Consequence analysis of a steam generator tube rupture accident

Journal Article · Sat Dec 01 00:00:00 EST 1984 · Nucl. Technol.; (United States)

OSTI ID:5924761

Wu, J M; Chuang, C F

A flashing droplet model was developed to examine the rupture flow of reactor coolant and its transport phenomena through the stream generator during a steam generator tube rupture accident. The model includes flashing flow; droplet formation; droplet removal by tube bundles, bubble scrubbing, steam separators, and steam dryers; and droplet size change by evaporation and condensation. The calculation follows the actual sequence of events during the accident. Those reactor coolant droplets escaping from the steam generator are used to estimate the radioactivity released into the environment. The steam generator tube rupture accident that occurred at the Prairie Island Plant on October 2, 1979, was studied using the model. The model estimated a release of 204 ..mu..Ci of /sup 131/I equivalent activity. The U.S. Nuclear Regulatory Commission estimated a 210-..mu..Ci release, assuming an iodine partition factor of 1/100 in the steam-generator. The model was also used to analyze a hypothetical steam generator tube rupture accident coupled with loss of off-site power in a large 1100-MW (electric) Westinghouse four-loop plant. The model estimated that 45 Ci of /sup 131/I equivalent activity could be released through the relief valves, which were stuck open for 30 min. The number is eight times higher than the estimate from the Westinghouse safety analysis report using a uniform mixing model.

• Research Organization: Rensselaer Polytechnic Institute Department of Nuclear Engineering, Troy, New York 12181
• OSTI ID: 5924761
• Journal Information: Nucl. Technol.; (United States), Vol. 67:3
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PRAIRIE ISLAND-1 REACTOR
LOSS OF COOLANT
PRAIRIE ISLAND-2 REACTOR
STEAM GENERATORS
RUPTURES
WESTINGHOUSE STANDARD REACTOR
POWER LOSSES
DROPLET MODEL
DROPLETS
EVAPORATION
IODINE 131
RADIOACTIVITY
RADIONUCLIDE MIGRATION
RELIEF VALVES
STEAM SEPARATORS
US NRC
VAPOR CONDENSATION
ACCIDENTS
BETA DECAY RADIOISOTOPES
BETA-MINUS DECAY RADIOISOTOPES
BOILERS
CONTROL EQUIPMENT
DAYS LIVING RADIOISOTOPES
ENERGY LOSSES
ENRICHED URANIUM REACTORS
ENVIRONMENTAL TRANSPORT
EQUIPMENT
FAILURES
FLOW REGULATORS
INTERMEDIATE MASS NUCLEI
IODINE ISOTOPES
ISOTOPES
LOSSES
MASS TRANSFER
MATHEMATICAL MODELS
NATIONAL ORGANIZATIONS
NUCLEAR MODELS
NUCLEI
ODD-EVEN NUCLEI
PARTICLES
PHASE TRANSFORMATIONS
POWER REACTORS
PWR TYPE REACTORS
RADIOISOTOPES
REACTOR ACCIDENTS
REACTORS
SEPARATION EQUIPMENT
US ORGANIZATIONS
VALVES
VAPOR GENERATORS
VAPOR SEPARATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled