### Abstract

Original language | English |
---|---|

Pages (from-to) | 878-885 |

Number of pages | 8 |

Journal | Annals of Nuclear Energy |

Volume | 36 |

Issue number | 7 |

DOIs | |

Publication status | Published - 2009 |

MoE publication type | A1 Journal article-refereed |

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### Keywords

- neutron transport theory
- neutrons

### Cite this

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**Two practical methods for unionized energy grid construction in continuous-energy Monte Carlo neutron transport calculation.** / Leppänen, Jaakko (Corresponding Author).

Research output: Contribution to journal › Article › Scientific › peer-review

TY - JOUR

T1 - Two practical methods for unionized energy grid construction in continuous-energy Monte Carlo neutron transport calculation

AU - Leppänen, Jaakko

PY - 2009

Y1 - 2009

N2 - A considerable speed-up in continuous-energy Monte Carlo neutron transport calculation can be achieved by using the same unionized energy grid for all point-wise reaction cross sections. This speed-up results from the fact that time-consuming grid iteration is reduced to minimum, and if the unionized grid is constructed by combining the grids of all nuclides, there is no loss of data or accuracy in the calculation. The drawback of this approach is that computer memory is wasted for storing a large number of redundant data points. Memory usage may become a problem, especially in burnup calculation, in which the irradiated materials consist of several hundred actinide and fission product nuclides. The grid size easily increases to over 1 million points, requiring tens of gigabytes of memory for storing the cross section data. This paper presents two practical methods for reducing the memory demand, while trying preserve the accuracy of the original data. The calculation routines are included in the PSG2/Serpent Monte Carlo reactor physics burnup calculation code and the methods are tested in a BWR assembly burnup calculation.

AB - A considerable speed-up in continuous-energy Monte Carlo neutron transport calculation can be achieved by using the same unionized energy grid for all point-wise reaction cross sections. This speed-up results from the fact that time-consuming grid iteration is reduced to minimum, and if the unionized grid is constructed by combining the grids of all nuclides, there is no loss of data or accuracy in the calculation. The drawback of this approach is that computer memory is wasted for storing a large number of redundant data points. Memory usage may become a problem, especially in burnup calculation, in which the irradiated materials consist of several hundred actinide and fission product nuclides. The grid size easily increases to over 1 million points, requiring tens of gigabytes of memory for storing the cross section data. This paper presents two practical methods for reducing the memory demand, while trying preserve the accuracy of the original data. The calculation routines are included in the PSG2/Serpent Monte Carlo reactor physics burnup calculation code and the methods are tested in a BWR assembly burnup calculation.

KW - neutron transport theory

KW - neutrons

U2 - 10.1016/j.anucene.2009.03.019

DO - 10.1016/j.anucene.2009.03.019

M3 - Article

VL - 36

SP - 878

EP - 885

JO - Annals of Nuclear Energy

JF - Annals of Nuclear Energy

SN - 0306-4549

IS - 7

ER -