THATCH: A Computer Code for Modelling Thermal Networks of High-Temperature Gas-Cooled Nuclear Reactors

摘要

The report documents the THATCH code (Thermal Analysis for Temperatures in the Core of an HTGR), which models thermal and flow networks of solids and coolant channels in two-dimensional geometries. The main application of THATCH is for reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The code simulates core heatup transients, heat transfer to general sinks or to specific air or water-cooled reactor cavity cooling systems. Graphite oxidation during air or water ingress can be modelled, including the chemical energy release. A point kinetics model is available for reactivity excursions. For most slow HTGR transients a user-selected nodalization of the core in r-z geometry is used. A separate model of heat transfer in the symmetry element of each fuel element is available for rapid transients. The report describes the mathematical models and the method of solution. It describes the code structure and its various procedures. Details of the input data and file usage, is given for the code and for the preprocessing and postprocessing options. The THATCH model of the current 350 MW (sub th) reactor is described. Input data for four sample cases are given, with output available in fiche form. Installation requirements, code limitations, and some common error indications are listed.