Comparison of reactivity feedback models for the FFTF passive safety tests
The FFTF Loss-of-Flow-Without-Scram Test from 50% power to natural circulation flow was analyzed with the SASSYS code using both the SASSYS reactivity feedback models and the semiempirical reactivity feedback equations for the FFTF oxide-fuel core. The experimental data for primary loop flow and reactor power were used as inputs to obtain the same fuel, sodium, and structure temperatures for both sets of reactivity feedbacks. A detailed comparison was made for each of the reactivity feedbacks: Doppler, sodium density, control rod expansion, axial fuel expansion, radial expansion, and bowing. The major differences between the SASSYS reactivity models and the FFTF reactivity equations were in the radial expansion and bowing feedback. The sensitivity of the results to the input for the SASSYS radial expansion and bowing model was investigated.
- Research Organization:
- Westinghouse Hanford Co., Richland, WA (United States)
- DOE Contract Number:
- AC06-87RL10930
- OSTI ID:
- 7013301
- Report Number(s):
- WHC-SA-0275; CONF-880506-21; ON: DE88013580
- Resource Relation:
- Conference: Safety of next generation power reactors, Seattle, WA, USA, 1 May 1988; Other Information: Portions of this document are illegible in microfiche products
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FFTF REACTOR
REACTIVITY
CONTROL ELEMENTS
DEFORMATION
EXPANSION
EXPERIMENTAL DATA
FEEDBACK
FUEL ELEMENTS
REACTOR SAFETY
S CODES
SODIUM
ALKALI METALS
COMPUTER CODES
DATA
ELEMENTS
EPITHERMAL REACTORS
FAST REACTORS
INFORMATION
LIQUID METAL COOLED REACTORS
METALS
NUMERICAL DATA
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
SODIUM COOLED REACTORS
TEST REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210500 - Power Reactors
Breeding
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors