TRAC-PF1-EN/MOD3: Code System for Coupled 3D Neutronics-Thermalhydraulics Calculations.

Synthesis Srl, Milano, Italy, through the NEA Data Bank, Issy-les-Moulineaux, France.

Fortran 77; PC X86 (P00477PCX8601).

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 one 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 unborated water), involving all of the primary system individual components.

The TRAC two-phase, two-fluid nonequilibrium hydrodynamics model with an incondensable 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, _{} *,

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 coworkers at MIT and coded in the QUANDRY program. 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 homogenization 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) substeps is applied.

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 2.7210^{6} and
410^{6} bytes) can be changed by modifying some PARAMETER statements.

The transient of the first sample problem (rod ejection at Hot Zero-Power) featured by a mesh of over 4400 neutronic nodes in a quarter-core and 23 plant components including a full vessel divided into 384 thermal-hydraulic nodes requires 170 min (2.8 h) of CP time on a PC-486/100 for a sequence of 547 thermal-hydraulic (TRAC) steps and 2066 neutronic (QUANDF) steps. Notice that the neutronic calculations account for 85% of the total CPU time. Such a disproportion is explained not only by the greater number of neutronic steps but, mainly, by the fact that the neutronic analysis is carried out on an optimal mesh of assembly-sized (20 cm) nodes while the thermal-hydraulic calculation is performed on a tentative mesh of much larger nodes.

A PC with 486 or Pentium processor and at least 16 Mb of RAM.

WINDOWS provided with MS Fortran Power Station Compiler version 1.0 or higher. LINUX or Unix with a F77 compiler.

G. Alloggio, D. Basile, E. Brega, R. Guandalini, and L.
Pollachini, “Simulation of PWR Plant by a New Version of TRAC-PF1 Code
Including a Three-Dimensional Neutronic Model and a Transport Boron Model,” pp
1019-1027 in *Proceedings of the ASME-JSME 4 ^{th} International
Conference on Nuclear Engineering*, A. S. Rao, R. B. Duffey, and D. Ilias,
editors (No.I389A2-1996).

D. Basile, E. Salina, and E. Brega, “TRAC-PF1-EN/MOD3: A TRAC-PF1 Revised Version Inclusive of a Three-Dimensional Neutron Kinetics Model Based on High-Accuracy Two-Group Nodal Diffusion Methods,” Job No. 1037/96, Rep. No. 1037/3 (February 28, 1997).

**b) Background information:**

D.R. Liles *et al.*, “TRAC-PF1, An Advanced
Best-Estimate Computer Program for Pressurised Water Reactor Analysis. Input
Specifications TRAC-PF1 7.0/EXTUP 7.6,” LA-TIA-TN-82-1 (June 1982).

Included on one CD-ROM compressed in a self-extracting executable file are the referenced documents, Fortran source files, and the input/output files for the three sample problems.

March 2000, March 2002, October 2010.

**KEYWORDS:** LWR; REACTOR SAFETY; THERMAL
HYDRAULICS; HEAT TRANSFER