last modified: 01-FEB-1979 | catalog | categories | new | search |

NEA-0498 SCORCH-B2.

SCORCH-B2, BWR Core Heating During LOCA

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1. NAME OR DESIGNATION OF PROGRAM:  SCORCH-B2.
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2. COMPUTERS
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Program name Package id Status Status date
SCORCH-B2 NEA-0498/01 Tested 01-FEB-1979

Machines used:

Package ID Orig. computer Test computer
NEA-0498/01 IBM 370 series IBM 370 series
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3. NATURE OF PHYSICAL PROBLEM SOLVED

SCORCH-B2 is a simulation code of reactor core heatup during a LOCA of BWR's. The program analyzes transient heat transmission on a horizontal plane of a fuel assembly and evaluates the peak cladding temperature and the maximum oxide thickness, both of which determine the soundness of the core during  the accident. Fuel rods are arbitrarily classified into a smaller number of groups and each fuel rod is divided into several annuli.
Heat conduction within fuel rods, heat convection from rods to coolant and radiation among rods and the channel box are calculated  for each time step.
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4. METHOD OF SOLUTION

Forward explicit differential method is used to  solve the heat transmission equations and the shadow area method which was newly developed is used to calculate the radiation view factors between ballooning rods.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

The maximum number of columns (= number of rows) of the fuel rods in a fuel assembly is 8. The maximum number of radial nodes in a fuel pellet is 10. Any grouping of rods in an assembly is applicable.
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6. TYPICAL RUNNING TIME

The samle run (10 fuel rod group, 7 pellet radial node, 220 sec physical time) takes 35 seconds on FACOM 230/75.
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7. UNUSUAL FEATURES OF THE PROGRAM

Radiation view factors are analytically recalculated whenever cladding are calculated to balloon.
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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-0498/01 01-FEB-1979 Tested at NEADB
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10. REFERENCES
NEA-0498/01, included references:
- K. Abe and K. Sato:
  SCORCH-B2: Simulation Code of Reactor Core Heatup During LOCA
  (for BWR, 2nd version) JAERI-M 6678 (July 28, 1976)
  (japanese+translation).
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11. HARDWARE REQUIREMENTS: MACHINE REQUIREMENTS
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0498/01 FORTRAN-IV
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13. OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED:  FACOM-230/75, monitor 6/7.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS
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15. NAME AND ESTABLISHMENT OF AUTHOR

       K. Abe and K. Sato
       Tokai Research Establishment
       Japan Atomic Energy Research Institute
       Tokai-Mura, Naka-Gun, Ibaraki-Ken
       JAPAN
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16. MATERIAL AVAILABLE
NEA-0498/01
File name File description Records
NEA0498_01.001 SOURCE PROGRAM (F4,EBCDIC) 2181
NEA0498_01.002 SAMPLE PROBLEM INPUT DATA 126
NEA0498_01.003 SAMPLE PROBLEM PRINTED OUTPUT 8489
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17. CATEGORIES
  • G. Radiological Safety, Hazard and Accident Analysis

Keywords: BWR reactors, cladding, fuel assemblies, fuel rods, fuel-cladding interaction, heat transfer, loss-of-coolant accident, reactor kinetics, reactor safety, simulation, transients.