Evaluation of the Induced to Spontaneous Fission Neutron Multiplicity Distribution Coefficients for Passive Neutron Multiplicity Counting of Plutonium and MOX

Shift Register based passive neutron multiplicity counting of PuO2 or MOX samples relies on models to interpret the measured data. To attain the highest accuracy these models require an appropriate description of the neutron multiplets produced in the nuclear material. In order to account for the differences in neutron multiplicity distributions between the initial spontaneous fission events and subsequent prompt induced fission events, a series of correction coefficients must be derived. These coefficients, the so-called ?-coefficients, are introduced into the neutron multiplication terms of the passive neutron multiplicity counting interpretative model. They are expressed as overlap integrals involving the isotopic composition of the sample, the first three factorial moments of the induced (?I1, ?I2, ?I3) and spontaneous (?S1, ?S2, ?S3) fission neutron multiplicity distributions, the induced fission cross section of the fissile isotopes and the energy spectra of neutrons that induce fission. In the absence of experimental evaluation of the induced fission factorial moments as a function of incident neutron energy for plutonium isotopes other than 239Pu, the authors have used a limited data set (239Pu, 235U and 238U) in conjunction with a normal probability distribution model of emitted multiplicity to derive approximate values of all of the important induced fission factorial moments. Several empirical relationships between ?I1,?I2 and ?I3 as a function of energy have been used to produce alternative values of these parameters in order to evaluate the sensitivity of the results to the approach adopted. Propagation of the uncertainties has been performed throughout the calculations so that the resulting ? values derived in this work could be used with confidence for the purpose of quantitative neutron multiplicity counting. Having established a calculational scheme, ?-coefficients for doubles and triples counting of isotope enriched PuO2 samples have been compared with those of typical reactor grade PuO2 and MOX. The principal findings of this study were: firstly that ?-coefficients exhibit only a small dependence on the isotopic composition of a given type of plutonium product; secondly that, for best accuracy, the neutron energy spectral effect from (a,n) neutrons should be accounted for, especially when material with a significant (a,n) contribution are encountered; finally, that significantly different values of the ?-coefficients should be used when MOX or PuO2 material assay is performed.