SCORE-EVET; three-dimensional hydraulic reactor core analysis. [CDC7600,CYBER175; FORTRAN IV]

Technical Report ·

OSTI ID:6392556

Fowler, T B; Tobias, M L; Fox, J N; Lawler, B E; Koppel, J U; Triplett, J R; Lynn, L L; Waldman, L A; Goldberg, I; Greebler, P; Kelley, M D; Davis, R A; Keck, C E; Redfield, J A; Murphy,; Kiser, D M

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.CDC7600,CYBER175; FORTRAN IV; SCOPE 2.1 (CDC7600), NOS 1.3 (CDC CYBER175); 200,000 (octal) words of memory are required for execution.

• Research Organization: Los Alamos National Lab., NM (USA)
• OSTI ID: 6392556
• Report Number(s): ANL/NESC-931; ON: DE83048931
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220300* -- Nuclear Reactor Technology-- Fuel Elements
COMPUTER CODES
FLOW BLOCKAGE
FLUID FLOW
FLUID MECHANICS
FUEL ELEMENTS
FUEL RODS
HYDRAULICS
INCOMPRESSIBLE FLOW
MECHANICS
PHYSICAL PROPERTIES
REACTOR COMPONENTS
REACTORS
S CODES
THERMODYNAMIC PROPERTIES
THREE-DIMENSIONAL CALCULATIONS
WATER COOLED REACTORS