Fuel Analysis under Steady-state and Transients (FAST) is the US 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, and
cladding oxidation.
The code contains necessary material and coolant properties, as well as clad-to-coolant heattransfer
correlations, for normal operation through postulated accidents for today’s US-based light
water reactor (LWR) fuel designs. FAST-1.0 also contains preliminary materials 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-1.0, which is the first official version of this code. This document
is one of a series of documents on FAST; the other documents detail the material properties used
by FAST as well as its integral assessment to experiments and commercial data.