Spectral averaged cross sections as a probe to a high energy tail of 235U PFNS

Abstract

The systematic evaluations of spectrum averaged cross sections of dosimetric reactions over a broad range of energies were performed in 252Cf (spontaneous fission) and 235U(nth,f) neutron fields. The neutron sources used in this study were LR-0, VR-1 zero power research light water reactors, LVR-15 10 MW research light water reactor, and 252Cf neutron source with emission specified precisely by the manganese sulphate bath. All spectral averaged cross sections were inferred from measured reaction rates which were derived from gamma spectrometry. These gamma spectrometry measurements were performed using a single detector in all cases. The ratios of 252Cf and 235U spectral averaged cross sections can be used to specify the high energy tail of the 235U prompt fission neutron spectrum as the 252Cf spontaneous fission spectrum is considered as a standard. Furthermore, ratios are independent of cross section uncertainties since uncertainties in the cross sections are eliminated. Theoretical models of fission can be tested based on our measurements. The calculations were performed in MCNP6.2 transport code using different prompt fission neutron spectra and IRDFF-II cross sections for threshold reactions. The ratios are in good agreement using only ENDF/B-VIII.0 235U prompt fission neutron spectrum suggesting it to be harder than in other evaluations.