based on relevant experimental data for PWR (e.g. OECD Benchmark PSBT Tests) and BWR (OECD Benchmark BFBT) that includes stationary and transient tests, where key parameters such as pressure drop, void fraction, critical power, etc. are measured.

User Community

The user community of SubChanFlow consists of 22 institutions (universities, research centres, TSOs) of Europe, Middle East, Asia and Latin America.


  1. U. Imke and V. Sanchez; Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT). Science and Technology of Nuclear Installations. Volume 2012, Article ID 465059, 12 pp.
  2. A. Gomez Torres, W. Jaeger, V.H. Sánchez Espinoza and U. Imke; On the Influence of Shape Factors for CHF Predictions with SUBCHANFLOW during a Rod Ejection transient. 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9). P0057. Kaohsiung, Taiwan, September 2012.
  3. A. Berkhan, V. Sánchez and U. Imke; Validation of PWR relevant models of SUBCHANFLOW using the NUPEC PSBT Data. The 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14). Log Number: 191. Hilton Toronto Hotel, Toronto, Ontario, Canada, September 25-29, 2011
  4. V. Sánchez, U. Imke, R. Gomez; SUBCHANFLOW: A Thermal-Hydraulic Sub-Channel Program to Analyse Fuel Rod Bundles and Reactor Cores. 17th Pacific Basin Nuclear Conference. Cancún, Q.R., México, October 24-30, 2010.
  5. U. Imke, V. Sanchez, R. Gomez; SUBCHANFLOW: An empirical knowledge based subchannel code. Annual Meeting on Nuclear Technology. 4-6 may 2010. Berlin Germany.