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NEA-0235 FLARE-JAERI. SCOPERS

FLARE-JAERI, 3-D BWR and ATR Simulation
SCOPERS-2, BWR and PWR Core Performance Simulation

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1. NAME OR DESIGNATION OF PROGRAM:  FLARE-JAERI, SCOPERS-2.
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2. COMPUTERS
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Program name Package id Status Status date
SCOPERS-2 NEA-0235/02 Tested 08-NOV-1988

Machines used:

Package ID Orig. computer Test computer
NEA-0235/02 FACOM M-380 IBM 3090
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3. NATURE OF PHYSICAL PROBLEM SOLVED

(A) FLARE-JAERI: The three- dimensional boiling water reactor simulator FLARE is revised for the design of ATR and further analysis of BWR. Following modifications are made:
1. A control rod positioning logic is now in the program.
2. The direction of coolant flow is optional, namely upward or down-    ward or channelwise alternatively.
3. The effect of self-evaporation is considered.
4. An additional variety of fitting formula is available.
5. Vertical fuel shuffling or reloading is performed by users    option after the core reactivity is less than an input value.
6. The horizontal fuel shuffling is easily performed within one run.
(B) SCOPERS-2: This new version of FLARE-JAERI uses a generalized FLARE-type nodal equation together with the nodal-form neutron migration kernel derived in a relatively rigorous theoretical way. It replaces the empirical kernel used in FLARE. SCOPERS-2 can be applied to both PWR and BWR with wider options.
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4. METHOD OF SOLUTION

Basic structure of FLARE is reserved. The neutron balance equation using migration area and albedo is solved by either point SOR or point Jacobi or Gauss-Seidel iterative method. The control rod pattern - positions and driving ratio of the adjusting rods - is an input. Rod positions are searched by using parabolic interpolation or extrapolation. Rod positioning is discontinued when at least one of the adjusting rods is out of the driving range.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM.
One-energy-group approximation
Number of fuel types LE 13.
Number of nodes in x-direction LE 15.
Number of nodes in y-direction LE 15.
Number of nodes in z-direction LE 20.
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6. TYPICAL RUNNING TIME

Typical CPU time on FACOM M-380 is 8.6 seconds for a steady-state solution of Nuclear Ship Mutsu problem with 432 nodes.
NEA 0235/02: The test cases included in this package were run by NEA-DB on an IBM 3090 computer. Execution times were 1.40 seconds (case 1); 5 seconds (case 2).
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7. UNUSUAL FEATURES: UNUSUAL FEATURES OF THE PROGRAM. None.
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8. RELATED OR AUXILIARY PROGRAMS: RELATED AND AUXILIARY PROGRAMS.
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9. STATUS
Package ID Status date Status
NEA-0235/02 08-NOV-1988 Tested at NEADB
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10. REFERENCES

- D.E. Delpe, D.L. Fisher, J.M. Harriman and M.J. Stenwell
  FLARE A Three Dimensional Boiling Water Reactor Simulator
  GEAP-4598 (1964).
NEA-0235/02, included references:
- T. Shimooke and M. Itagaki:
  A User's Manual for SCOPERS-2: A Static Core Performance Simulator
  for Light Water Reactors [Version 2].
  JAERI-M 86-063  (March 1986)
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11. MACHINE REQUIREMENTS

NEA 0235/02: 612K bytes of IBM 3090 main storage were required to run the test cases.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0235/02 FORTRAN-IV
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13. OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED:  FACOM
OSN/F4.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
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15. NAME AND ESTABLISHMENT OF AUTHOR

             Takanori SHIMOOKE
             Department of Reactor Safety Engineering
             Japan Atomic Energy Research Institute
             Tokai-mura, Naka-gun, Ibaraki-ken, 319,11, Japan

             Masafumi ITAGAKI
             Department of Nuclear Ship Engineering
             Japan Atomic Energy Research Institute
             Toranomon 1-151-16, Minato-ku, Tokyo 105, Japan
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16. MATERIAL AVAILABLE
NEA-0235/02
File name File description Records
NEA0235_02.001 Information file 133
NEA0235_02.002 FORTRAN source without sub. MOVEC 5329
NEA0235_02.003 Additional FORTRAN source by NEA 15
NEA0235_02.004 JCL from the author 15
NEA0235_02.005 JCL used at NEA Data Bank 70
NEA0235_02.006 Sample case 1 (JPDR) 69
NEA0235_02.007 Sample case 2 (MUTSU) 53
NEA0235_02.008 Output of sample case 1 from the author 1120
NEA0235_02.009 Output of sample case 2 from the author 876
NEA0235_02.010 Output of sample case 1 by NEA Data Bank 1094
NEA0235_02.011 Output of sample case 2 by NEA Data Bank 876
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17. CATEGORIES
  • D. Depletion, Fuel Management, Cost Analysis, and Power Plant Economics

Keywords: BWR reactors, control elements, coolants, fuel management, pwr reactors, simulation, three-dimensional.