On the effects of oxygen cross-sections in the fast neutron fluence analysis for a reactor pressure vessel scraping test

摘要

Studies in support of the assessment of aging structural materials in pressurized water reactors are being performed at the Paul Scherrer Institut. To that aim, a state-of-the-art methodology based on applying a CASMO-4/SIMULATE-3/MCNPX calculation scheme has been developed. In the frame of the methodology validation, an investigation is currently reported pertaining to the sensitivity of the calculated results, for a specific reactor pressure vessel scraping test, to the nuclear data used with the Monte Carlo code. Thus, the MCNPX-2.4.0 calculations have been carried out using three different data libraries, based on JEF-2.2, ENDF/B-VI.8 and JENDL-3.3 evaluations, respectively. A significant discrepancy (～10%) in fast neutron flux (E > 1 MeV) estimations has been observed between the results obtained with JENDL-3.3, versus the two other libraries, while noting that the later both provide very satisfactory agreement (within ±5%) with the reference results based on the experimental data. Subsequent analysis has indicated that the observed discrepancy can be attributed mostly to differences in the oxygen data. The discussed cross-section differences could potentially lead to more sizable discrepancies for other applications, and thus need to be brought to the attention of neutron data evaluators and users.