The FRAPTRAN-1.3 computer code has been evaluated regarding its applicability, modelling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under reactor power and coolant transients, such as overpower transients, reactivity initiated accidents (RIA), boiling-water reactor power oscillations without scram, and loss of coolant accidents (LOCA). Its experimental database covers boiling- and pressurized water reactor fuel rods with UO2 fuel up to rod burnups around 64 MWd/kgU.
In FRAPTRAN-1.3, the fundamental equations for heat transfer and structural analysis are solved in one-dimensional (in the radial direction) and transient (time-dependent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow, rod internal gas pressure and optionally axial flow of fission gases. The clad-to-coolant heat transfer conditions can either be specified as pre-calculated data or can be determined by a coolant channel model in the code. The code provides different clad rupture models depending on cladding temperature and amount of cladding plastic hoop strain. For LOCA analysis, a model calculating local clad shape (ballooning) and associated local stresses is available to predict clad burst. A strain- based failure model is present for cladding rupture driven by pellet-cladding mechanical interaction. Two models exist for computation of high-temperature clad oxidation under LOCA (i) the Baker-Just model for licensing calculations and (ii) the Cathcart-Pawel model for best-estimate calculations.
The code appears to be fairly easy to use, however, the applicability of the current version as a self-standing analysis tool for LOCA and RIA analyses depends highly on the numerical robustness of the coolant channel model for generation of clad-to-coolant heat transfer boundary conditions.