Nuclear Graphite Irradiated Material Property Normalization, Interpolation, and Extrapolation Approach

Abstract

Abstract Nuclear grade graphite has several physical characteristics that are advantageous for its use in high temperature reactors, including a high scatter cross-section and a low absorption cross-section (high moderating ratio). In addition to the neutronics benefits of graphite as a moderator, nuclear grade graphite has high thermal conductivity combined with adequate structural strength at high temperatures. However, the material properties associated with graphite grades that are candidates for high temperature environments exhibit a strong dependance on fast neutron fluence (E > 0.1 MeV). The material properties of interest are typically available only at discrete temperature and fluence levels, which results in an expensive process to characterize the material properties across a wide range of operating conditions. In addition, this approach may limit the reactor design to operating conditions and parameters established early in the design process. The purpose of this paper is to outline a technique to interpolate available irradiated material property data, within the bounds of the parameters examined in applicable testing programs, and also extrapolate to irradiation temperatures beyond those provided in the available test reports. Implementing this approach would permit a smaller data set to characterize the irradiation induced material property changes than if tabulated test results are directly used to represent the full range of operation. Furthermore, this paper will outline an example of the application of this approach in which the graphite material is exposed to temperatures that may exceed the upper bound temperatures examined in the Advanced Graphite Creep (AGC) testing program.