High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels.
|Title of host publication||Mathematics and Computations and Supercomputing in Nuclear Applications (M&C+SNA 2007)|
|Subtitle of host publication||Monterey, CA, April 15-19, 2007|
|Publisher||American Nuclear Society ANS|
|Number of pages||12|
|ISBN (Electronic)||978-0-89448-059-1, 0-89448-059-6|
|Publication status||Published - 2007|
|MoE publication type||A4 Article in a conference publication|
|Event||Joint International Topical Meeting on Mathematics and Computations and Supercomputing in Nuclear Applications, M&C + SNA 2007 - Monterey, United States|
Duration: 15 Apr 2007 → 19 Apr 2007
|Conference||Joint International Topical Meeting on Mathematics and Computations and Supercomputing in Nuclear Applications, M&C + SNA 2007|
|Period||15/04/07 → 19/04/07|
Leppänen, J. (2007). Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code. In Mathematics and Computations and Supercomputing in Nuclear Applications (M&C+SNA 2007): Monterey, CA, April 15-19, 2007 American Nuclear Society ANS.