The experimental determination and evaluation of the spectral indices of the IPEN/MB-01 reactor for the IRPhE project

Highlights: • A new experimental method for the spectral index measurements was developed. • The method eliminates the need of calculated correction factors for the case of 28ρ. • The experimental results were approved for inclusion in the IRPhE handbook. • MCNP5 with ENDF/B-VII.0 described well the experimental data. - Abstract: New experimental results for the spectral indices of the IPEN/MB-01 reactor are presented in this work. The experimental approach considers a new technique for the 28ρ case which does not require any sort of calculated correction factors. This aspect gave to the IPEN/MB-01 experiment an excellent quality and free of possible bias due to these calculated correction factors. The uncertainty analysis show, even considering the uncertainties of the geometric and material data of the facility, that the final uncertainties are small enough and well understood for a benchmark problem. The experiment was evaluated and included in the IRPhE handbook. The theory/experiment comparison reveals that there is considerable progress in the 238U nuclear data for application in thermal reactors. The theory/experiment comparison employs the nuclear data libraries ENDF/B-VII.0, ENDF/B-VI.8, JEF3.1, and JENDL3.3 and reveals an excellent agreement for the spectral index 28ρ* independently of the nuclear data library considered; aspect never found before in several other comparisons. The long term overprediction of the 238U neutron epithermal capture appears to be eliminated with the improvements in the recent nuclear data libraries. The experimental performed at the IPEN/MB-01 reactor supports the changes in the 238U nuclear data incorporated in ENDF/B-VII.0 as well as in the other libraries studied in this work. The theory/experiment comparison of 25δ* and (C8/F)ept show that these spectral indices are in general slightly overpredicted, thus suggesting that the thermal fission cross section of 235U might be a little bit underestimated. The overall analysis of the theory/experiment comparisons show the excellent applicability of ENDF/B-VII.0 for thermal reactors fuelled with slightly enriched uranium