During normal operation of a Nuclear Power Plant (NPP), the Reactor Pressure Vessel (RPV) is exposed to an intense neutron irradiation due to its proximity to the reactor core. Therefore, Fast Neutron Fluence (FNF) analysis is an important part of RPV ageing/life-time assessments and associated safety analyses of NPPs.
Thus, one of the activities conducted at the Paul Scherrer Institut (PSI) in Switzerland is the establishment of a computational platform containing core models for all the Swiss light water reactors (4 NPPs) for high-fidelity FNF calculations.
The PSI FNF modeling methodology is based on the use of the CASMO/SIMULATE/MCNPX system of codes. The two deterministic codes CASMO-4 and SIMULTES-3 are combined to estimate the neutron source. The Monte Carlo code MCNPX is used for the neutron transport. The necessary input cards for MCNPX (source definition) are directly produced from the results of the CASMO/SIMULATE core-follow calculations through the dedicated "SOURCE4MC" tool developed at PSI.
By present time a considerable amount of dosimetry experimental data from both Swiss PWRs and BWRs has been analyzed at PSI for validation of the above mentioned FNF modeling methodology. However, for the specific BWR plant under consideration in this paper, the past validation studies concerned only single cycle-long irradiation programs.
Now, the paper at hand presents the on-going work on additional validation studies based on longer irradiation durations in order to complete the validation database. Indeed, the reactor parameters (operating power, fuel used, core loading scheme, etc…) may vary considerably during its lifetime, inducing important changes in the irradiation history of the RPV. Therefore, it is of importance to look at the RPV FNF not only on the duration of one cycle, but also on longer periods of operation. In this context, it was proposed to do a new validation study based on the experimental data provided by the plant and obtained from the dosimeter set of the first surveillance capsule irradiated during cycles 1 through 11 at the given BWR. Moreover, certain steps towards improvement of the modeling/calculation accuracy and precision are being planned and realized. Finally, an impact of nuclear data library choice on the calculation and validation results is investigated and the relevant cross-section data differences are discussed.