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NEA-1593 TRAC-PF1/EN MOD 3.

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

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1. NAME OR DESIGNATION OF PROGRAM:  TRAC-PF1/EN MOD 3.
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2. COMPUTERS
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Program name Package id Status Status date
TRAC-PF1/EN MOD 3 NEA-1593/02 Tested 14-MAR-2002

Machines used:

Package ID Orig. computer Test computer
NEA-1593/02 IBM PC PC Pentium III,DEC ALPHA W.S.
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3. 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 unborated water), involving all of the primary system individual components.
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4. METHODS

The TRAC two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable 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 coworkers 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) substeps is applied.
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5. 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).
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6. TYPICAL RUNNING TIME

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 a optimal mesh of assembly-sized (*20 cm) nodes while the thermal-hydraulic calculation is performed on a tentative mesh of much larger nodes.
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7. UNUSUAL FEATURES
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8. RELATED OR AUXILIARY PROGRAMS

The thermal-hydraulic section is a revised version of the TRAC-PF1 code (IBM version) which is also available from the NEA Data Bank as package NESC 0836/07. Since neither the physical and mathematical models of TRAC nor the stream of the input data have been changed except, possibly, for the inclusion of those data activating the Boron transport model, the User's Manual of the original TRAC-PF1 code (see 10. REFERENCES) 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 viceversa 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 viceversa 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.
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9. STATUS
Package ID Status date Status
NEA-1593/02 14-MAR-2002 Tested at NEADB
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10. REFERENCES

- 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, Los Alamos National Laboratory,
  Safety Code Development Group, June 1982
NEA-1593/02, included references:
- 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
(ENEL-ATN/GNUM) Rep. No. 1037/3 (28 Feb. 97)
- 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
Reprinted from Proceedings of the ASME-JSME 4th Int. Conf. on Nucl. Eng.
Book No.I389A2-1996
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11. HARDWARE REQUIREMENTS

A Personal Computer with 486 or Pentium processor and at least 16 Mb of RAM.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-1593/02 FORTRAN-77
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13. SOFTWARE REQUIREMENTS

DOS or WINDOWS provided with MS FORTRAN Power Station Compiler version 1.0 or higher.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS
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15. NAME AND ESTABLISHMENT OF AUTHORS

D. Basile, E. Salina
Synthesis Srl
Via B. Garofalo 10
20133 Milano, Italy

E. Brega
ENEL SpA
Via Pozzobonelli, 6
20162 Milano, Italy
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16. MATERIAL AVAILABLE
NEA-1593/02
Quandf.for FORTRAN source file
Quandf.doc Code domumentation
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17. CATEGORIES
  • F. Space - Time Kinetics, Coupled Neutronics - Hydrodynamics - Thermodynamics
  • G. Radiological Safety, Hazard and Accident Analysis

Keywords: LWR reactors, boron, fuel rods, heat transfer, neutron flux, power distribution, power plants, reactivity, reactor cores, reactor kinetics, reactor safety, thermodynamics, three-dimensional, transients, two-group theory, two-phase flow.