Feasibility study of U-235, Pu-239 and Pu-240 content determination in an irradiated fuel by neutron transmission analysis

A proposed nondestructive method and its feasibility for the determination of U-235, Pu-239 and Pu-240 contents in an irradiated fuel is described. The method is based on the use of shape fit analysis of the Time-Of-Flight (TOF) neutron transmission data of the irradiated fuel for neutron energies below 3 eV. The neutron transmission experiment of the irradiated fuel is planned to carry out using one of the TOF spectrometers installed at ET-RR-J reactor. The computer code SHAPE is adapted taking into account the known parameters of resonances of certain fissile and fission product nuclei to provide the fit analysis. The content of the gross-fissile and fission product isotopes are determined from the burn-up calculations of the fuel assembly of the ET-RR-I reactor with defined history. The effect of both uncertainties in resonance parameters on the deduced contents of fissile nuclei and statistical accuracy of the TOF measurements are estimated.