TRAC-PF1/EN MOD 3, Best Estimate Coupled 3-D Neutronics-Thermohydraulics

1 - Description of program or function: TRAC-PF1-EN/MOD3 is a combined computer program comprising a revised version of the TRAC-PF1 transient reactor analysis code and a specially implemented three-dimensional two-group neutron kinetics code (QUANDF). Starting from either a critical steady-state (k-effective or critical dilute Boron problem) or a subcritical steady-state (fixed source problem) in a PWR plant, the code allows to simulate the neutronic and thermal-hydraulic core transient response to reactivity accidents initiated both inside the vessel (such as a control rod ejection) and outside the vessel (such as the sudden circulation of a stagnant slug of un-borated water), involving all of the primary system individual components. 2 - Methods: The TRAC two-phase, two-fluid non-equilibrium hydrodynamics model with a non-condensable gas field is based on six partial differential equations that describe the transfer of mass, energy and momentum between the water liquid and vapor phases and the interaction of the individual phases with the system structures. Because these interactions are dependent on the flow topology, a flow-regime dependent constitutive equation package has been incorporated into the code. The one-dimensional (z) or three-dimensional (r, *, z) fluid dynamics equations as well as the one-dimensional (r or z) or two-dimensional (r, z) equations that model the heat transfer in solid structures are approximated by finite differences. The fluid dynamics equations in the one-dimensional components use a multistep procedure that allows the material Courant condition to be violated. The optional three-dimensional component (vessel) uses a semi-implicit scheme, subject to the Courant condition, The finite-difference equations for hydrodynamic phenomena form a system of coupled nonlinear equations that are solved by a Newton-Raphson iteration procedure. The heat-transfer equations are treated implicitly in the radial direction and explicitly in the axial direction. The neutronic module is based on the Analytical Nodal Method (ANM) for two-group neutron diffusion equation in three-dimensional cartesian geometry, developed by A. F. Henry and his co-workers at MIT and coded in the QUANDRY program, but, instead of solving the nodal equations for node-averaged fluxes and directional leakages, it adopts the more efficient approach of solving Coarse-Mesh Finite-Difference (CMFD) equations corrected by Equivalence Theory Discontinuity Factors which are internally computed so as to match just the accuracy of the Analytical Nodal Method. The cross-sections and the discontinuity factors correcting for homogenisation effects are updated for thermal (fuel temperature) and thermal-hydraulic feedback (coolant temperature and density) and dilute Boron effect, either by applying temperature and density coefficients (quadratic at the most) or by interpolating in input multiple-entry libraries of reference values. At each thermal-hydraulic (TRAC) time step whose size is automatically selected by inhibitive and promotional algorithms between input minimum and maximum values, the coefficients of the neutronic nodal equations are recomputed and a refined logic to control also the neutronic (QUANDF) sub-steps is applied. 3 - Restrictions on the complexity of the problem: Most of the data-dependent arrays are contained in two named Common blocks, viz., BLANK for the thermal-hydraulic section and BLANKQ for the neutronic section, whose standard lengths (respectively 2.72*106 and 4*106 bytes) can be changed by modifying some PARAMETER statements (see the Installation Directions). original TRAC-PF1 code as well as the additional documentation here enclosed are required to use the TRAC/QUANDF combined code. To mention the main revisions made in TRAC-PF1, the coding needed to include the neutronic module as a subprogram and to convert the neutronic and thermal-hydraulic field variables from the generally fine cartesian mesh of QUANDF to the generally coarser cylindrical mesh of TRAC and vice versa has been implemented, the dilute Boron transport model has been activated, the continuous rolling of data from the Small Core Memory (SCM) to the Large Core Memory (LCM) and vice versa as the solution procedure advances on the nodal grid, has been eliminated, the logic to control the automatic selection of the time steps has been refined and a control on the fission power distribution has been included in the PWR and Generalised Initialisation procedures