Containment margins in FFTF for postulated failure of in-vessel post-accident heat removal

Technical Report · Wed Dec 19 00:00:00 EST 2018

DOI:https://doi.org/10.2172/7213728· OSTI ID:7213728

Bankert, S F; Hale, J P; Kunkel, W P; Martin, H C; Peak, R D; Rock, A O; Shire, P R; Simpson, D E; Stepnewski, D D

This study was performed to complete the assessment of potential melt-through consequences for the postulated failure of post-accident heat removal following a hypothetical core disruptive accident (HCDA) in the FFTF. The SPRAY, SOFIRE, CACECO, HAA-3B, and COMRADEX codes were used to determine potential site boundary radiation doses. Parametric comparisons were made to evaluate the effects of natural hydrogen-oxygen recombination, reactor cavity liner failures and containment coolers. Containment pressure relief, vent/purge options, and the effects of exhaust filtration were also included in the study. The potential site boundary radiation exposures for all options considered were within 10 CFR 100 guideline values.

• Research Organization: Hanford Engineering Development Lab., Richland, WA (United States)
• DOE Contract Number: EY-76-C-14-2170
• OSTI ID: 7213728
• Report Number(s): HEDL-TME-77-18
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FFTF REACTOR
REACTOR CORE DISRUPTION
AFTER-HEAT REMOVAL
COMPUTER CALCULATIONS
CONTAINMENT SYSTEMS
FISSION PRODUCT RELEASE
RADIATION DOSES
ACCIDENTS
CONTAINMENT
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ENGINEERED SAFETY SYSTEMS
EPITHERMAL REACTORS
FAST REACTORS
LIQUID METAL COOLED REACTORS
REACTOR ACCIDENTS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
TEST REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors