1. NAME AND TITLE
SCORCH-B2: BWR Core Heating During LOCA.
Atomic Energy Research Institute, Tokai-Mura, Naka-Gun, Ibaraki-Ken, Japan through
the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France.
3. CODING LANGUAGE AND COMPUTER
Fortran, IBM 370 Series (P00601I037000).
4. NATURE OF 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. Radiation view factors are analytically recalculated whenever cladding are calculated to balloon.
5. 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.
6. RESTRICTIONS OR LIMITATIONS
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.
7. TYPICAL RUNNING TIME
The sample run (10 fuel rod group, 7 pellet radial node, 220 sec physical time) takes 35 seconds on FACOM 230/75.
8. COMPUTER HARDWARE REQUIREMENTS
9. COMPUTER SOFTWARE REQUIREMENTS
FACOM-230/75, monitor 6/7.
a) Included Documentation:
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).
11. CONTENTS OF CODE PACKAGE
The package is distributed on a CD with a compressed zip file including source files, documentation, sample input and output.
12. DATE OF ABSTRACT
KEYWORDS: BWR REACTORS, CLADDING, FUEL ASSEMBLIES, FUEL RODS, FUEL-CLADDING INTERACTION, HEAT TRANSFER, LOSS-OF-COOLANT ACCIDENT, REACTOR KINETICS, REACTOR SAFETY, SIMULATION, TRANSIENTS