Simulated annealing tomography for nuclear fuel pin analysis

Gamma Emission Computed Tomography was used in the last years to evaluate, in a nondestructive manner, the distribution of the gamma radioactive fission products in the cross section of irradiated nuclear fuel rods. The shape of this distribution depends on the neutron flux but also on the behavior of the nuclear fuel during irradiation, especially on the migration of some fission products. Consequently, it can be used in diffusion kinetics studies in the determination of the release rate of fission products leaving overheated fuel in case of a severe accident or can offer information about the chemical species of the fission products. In the usual post-irradiation examination, the mentioned distribution is used to evaluate the self-absorption coefficient defined as the ratio between the number of photons escaping absorption in the fuel and the total number of emitted photons. The large value of the nuclear fuel density involves the absorption of an important number of photons, so, a proper burn-up evaluation needs a careful evaluation of the self-absorption coefficient. This evaluation can be done using Monte Carlo transport technique which consists of tracing the processes of a large number of photons in the cross section of the fuel pin. A tomographic method based on simulated annealing - a Monte Carlo global minimization procedure for solving nonlinear optimization problems - is used to reconstruct the three dimensional distribution of radioactive fission products in irradiated CANDU-type fuel rods. Good reconstruction was obtained for the case of a strong attenuation and for a small number of projections, for fuel pins irradiated in a TRIGA reactor. The material of the fuel pin of diameter 12.15 mm is sintered uranium dioxide surrounded by a cladding of Zircaloy-4 of thickness 0.4 mm. It contains 5.7% U-235 and it was irradiated during a period of 6048 hours. At the end of the irradiation a ramp has been performed. The final burn-up value obtained was 200 MWh/kgU. The tomographic reconstruction was performed on a 50 x 50 pixel plane using 5 projection angles. The reconstruction of the La-140 (1596 KeV) is presented. This distribution was used to evaluate the self-absorption needed in the burn-up calculation. The difference between the burn-up value evaluated in this manner and that obtained by radiochemical analysis was lower than 3%. This demonstrate that simulated annealing tomography is a suitable method for nuclear fuel rod analysis. The method has to be implemented on a powerful computer (like a HP-9000 workstation) to achieve a reasonable computing time. (authors)