FFTF containment of hypothetical accidents

Conference · Wed Aug 01 04:00:00 EDT 1979

OSTI ID:5861341

Peak, R. D.; Martin, H. C.; Jensen, R. L.; Stepnewski, D. D.; Hale, J. P.

The FFTF facility was evaluated for the consequences of an HCDA followed by failure of in-vessel post-accident heat removal, reactor vessel melt-through, and release of core debris and sodium coolant to the reactor cavity. Two cases are presented based on parameters considered to represent upper limits for rates of chemical and thermal attack of the reactor cavity concrete containment structure. The reactor containment building temperature, pressure, and leak rate histories were computed with the CACECO code which provided input into the HAA-3C code for prediction of aerosol behavior, and to the COMRADEX-H code for prediction of radioactivity dispersion. The resultant 30-day doses at the site boundary were judged to be acceptable considering the conservatism in the analysis and the low probability of the event.

• Research Organization: Hanford Engineering Development Lab., Richland, WA (USA)
• Sponsoring Organization: USDOE
• DOE Contract Number: EY-76-C-14-2170
• OSTI ID: 5861341
• Report Number(s): HEDL-SA-1736-FP; CONF-790816-70
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CALCULATIONS
CONTAINMENT
CONTAINMENT BUILDINGS
CONTAINMENT SYSTEMS
ENGINEERED SAFETY SYSTEMS
EPITHERMAL REACTORS
FAST REACTORS
FFTF REACTOR
LIQUID METAL COOLED REACTORS
MATHEMATICAL MODELS
PERFORMANCE
REACTOR ACCIDENTS
REACTOR CORE DISRUPTION
REACTOR SAFETY
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
SODIUM COOLED REACTORS
STRESSES
TEST REACTORS
THERMAL STRESSES