Description
Chairs: Marina Demeshko and Yamano Hidemasa
The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy, Office of Nuclear Energy. The mission of the VTR is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled...
Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commision), a methodology is proposed to enable the early consideration of safety in the undergoing design process. Before the use of mechanistic tools (CATHARE, SIMMER, APOLLO, etc.) that usually requires an advanced knowledge of the reactor design, this...
To improve the safety of future sodium cooled fast reactor, alternative Decay Heat Removal (DHR) path is envisaged through secondary sodium main circuit. Secondary Sodium Decay Heat Removal System (SSDHRS) transfers heat from secondary sodium to ambient through Air Heat Exchanger (AHX). SSDHRS is planned to operate in forced circulation mode using Class 3 power supply. SSDHRS in addition to...
The Fukushima accident highlighted the need to improve the design of the safety systems in order to cope the consequences of a full blackout accident. In particular, the decay heat removal function must always be ensured in order to guarantee the integrity of the first barrier and the second one confining the radioactivity. The paper shows the main features of the pre-design of an innovative...
The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy (USDOE), Office of Nuclear Energy. The mission of the VTR is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled...
Lead cooled reactor BREST-OD-300 design is under development as a part of Russian federal project "PRORYV". Two circuits are used for heat removal from the reactor. The special feature of two-circuit heat removal system is the potential risk of steam ingress into the core in case of large leak in the steam generator as a result, for example, of steam generator tube rupture (SGTR). The main...
Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor, operating at Kalpakkam. FBTR core is originally designed to operate at 40 MWt using mixed oxide (MOX) fuel with 30% PuO2 and 70% UO2 (85% enriched U). However, due to the non-availability of enriched uranium, mixed carbide fuel (Mark-I (70% PuC+30%UC)) was chosen for the initial core. After first criticality in 1985,...
The MSFR (Molten Salt Fast Reactor) consists in a concept of high power molten salt reactor. In molten salt reactors, the fissile and fertile nuclei are dissolved in a circulating salt that acts as fuel and coolant. The physical state of the fuel permits to consider draining as a way to mitigate hypothetical accidents. Contrary to solid fuel FNR (Fast Neutron Reactor) concepts, in the MSFR,...
