Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission
(NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor
fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design
basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure,
and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary
conditions. The phenomena modeled by the code include:
• heat conduction through the fuel and other materials
• heat transfer from the cladding-to-coolant
• cladding elastic and plastic deformation, including creep
• fuel-cladding mechanical interaction
• fission gas release from the fuel
• rod internal pressure and void volume
• cladding oxidation
The code contains necessary material and coolant properties, as well as clad-to-coolant heat-
transfer correlations, for normal operation through postulated accidents and AOOs for today’s
U.S.-based light water reactor (LWR) fuel designs. FAST-1.2 also contains preliminary materi- als
and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel
concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on
Windows and Linux operating systems.
This document describes FAST-1.2 and is one of a series of documents on the code; the other
documents detail the material properties used by FAST as well as its integral assessment to ex-
periments and commercial data.