Abstract
Thermal neutron distribution in a square lattice of water moderated UO2 (enriched to 2.596 w/0 in 235U) was measured with dysprosium micro-foils. Diameters of the fuel pellet and the aluminum cladding tube were 12.5mm and 14.12mm respectively. The pitch of the square lattice was 19.56mm, and the water to fuel volume ratio 1.844. To obtain the integrated flux in the fuel and moderator regions, the solution for the P-1 diffusion equation was used. A disadvantage factor of 1.27±0.02 and a thermal utilization factor of 0.891±0.002 were obtained. The theoretical value of the disadvantage factor obtained by the 30 group integral transport theory in a square call is 1.341 and is larger than the experimental value by a discrepancy exceeding the experimental error. The result is also compared with some approximate calculations.
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