Analysis of Advanced Sodium-cooled Fast Reactor Core Designs with Improved Safety Characteristics
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The present doctoral work was performed to contribute to the conceptual design development and safety assessment of a Generation IV Sodium Fast Reactor (SFR) in the frame of the European Sodium Fast Reactor Safety Measures Assessment and Research Tools (ES ...
Characteristics of the spent nuclear fuel (SNF) are typically calculated, requiring validation a priori. The validation process relies on the difference between calculations and measurements, namely the bias. Usually, predicting the bias based on benchmark ...
The mechanical analysis of nuclear fuel behavior under base-irradiation conditions has traditionally been performed adopting the small-strain approximation. However, many cases of interest for fuel behavior involve the occurrence of large rod deformations ...
Zirconium alloys used in the nuclear industry are exposed to extreme conditions undergoing high levels of irradiation damage and corrosion. Zircaloy-2 is used as nuclear fuel cladding in boiling water reactors and for the encapsulation of the spallation ta ...
The calculations performed for the design and operation of a Nuclear Power Plant (NPP) are a key factor for their safety analyses. The standard for the computational analysis of NPPs is the so called conventional approach, which relies on coarse mesh diffu ...
As one of promising candidate materials for fuel claddings and structural components in the Gen-IV fission reactors, FeCrAl(Zr)-ODS ferritic steels were studied to well understand the radiation hardening behavior. Nanoindentation (NI) hardness and plastica ...
The sequence of codes Serpent/DYN3D has been developed by the Helmholtz-Zentrum Dresden-Rossendorf and successfully applied to core static and transient analyses of sodium-cooled fast reactors (SFRs). The successful application of the sequence to SFRs was ...
In this thesis, the development and testing of a system for measuring the axial distribution of fast neutron emission of spent nuclear fuel rods is presented. Emphasis is placed on the novel fast neutron detector used which can reliably work in extremely h ...
This work presents a preliminary safety analysis of the accelerator-driven system under development at TRANSMUTEX. It is a sub-critical lead-cooled fast reactor with a nominal power of 300 MWth having the following features: (1) an inert-matrix fuel; (2) a ...
The PETALE program aims to provide new experimental data to constrain the stainless steel nuclear data. In this frame, a preliminary measurement campaign has been performed to characterize the neutron flux in key positions of the CROCUS reactor and to deve ...