Preliminary Analysis of FFTF Loss-of-Flow Tests with SAS4A/SASSYS-1

Journal Article · Sat Jul 01 04:00:00 EDT 2017 · Transactions of the American Nuclear Society

OSTI ID:23050256

Zhang, Guanheng; Moisseytsev, Anton; Sumner, Tyler; Fanning, Thomas ^[1]

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1. Argonne National Laboratory: 9700 Cass Ave, Lemont, IL 60439 (United States)

The Fast Flux Test Facility (FFTF) was the most recent Sodium-cooled Fast Reactor (SFR) built in the United States. The Passive Safety Test (PST) program was carried out on FFTF between 1984 and 1986 to demonstrate the capability of passive reactor shutdown and decay heat removal in response to protected and unprotected transients. In order to recover the information and knowledge that has been gained from FFTF, a research project was initiated focusing on the simulation of Unprotected Loss of Flow Without Scram (LOFWOS) tests similar to the knowledge preservation activities conducted for EBR-II by Argonne National Laboratory in earlier years. These knowledge preservation activities are intended to support the validation of computational tools and methods in the area of SFR passive safety. Comparisons with experimental data and other safety codes provide opportunities to improve SFR design and safety analysis codes and methods.

• OSTI ID: 23050256
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 116; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING
AFTER-HEAT REMOVAL
ANL
COMPUTERIZED SIMULATION
EBR-2 REACTOR
FFTF REACTOR
KNOWLEDGE PRESERVATION
LOSS OF FLOW
REACTOR DESIGN
SAFETY ANALYSIS
SCRAM
TRANSIENTS
VALIDATION