Direct Determination of Uranium-235 Capture-to-Fission Ratio in a Zero-Power Reactor

An experimental method for the determination of the spectral average of the capture-to-fission ratio ᾱ for materials inserted in a low-flux reactor is described. The procedure involves a comparison of reactor response to oscillated samples of a fissile material, an absorber, and a spontaneous fission neutron source, plus an experimental determination of fission rate for the fissile material and capture rate for the absorber. In addition, it is necessary that the neutron source be calibrated. These experimental results, combined with a knowledge of the number of neutrons per fission for the fissile material, yield a value of the quantity 1 + ᾱ. This method has been tested in Hi-C Core 10, a critical assembly of 3%-enriched-UO2 fuel pins, moderated and reflected by light water, in a lattice spacing which yields a H-to-238U atom ratio of 2:91. The oscillator and absolute counting data yield a value of 0.217 for the central capture-to-fission ratio of 235U, with a standard deviation of ± 0.015. This agrees well with values derived from a combination of measured 235U fission cadmium ratios and calculated thermal and epithermal values for α.