CONTAIN code calculations for the LA-4 experiment
The CONTAIN code is a system-level analysis tool developed for the USNRC, and is intended for best-estimate prediction of conditions which might occur in the containment building of a nuclear power plant during a severe accident. A key feature of the code is that it models the containment phenomena in an integrated manner. In particular, the CONTAIN code models some of the complex ways that thermal hydraulics and aerosol phenomena interact with each other. The Light Water Reactor Aerosol Containment Experiment (LACE) progarm is a program to aid researchers in their understanding of thermal hydraulic and aerosol behavior within containments. The purpose of this paper is to report on best-estimate LA-4 post-test calculations that have been completed with the most recent version of the CONTAIN code, version 1.11. An analysis of experimental data and review of the blind post-test CONTAIN calculations is used to justify a re-calculation of the experiment and to establish a best-estimate calculation. The best-estimate calculation shows that reasonably good agreement between thermal hydraulic predictions and data can be obtained with the current CONTAIN models by varying experimental parameters within their uncertainties. Furthermore, with the recently added solubility model for hygroscopic aerosols, the best-estimate calculation gives aerosol behavior that is in good agreement with aerosol data. 10 refs., 16 figs.
- Research Organization:
- Sandia National Labs., Albuquerque, NM (USA)
- Sponsoring Organization:
- NRC
- DOE Contract Number:
- AC04-76DP00789
- OSTI ID:
- 6675130
- Report Number(s):
- SAND-89-3010C; CONF-901024--4; ON: DE90012753
- Country of Publication:
- United States
- Language:
- English
Similar Records
Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2
MELCOR 1. 8. 1 assessment: LACE aerosol experiment LA4
Related Subjects
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCIDENTS
AEROSOLS
BUILDINGS
BWR TYPE REACTORS
C CODES
COLLOIDS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTAINMENT
CONTAINMENT BUILDINGS
DISPERSIONS
ENERGY TRANSFER
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
MECHANICS
NUCLEAR FACILITIES
NUCLEAR POWER PLANTS
PARTICLE SIZE
POWER PLANTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
SIZE
SOLS
STEAM
TEST FACILITIES
THERMAL POWER PLANTS
VAPOR CONDENSATION
WATER COOLED REACTORS
WATER MODERATED REACTORS