### Abstract

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.

Original language | English |
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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

Conference | Joint International Topical Meeting on Mathematics and Computations and Supercomputing in Nuclear Applications, M&C + SNA 2007 |
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Country | United States |

City | Monterey |

Period | 15/04/07 → 19/04/07 |

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## Cite this

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.