Recently, several advanced multidimensional computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. One of such advanced, best-estimate systems codes is TRAC/RELAP Advanced Computational Engine (TRACE), developed by the U.S. Nuclear Regulatory Commission. The TRACE comes with a graphical user interface called SNAP (Symbolic Ana...

The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design WCLL-BB (water-cooled lead-lithium breeding blanket). Research activities are ongoing master phenomena and processes that occur during postulated accident, enhance predictive capability reliability numerical tools, validate computer models, codes, procedures for their applications. Followi...

The present paper describes the experimental campaign executed at ENEA Brasimone Research Centre aiming supporting development of a PbLi/water heat exchanger suitable for lithium–lead loops dual coolant lithium lead and water cooled breeding blankets EU DEMO fusion reactor. experiments were performed in test section named HERO, installed inside main vessel lead–bismuth eutectic-cooled pool-type...

A web-based nuclear reactor simulator has been developed using the best-estimate nuclear system analysis code RELAP5 as its engine, and abVIEW for graphical user interface and web-casting. Simulator retains the accuracy of the best-estimate code. Results are displayed in user riendly graphical format. Color-coded nominal values are displayed along with the current status of different variables ...

The Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR) was modeled using the neutronics analysis code SCALE6.0 and thermal-hydraulics kinetics modeling RELAP5-3D with objective to devise, analyze, evaluate feasibility stability of a start-up procedure for this reactor natural circulation coolant under Loss Of Offsite Power (LOOP) conditions. This Generation IV design has been initially ...

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As part of an I-NERI project to develop safety analysis codes and experimental validation for a VeryHigh Temperature Gas-Cooled Reactor (VHTGR), we have developed MCNP5 [1] models to represent material heterogeneities inherent in the microsphere fuel particles and fuel compacts for a GT-MHR design [2]. We have also performed preliminary coupled nuclear-thermalhydraulic (NTH) analysis to obtain ...