TY - GEN
T1 - Property evaluation and microstructure characterization of the A3-3 matrix graphite
AU - Zhou, Xiangwen
AU - Campbell, Anne A.
AU - Katoh, Yutai
AU - Lu, Zhenming
AU - Zhang, Jie
AU - Contescu, Cristian I.
AU - Liu, Bing
PY - 2016
Y1 - 2016
N2 - Spherical fuel elements with TRISO fuel will be used in the pebble-bed high temperature gas-cooled nuclear reactor. The fuel will be located within the center region of the elements, and will be suspended in, and surrounded by a shell of matrix graphite. The matrix graphite provides structural support to the fuel, prevents fuel damage, and allows for heat transfer from the fuel to the coolant. The raw materials for the matrix graphite, type A3-3, includes 64wt% natural flake graphite, 16wt% artificial graphite and 20wt% phenol resin binder. After the final heat treatment, the resulting A3-3 matrix is composed of71wt% natural graphite, 18wt% artificial graphite, and 11wt% phenol resin-derived coke carbon. The microstructure of A3-3 matrix graphite has direct influences on the material properties, which are crucial to maintaining the integrity and safety of the fuel elements. This presentation will discuss microstructure features, such as lattice, and crystal parameters, porosity, pore size and distribution, and anisotropy. The multiple physical, mechanical and thermal properties of this material, including: elastic modulus, mechanical strength, and oxidation property will also be presented. Support from the State Scholarship Foundation of China (201406215002), the Chinese National S & T Major Project (ZX06901), and Tsinghua University Initiative Scientific Research Program (20121088038) are acknowledged. The work performed at Oak Ridge National Laboratory was supported partially by U.S. Department of Energy Advanced Reactor Technology Program, Office of Nuclear Energy. Oak Ridge National Laboratory is managed by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 for the U.S. Department of Energy.
AB - Spherical fuel elements with TRISO fuel will be used in the pebble-bed high temperature gas-cooled nuclear reactor. The fuel will be located within the center region of the elements, and will be suspended in, and surrounded by a shell of matrix graphite. The matrix graphite provides structural support to the fuel, prevents fuel damage, and allows for heat transfer from the fuel to the coolant. The raw materials for the matrix graphite, type A3-3, includes 64wt% natural flake graphite, 16wt% artificial graphite and 20wt% phenol resin binder. After the final heat treatment, the resulting A3-3 matrix is composed of71wt% natural graphite, 18wt% artificial graphite, and 11wt% phenol resin-derived coke carbon. The microstructure of A3-3 matrix graphite has direct influences on the material properties, which are crucial to maintaining the integrity and safety of the fuel elements. This presentation will discuss microstructure features, such as lattice, and crystal parameters, porosity, pore size and distribution, and anisotropy. The multiple physical, mechanical and thermal properties of this material, including: elastic modulus, mechanical strength, and oxidation property will also be presented. Support from the State Scholarship Foundation of China (201406215002), the Chinese National S & T Major Project (ZX06901), and Tsinghua University Initiative Scientific Research Program (20121088038) are acknowledged. The work performed at Oak Ridge National Laboratory was supported partially by U.S. Department of Energy Advanced Reactor Technology Program, Office of Nuclear Energy. Oak Ridge National Laboratory is managed by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 for the U.S. Department of Energy.
UR - http://www.scopus.com/inward/record.url?scp=85026315582&partnerID=8YFLogxK
M3 - Conference contribution
AN - SCOPUS:85026315582
T3 - International Topical Meeting on High Temperature Reactor Technology, HTR 2016
SP - 772
EP - 775
BT - International Topical Meeting on High Temperature Reactor Technology, HTR 2016
PB - American Nuclear Society
T2 - 8th International Topical Meeting on High Temperature Reactor Technology, HTR 2016
Y2 - 6 November 2016 through 10 November 2016
ER -
