RSICC CODE PACKAGE PSR-442

1. NAME AND TITLE

SCORE-EVET: Code System for Three Dimensional Hydraulic Reactor Core Analysis.

2. CONTRIBUTORS

EG&G Idaho, Idaho Falls, ID through the Energy Science and Technology Software Center, Oak Ridge, Tennessee.

3. CODING LANGUAGE AND COMPUTER

SCORE-EVET was developed using Fortran IV on a CDC 7600 (P00442C760000).

4. NATURE OF PROBLEM SOLVED

SCORE-EVET was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code contains a one-dimensional steady state solution scheme to initialize the flow field, steady state and transient fuel rod conduction models, and comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions, such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage. The basic volume-averaged transient three-dimensional equations for flow in porous media are solved in their general form with constitutive relationships and boundary conditions tailored to define the porous medium as a matrix of fuel rods. By retaining generality in the form of the conservation equations, a wide range of fluid flow problem configurations, from computational regions representing a single fuel rod subchannel to multichannels, or even regions without a fuel rod, can be modeled without restrictive assumptions. The completeness of the conservation equations has allowed SCORE-EVET to be used, with modification to the constitutive relationships, to calculate three-dimensional laminar boundary layer development, flow fields in large bodies of water, and, with the addition of a turbulence model, turbulent flow in pipe expansions and tees.

5. METHOD OF SOLUTION

The numerical technique used to solve the volume-averaged three-dimensional equations is based on the marker and cell (MAC) method for incompressible flow, as modified by Hirt and Cook, and the implicit continuous fluid Eulerian (ICE) method for compressible flow. The one-dimensional transient heat conduction equation used in the fuel rod thermal transport model is solved by dividing the fuel and cladding into a finite number of nodes and applying a Crank-Nicolson implicit finite difference formulation.

6. RESTRICTIONS OR LIMITATIONS

Maxima of 64 flow channels, 40 cells in any one direction (boundary plus real cells), and 40 possible loss coefficients. For the fuel rod thermal transport model, five nodes are allowed in the fuel and two nodes in the cladding.

7. TYPICAL RUNNING TIME

NESC executed the three sample problems in 12, 15, and 2 CP minutes, respectively, on a CDC CYBER175.

8. COMPUTER HARDWARE REQUIREMENTS

200,000 (octal) words of memory are required for execution on CDC 7600 or Cyber computers.

9. COMPUTER SOFTWARE REQUIREMENTS

SCORE-EVET was run under SCOPE 2.1 on CDC7600 and under NOS 1.3 on CDC CYBER175. A Fortran IV compiler is required. The software was tested in February 1981 when the package was contributed to NESC. No testing or modifications were made when the package was transferred to RSICC and made available in March 2001.

10. REFERENCES

a) Included in documentation:

L. Eyberger, "SCORE-EVET, NESC No. 931, SCORE-EVET Tape Description and Implementation Information," NESC Note 82-89 (August 14, 1982).

R. L. Benedetti, L. V. Lords, and D. M. Kiser, "SCORE-EVET: A Computer Code for the Multidimensional Transient Thermal-Hydraulic Analysis of Nuclear Fuel Rod Arrays," TREE-NUREG-1133 (February 1978).

b) Background information:

W. J. Wnek, J. D. Ramshaw, J. A. Trapp, E. D. Hughes, and C. W. Solbrig, "Transient Three-Dimensional Thermal-Hydraulic Analysis of Nuclear Reactor Fuel Rod Arrays: General Equations and Numerical Scheme," ANCR-1207 (November 1975).

C. W. Hirt and J. L. Cook, "Calculating Three-Dimensional Flows around Structures and over Rough Terrain," Journal of Computational Physics, 10, 324-340 (1972).

11. CONTENTS OF CODE PACKAGE

Included with the package are the referenced documents in (10.a) and one 3.5" diskette which includes the Fortran source, data library and test case input written in a self-extracting compressed DOS file.

12. DATE OF ABSTRACT

March 2001.

KEYWORDS: COMPLEX GEOMETRY; HEAT TRANSFER; LOCA; THERMAL HYDRAULICS