Neutron Behavior in Cluster-Type Fuel Lattices, (II)

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  • Theory and Analysis of Experiment

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A multigroup method of calculation is presented for describing the neutron behavior in a clustertype fuel lattice. It solves the integral transport equation by a semi-analytical method proposed in a previous paper for calculating collision probabilities in the lattice of a clustered fuel element. Using only fundamental nuclear data, it gives space and energy dependent neutron flux. The method has been programmed for HITAC-5020F (computer code named CLUSTER-III).<br>The accuracy of the method has been tested by comparing the calculation with the experiment described in Part (I) of this paper. The lattices are 28-pin clusters of UO2 or PuO2+UO2 fuel pins, with heavy- or light-water moderators and with light-water coolant containing varying void ratios. The quantities studied are micro-parameters, reaction distributions in energy and space, thermal disadvantage factors and the multiplication factors. It is found that the calculated results are generally in good agreement with experiment, typically within 10% for micro-parameters and thermal disadvantage factor, and within 1% for the multiplication factor.

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Details 詳細情報について

  • CRID
    1390282679074711936
  • NII Article ID
    130000822028
  • DOI
    10.3327/jnst.9.705
  • COI
    1:CAS:528:DyaE3sXltFaktg%3D%3D
  • ISSN
    18811248
    00223131
  • Text Lang
    en
  • Data Source
    • JaLC
    • Crossref
    • CiNii Articles
  • Abstract License Flag
    Disallowed

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