Enhanced Nuclear Engineering Simulators

Engineering simulation is a sophisticated multi-purpose technology allowing the users of simulators to run a variety of engineering activities due to the possibility of modifying the simulated plant architecture and components, to adjust parameters, to test alternative solutions. Engineering Simulators (ES) have been built and used worldwide for a variety of purposes: - Development and refinement of the plant design or plant modifications - Safety analyses focused on the overall system behaviour - Verification and Validation (V&V) of systems and components - Development of Operational and Emergency Procedures - Pre-Training of operators and supervisors - High level education and Communication activities - Human Factor Engineering Analysis - Adaptive Control System training Engineering Simulators also play a role in developing and maintaining key nuclear skills, as knowledge repositories and tools for training at various levels of expertise

Analisi e modellazione del trasporto pedonale in aree urbana

Comunicazione all'interno del seminario "Metodi e tecnologie dell'ingegneria dei trasport

Feasibility Analysis and Uncertainty Quantification for a "Fast-Running" Chain of Codes for the NPP Accident Management

This work briefly describes the computer codes RAVEN and DAKOTA, illustrating how to use them in the ambit of uncertainty quantification during PRA thermal hydraulic analyses. The second part shows instead two simple applications of RAVEN code to study SBO transient. In the first one, it has been coupled to RELAP-7 to analyze a PWR, in simplified manner (due to the current limitations of the RELAP-7 code, that is under development). In the second one, it was coupled to RELAP-5 3D code in order to analyze a BWR MARK-1. These two applications demonstrate the potential of the RAVEN code as an optimal tool for the coupling with codes for TH analyses

Effects of cross sections libraries parameters on the OECD/NEA OSKARSHAMN-2 benchmark solution

The OECD/NEA proposes a new international benchmark based on the data collected from an instability transient occurred at the Oskarshamn-2 NPP with the aim to test the coupled 3D Neutron Kinetic/Thermal Hydraulic codes on challenging situations. The ENEA “Casaccia” Research Center is participating to this benchmark, developing a computational model using RELAP5-3D© code. The 3DNK model was developed starting from the cross sections datasets calculated by OKG, the Oskarshamn-2 licensee, using the CASMO lattice code. Integration of neutron cross sections database in RELAP5-3D© required data fitting by a n-dimensional polynomials, calculations of the various polynomial coefficients and of the base cross sections values. An ad-hoc tool named PROMETHEUS has been developed for automatically generate the RELAP5-3D©-compatible cross sections libraries. Thanks to this software it has been easily possible to visualize the complex structure of the neutronic data sets and to derive different cross sections libraries for evaluating the effects of some neutronic parameters on the prediction of the reactor instability. Thus, the effects of the fuel temperature and control rod history, of the discontinuity factors (averaged/not averaged), and of the neutron poisons has been assessed. A ranking table has been produced, demonstrating the relevance of the not-averaged discontinuity factors and of the on-transient neutron poisons calculations for the correct prediction of the Oskarshamn-2 even