MELCOR 1. 8. 1 assessment: ACRR source term experiments ST-1/ST-2
The MELCOR code has been used to simulate the ST-1 and ST-2 in-pile product source term experiments performed in the ACRR facility. As expected, there were no major differences observed in the results calculated for the different test conditions. The CORSOR, CORSOR-M and CORSOR-Booth release models all were tested, and the effect of including the surface-volume correction term was evaluated. MELCOR results were compared to test data and to VICTORIA results, and also directly to the correlations and to ST-1/ST-2 results predicted by Battelle using their stand-alone CORSOR code to verify that the models have been implemented correctly in MELCOR. The release rates and total release fractions calculated by MELCOR generally agreed well with the test data, for both volatile and refractory species, with none of the release model options available yielding consistently better agreement with data for species. Sensitivity studies checking for time step and noding effects and machine dependencies were done, and some machine dependencies associated with very small numbers were identified and corrected in the code. Additional sensitivity studies were run on parameters affecting core heatup and core damage, including both variations in code models such as convective heat transfer coefficients, radiation view factors, candling assumptions, and in experimental conditions such as pressures, flow rates, power levels, and insulation thermal conductivity. Code and user input modeling errors encountered in these analyses are described.
- Research Organization:
- Sandia National Labs., Albuquerque, NM (United States)
- Sponsoring Organization:
- USDOE; USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC04-76DP00789
- OSTI ID:
- 5282657
- Report Number(s):
- SAND-91-2833; ON: DE92013782
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
FISSION PRODUCT RELEASE
M CODES
SOURCE TERMS
AEROSOLS
C CODES
COMPUTERIZED SIMULATION
EXPERIMENTAL DATA
FISSION PRODUCTS
FUEL CANS
FUEL ELEMENTS
FUEL-CLADDING INTERACTIONS
HEAT TRANSFER
NUMERICAL ANALYSIS
SENSITIVITY ANALYSIS
COLLOIDS
COMPUTER CODES
DATA
DISPERSIONS
ENERGY TRANSFER
INFORMATION
ISOTOPES
MATERIALS
MATHEMATICS
NUMERICAL DATA
RADIOACTIVE MATERIALS
REACTOR COMPONENTS
SIMULATION
SOLS
220502* - Nuclear Reactor Technology- Environmental Aspects- Radioactive Effluents
220900 - Nuclear Reactor Technology- Reactor Safety
990200 - Mathematics & Computers