DYN3D Modellierung

Description of the Physical Model of DYN3D


Neutron Kinetics

  • The reactor core is divided into volume elements(nodes), each characterized by homogenized nuclear data. The nodalisation is determined by the structure of the fuel assemblies (rectangular or hexagonal) and an axial division of the core into layers.
  • The three-dimensional neutron diffusion equation for two energy groups is solved by nodal methods 
  • The transversal integration method is used to solve the neutron diffusion equation within the nodes.
  • In the case of Cartesian geometry the three-dimensional diffusion equation is transformed into one-dimensional equations by using quadratic approximations for the transversal leakages. The nodes are coupled by the side averaged values of neutron flux and current.
  • Considering the hex-z geometry the diffusion equation is split into a two-dimensional equation in the hexagonal plane and a one-dimensional equation for the z-direction by the transverse integration. The nodes can be coupled by the side averaged values or both the sides averaged values and the edge averaged values of neutron flux and current.
  • An implicite difference scheme with exponential transformation is used for time integration

          Recent development:

  • The treatment of  the energy dependence of the neutrons was extended from two to multi groups
  • A nodal SP3 transport model was implemeted for improving the accuracy of the neutronic model based on diffusion approximation
  • Features:
  • Burn-up calculation
  • Criticality by change of neutron multiplication
  • Search for critical boron concentration
  • Search for critical power
  • Albedo boundary conditions different for each side of core or reflector surface
  • Assembly discontinuity factors (ADF)
  • Decay heat model
  • Equilibrium Xe/Sm concentration
  • Poison concentrations can be given by input from previous calculations
  • Moving control rods in transient calculations
  • Calculation of dynamic reactivity
  • Nodal flux reconstruction for pin power calculation

Thermal Hydraulics and Fuel Rod Model

  • The reactor core is modelled by parallel cooling channels describing one or more fuel assemblies. The parallel channels are coupled hydraulically by the condition of equal pressure drop over all core channels.
  • A one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy and momentum of the two-phase mixture and the mass balance for the vapour phase allows the description of thermodynamic non-equilibrium between the phases.
  • Time-dependent thermo-hydraulic boundary conditions for the core like coolant inlet temperature, pressure, coolant mass flow rate, or pressure drop can be given as input or obtained from the coupled thermohydraulic system code.
  • Features:
    • Heat transfer regime from one-phase liquid flow up to post critical heat transfer and superheated steam
    • Constitutive laws for heat, mass, and momentum transferr 
      • vapour generation at heated walls
      • condensation in the subcooled liquid
      • phase slip ratio
      • pressure drop at single flow resistances and friction along the flow channels
    • IFC-67 water and steam properties
    • Various correlations for the critical heat flux 
    • Consideration of hot channels with given power peaking factors connected to selected fuel assemblies during the transient calculation
    • Coupling with mixing models for the downcomer and lower  plenum and CFD for providing detailed boundary conditions at the core inlet

Fuel Rod