New method for analysis of X-ray computed tomography scans of TRISO fuel forms

Grant W. Helmreich, John D. Hunn, Daniel R. Brown, Brandon J. Blamer

Research output: Contribution to journalArticlepeer-review

11 Scopus citations

Abstract

X-ray computed tomography (XCT) is a powerful characterization tool that has been used for examination of both tristructural-isotropic (TRISO) coated particles and fuel forms. Rapid, nondestructive imaging of complete specimens by XCT provides significant characterization advantages in both speed and comprehensiveness over traditional materialographic methods. This paper discusses a method that has been developed to analyze XCT images of pebble fuel forms containing ~18,000 TRISO particles each. The XCT images are three-dimensional with contrast based on relative x-ray transparency. In the case of TRISO fuel forms, XCT image intensity distinguishes between the high-density kernel, the medium-density SiC, and the low-density matrix and carbon layers. Image analysis software has been developed to process XCT data sets to extract spatial distribution information, including particle nearest-neighbor distances, local packing fraction, and fuel-free zone thickness in TRISO fuel forms. Although this type of high-resolution XCT analysis is likely not feasible as a quality control measurement for commercial TRISO fuel, it is highly useful in generating feedback for the development of fuel form fabrication processes.

Original languageEnglish
Article number110418
JournalNuclear Engineering and Design
Volume357
DOIs
StatePublished - Feb 2020

Funding

This work was sponsored by the U.S. Department of Energy , Office of Nuclear Energy , through the Advanced Reactor Concepts ARC-Xe program and through the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office as part of the Advanced Gas Reactor Fuel Development and Qualification Program. This research was conducted, in part, using instrumentation within ORNL’s Materials Characterization Core provided by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the US Department of Energy. This work was sponsored by the U.S. Department of Energy, Office of Nuclear Energy, through the Advanced Reactor Concepts ARC-Xe program and through the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office as part of the Advanced Gas Reactor Fuel Development and Qualification Program. This research was conducted, in part, using instrumentation within ORNL's Materials Characterization Core provided by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the US Department of Energy.

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