### Abstract

Static and dynamic analyses of nuclear reactors are
usually carried out with diffusion-theory based nodal
methods, in which the fuel assemblies are modelled in the
horizontal direction as one or four and in the axial
direction as 10-30 homogenized volumes. The homogenized
two-group cross sections needed in the nodal analysis are
obtained from accurate single-node heterogeneous
multigroup transport calculations.
The nodal models differ from each other in the way the
intra-nodal flux shape is approximated in calculation of
the coupling coefficients between adjacent nodes. The
simplest methods use polynomials or other simple
functions for direct interpolation of the fast and
thermal fluxes; a more advanced approach, used e.g. in
the HEXBU and TRAB-3D nodal models (Kaloinen &
Kyrki-Rajamäki 1997), is to decompose the two-group flux
into eigenmodes, which have a better-defined shape than
the fast and thermal flux and can thus be interpolated
more accurately.
Especially in cores with large flux variations between
adjacent nodes, it is important to obtain as accurate an
estimate of the intra-nodal flux shape as possible.
Another need for improving the flux shape comes from pin
power reconstruction. Most of the current nodal codes
(including TRAB-3D and HEXBU) do not, however, use corner
point flux values in the nodal calculations and
information about the flux near node corners is therefore
limited. A new nodal model, based on the analytic
function expansion nodal model (AFEN), initially
developed in South Korea (Noh & Cho 1994), has been
developed for calculating the three-dimensional neutron
flux. In addition to improved accuracy within the core,
the analytic function expansion enables also calculation
of the flux in reflector areas, thus possibly enabling
better evaluation of e.g. the flux seen by ex-core
detectors. This paper briefly describes the model and the
results of the first test calculations.

Original language | English |
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Title of host publication | FINNUS: The Finnish Research Programme on Nuclear Power Plant Safety |

Subtitle of host publication | Interim Report 1999 - August 2000 |

Place of Publication | Espoo |

Publisher | VTT Technical Research Centre of Finland |

Pages | 161-168 |

ISBN (Electronic) | 951-38-5751-4 |

ISBN (Print) | 951-38-5750-7 |

Publication status | Published - 2000 |

MoE publication type | D2 Article in professional manuals or guides or professional information systems or text book material |

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## Cite this

Mattila, R. (2000). READY special report 2: Three-dimensional analytic function expansion nodal model for the TRAB-3D nodal code. In

*FINNUS: The Finnish Research Programme on Nuclear Power Plant Safety: Interim Report 1999 - August 2000*(pp. 161-168). VTT Technical Research Centre of Finland. http://www.vtt.fi/inf/pdf/tiedotteet/2000/T2057.pdf