Validation of the THIRMAL-1 melt-water interaction code
The THIRMAL-1 computer code has been used to calculate nonexplosive LWR melt-water interactions both in-vessel and ex-vessel. To support the application of the code and enhance its acceptability, THIRMAL-1 has been compared with available data from two of the ongoing FARO experiments at Ispra and two of the Corium Coolant Mixing (CCM) experiments performed at Argonne. THIRMAL-1 calculations for the FARO Scoping Test and Quenching Test 2 as well as the CCM-5 and -6 experiments were found to be in excellent agreement with the experiment results. This lends confidence to the modeling that has been incorporated in the code describing melt stream breakup due to the growth of both Kelvin-Helmholtz and large wave instabilities, the sizes of droplets formed, multiphase flow and heat transfer in the mixing zone surrounding and below the melt stream, as well as hydrogen generation due to oxidation of the melt metallic phase. As part of the analysis of the FARO tests, a mechanistic model was developed to calculate the prefragmentation as it may have occurred when melt relocated from the release vessel to the water surface and the model was compared with the relevant data from FARO.
- Research Organization:
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Organization:
- USDOE, Washington, DC (United States)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 211541
- Report Number(s):
- ANL/RE/CP-86472; CONF-950904-10; ON: DE96005219; TRN: 96:012114
- Resource Relation:
- Conference: 7. international topical meeting on nuclear reactor thermal-hydraulics (Nureth-7), Saratoga Springs, NY (United States), 10-15 Sep 1995; Other Information: PBD: May 1995
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS
99 MATHEMATICS
COMPUTERS
INFORMATION SCIENCE
MANAGEMENT
LAW
MISCELLANEOUS
T CODES
VALIDATION
CORIUM
INTERACTIONS
WATER
BWR TYPE REACTORS
MELTDOWN
PWR TYPE REACTORS
COMPUTER CALCULATIONS
REACTOR SAFETY
HEAT TRANSFER
HYDRAULICS