An Integrity Assessment Methodology for the VHTR Graphite Structure

Abstract

The VHTR, one of the next generations of nuclear power plants, uses graphite as fuel block, moderator, and support post. To decompose water into hydrogen and to improve thermal efficiency, the VHTR operates under conditions of high temperature above 950 degrees C. In case of the Prismatic Modular Reactor 600, one type of the Korean VHTR, graphite is used as a fuel block and support post inside the reactor core. It can be exposed to high temperature and neutron irradiation. Therefore it is necessary to consider its integrity by estimating its safety under this condition. In order to evaluate the integrity of the graphite structure, a new assessment methodology has been developed and applied. The proposed model turned out to be appropriate to assess graphite integrity. The safety of VHTR can be improved by the quantification of the graphite structure integrity and a more reliable design.