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ESTS0015 SCORE-EVET.

SCORE-EVET, 3-D Hydraulic Reactor Core Analysis

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1. NAME OR DESIGNATION OF PROGRAM:  SCORE-EVET.
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2. COMPUTERS
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Program name Package id Status Status date
SCORE-EVET ESTS0015/01 Arrived 23-MAY-2002

Machines used:

Package ID Orig. computer Test computer
ESTS0015/01 CDC 7600
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3. DESCRIPTION OF PROGRAM OR FUNCTION

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.
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4. 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.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

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.
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6. TYPICAL RUNNING TIME

NESC executed the three sample problems in 12, 15 and 2CP minutes, respectively, on a CDC CYBER175.
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7. UNUSUAL FEATURES OF THE PROGRAM:
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8. RELATED AND AUXILIARY PROGRAMS

An auxiliary program to convert the formatted water properties data to a binary data file for use by SCORE-EVET is included in the package. If the fuel rod thermal model is used, the fluid-solid interfacial heat transfer is calculated by  a slightly modified version of the heat transfer and critical heat flux correlations developed for RELAP4/MOD5.
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9. STATUS
Package ID Status date Status
ESTS0015/01 23-MAY-2002 Masterfiled Arrived
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10. REFERENCES

- Walter J. Wnek, John D. Ramshaw, John A. Trapp. E. Daniel Hughes,
  and Charles 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, Vol. 10, pp. 324-340 (1972).
ESTS0015/01, included references:
- 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).
- L. Eyberger:
  SCORE-EVET Tape Description and Implementation Information
  NESC 82-89 (August 14, 1982) + Attachments.
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11. MACHINE REQUIREMENTS

200,000 (octal) words of memory are required  for execution.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
ESTS0015/01 FORTRAN-IV
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:  SCOPE 2.1 (CDC7600); NOS 1.3 (CDC CYBER175).
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
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15. NAME AND ESTABLISHMENT OF AUTHORS

        Benedetti, R.L., Lords, L.V., Kiser, D.M.
   EG and G Idaho, Inc., Idaho Falls, ID (United States)
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16. MATERIAL AVAILABLE
ESTS0015/01
source program   mag tapeSCORE-EVET Source File                     SRCTP
test-case data   mag tapeSCORE-EVET Sample Problem 1                DATTP
test-case data   mag tapeSCORE-EVET Sample Problem 2                DATTP
test-case data   mag tapeSCORE-EVET Sample Problem 3                DATTP
symb data lib    mag tapeFormatted Water Properties Library         LBSTP
test-case data   mag tapeConversion Program to Create Binary Water  DATTP
test-case data   mag tapeJCL Control Information                    DATTP
test-case output listing Sample Problem 1, Cycle 600 Output         OUTLS
test-case output listing Sample Problem 2, Cycle 1000 Output        OUTLS
report                   TREE-NUREG-1133 (February 1978)            REPPT
prog. note               NESC Note 82-89 (August 14, 1982)          NOTPT
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17. CATEGORIES
  • G. Radiological Safety, Hazard and Accident Analysis

Keywords: equations of state, fluid flow, fuel rods, heat transfer, hydraulics, loss-of-coolant accident, transients.