TREAT source-term experiment STEP-1 simulating a PWR LOCA
In a hypothetical pressurized water reactor (PWR) large-break loss-of-coolant accident (LOCA) in which the emergency core cooling system fails, fission product decay heating causes water boil-off and reduced heat removal. Zircaloy cladding is oxidized by the steam. The noble gases and volatile fission products such as cesium and iodine that constitute a principal part of the source term will be released from the damaged fuel at or shortly after the time of cladding failure. TREAT test STEP-1 simulated the LOCA environment when the volatile fission products would be released using four fuel elements from the Belgonucleaire BR3 reactor. The principal objective was to collect a portion of the releases carried by the flow stream in a region as close as possible to the test zone. In this paper, the test is described and the results of an analysis of the thermal and steam/hydrogen environment are compared with the test measurements in order to provide a characterization for analysis of fission product releases and aerosol formation. The results of extensive sample examinations are reported separately.
- Research Organization:
- Argonne National Lab., IL
- OSTI ID:
- 7124346
- Report Number(s):
- CONF-861102-
- Journal Information:
- Trans. Am. Nucl. Soc.; (United States), Vol. 53; Conference: American Nuclear Society and Atomic Industrial Forum joint meeting, Washington, DC, USA, 16 Nov 1986
- Country of Publication:
- United States
- Language:
- English
Similar Records
Source-term experiment STEP-3 simulating a PWR severe station blackout
Summary results of the treat source term experiments project (STEP)
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
FISSION PRODUCT RELEASE
PWR TYPE REACTORS
REACTOR SAFETY
AEROSOLS
AFTER-HEAT REMOVAL
COMPUTERIZED SIMULATION
ECCS
FISSION PRODUCTS
FUEL CANS
FUEL ELEMENT FAILURE
HEAT TRANSFER
OXIDATION
RADIOACTIVITY TRANSPORT
REACTOR ACCIDENTS
TREAT REACTOR
ZIRCALOY
ACCIDENTS
AIR COOLED REACTORS
ALLOYS
CHEMICAL REACTIONS
COLLOIDS
DISPERSIONS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
EXPERIMENTAL REACTORS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HOMOGENEOUS REACTORS
ISOTOPES
MATERIALS
RADIOACTIVE MATERIALS
REACTOR PROTECTION SYSTEMS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
SAFETY
SIMULATION
SOLID HOMOGENEOUS REACTORS
SOLS
TEST REACTORS
THERMAL REACTORS
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors