Effect of graphite properties in thermal analysis of CHTR: A parametric study
- Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai-400 085 (India)
Compact High Temperature Reactor (CHTR) is a {sup 233}U-Thorium fuelled, lead-bismuth cooled reactor. The CHTR core mainly consists of graphite and beryllium oxide (BeO). The CHTR core consists of nineteen prismatic beryllium oxide (BeO) moderator blocks. These 19 blocks contain centrally located graphite fuel tubes. The BeO moderator blocks are surrounded by reflector blocks (partially graphite and partially BeO). The nuclear heat from the core is removed passively by natural circulation of the coolant between top and bottom plenums, upward through the fuel tubes and returning through the downcomer tubes at the periphery. The temperature gradient in fuel tubes, downcomer tubes and BeO is very high and therefore, to take care of the differential thermal expansion, gaps are provided in the core between the tubes and other core components. These gaps affect the heat transfer through the core in radial direction. In addition, there is a large variation in thermal properties of graphite which in turn affects the thermal behaviour of the core in various operating conditions. The fuel of CHTR is TRISO coated particle fuel. These particles are packed in with graphite powder as matrix and made into cylindrical compacts these compacts are packed in the bores of fuel tube. In this study, the effect of the thermal conductivity variation of the graphite on the temperature distribution of the core and density variation of the matrix graphite material in fuel compact on the maximum fuel kernel temperature is studied along with the overall role of graphite properties variation in heat transfer.
- OSTI ID:
- 22118572
- Journal Information:
- AIP Conference Proceedings, Vol. 1538, Issue 1; Conference: CCM12: International conference on carbon materials for energy harvesting, environment, nanoscience and technology, Mumbai (India), 1-3 Nov 2012; Other Information: (c) 2013 AIP Publishing LLC; Country of input: International Atomic Energy Agency (IAEA); ISSN 0094-243X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
BERYLLIUM OXIDES
BISMUTH
FUEL PARTICLES
GRAPHITE
LEAD
LMFBR TYPE REACTORS
MODERATORS
NATURAL CONVECTION
NUCLEAR FUELS
PARAMETRIC ANALYSIS
REACTOR CORES
TEMPERATURE DISTRIBUTION
TEMPERATURE GRADIENTS
THERMAL ANALYSIS
THERMAL CONDUCTIVITY
THERMAL EXPANSION
THORIUM
URANIUM 233