### Abstract

This paper presents utilization of the fission matrix (FM) methodology to analyze a spent fuel pool. This FM approach utilizes a pre-calculated MCNP-generated database of fission matrix coefficients which are created at different burnups and cooling times. Certain simplifying assumptions are made based on geometric and physical considerations, greatly reducing the amount of pre-computation required. This approach is capable of quickly and accurately determine pin-wise, axial-dependent fission density distribution and subcritical multiplication (M) or criticality (fc) of a spent fuel pool, in any arrangement, without recalculating FM coefficients. This paper examines the use of the FM approach for different test pool arrangements and conditions. Excellent agreements with an MCNP reference calculation have been achieved with several orders of magnitude reduction in computation time.

Original language | English (US) |
---|---|

Title of host publication | Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015 |

Publisher | American Nuclear Society |

Pages | 1351-1361 |

Number of pages | 11 |

ISBN (Electronic) | 9781510808041 |

State | Published - Jan 1 2015 |

Event | Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015 - Nashville, United States Duration: Apr 19 2015 → Apr 23 2015 |

### Publication series

Name | Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015 |
---|---|

Volume | 2 |

### Other

Other | Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015 |
---|---|

Country | United States |

City | Nashville |

Period | 4/19/15 → 4/23/15 |

### Fingerprint

### All Science Journal Classification (ASJC) codes

- Mathematics(all)
- Nuclear and High Energy Physics

### Cite this

}

*Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015.*Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015, vol. 2, American Nuclear Society, pp. 1351-1361, Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015, Nashville, United States, 4/19/15.

**A fission matrix approach to calculate pin-wise 3-D fission density distribution.** / Walters, William J.; Roskoff, Nathan; Haghighat, Alireza.

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution

TY - GEN

T1 - A fission matrix approach to calculate pin-wise 3-D fission density distribution

AU - Walters, William J.

AU - Roskoff, Nathan

AU - Haghighat, Alireza

PY - 2015/1/1

Y1 - 2015/1/1

N2 - This paper presents utilization of the fission matrix (FM) methodology to analyze a spent fuel pool. This FM approach utilizes a pre-calculated MCNP-generated database of fission matrix coefficients which are created at different burnups and cooling times. Certain simplifying assumptions are made based on geometric and physical considerations, greatly reducing the amount of pre-computation required. This approach is capable of quickly and accurately determine pin-wise, axial-dependent fission density distribution and subcritical multiplication (M) or criticality (fc) of a spent fuel pool, in any arrangement, without recalculating FM coefficients. This paper examines the use of the FM approach for different test pool arrangements and conditions. Excellent agreements with an MCNP reference calculation have been achieved with several orders of magnitude reduction in computation time.

AB - This paper presents utilization of the fission matrix (FM) methodology to analyze a spent fuel pool. This FM approach utilizes a pre-calculated MCNP-generated database of fission matrix coefficients which are created at different burnups and cooling times. Certain simplifying assumptions are made based on geometric and physical considerations, greatly reducing the amount of pre-computation required. This approach is capable of quickly and accurately determine pin-wise, axial-dependent fission density distribution and subcritical multiplication (M) or criticality (fc) of a spent fuel pool, in any arrangement, without recalculating FM coefficients. This paper examines the use of the FM approach for different test pool arrangements and conditions. Excellent agreements with an MCNP reference calculation have been achieved with several orders of magnitude reduction in computation time.

UR - http://www.scopus.com/inward/record.url?scp=84949499352&partnerID=8YFLogxK

UR - http://www.scopus.com/inward/citedby.url?scp=84949499352&partnerID=8YFLogxK

M3 - Conference contribution

AN - SCOPUS:84949499352

T3 - Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015

SP - 1351

EP - 1361

BT - Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015

PB - American Nuclear Society

ER -