Alemberti A.,Ansaldo Energia |
Smirnov V.,NIKIET |
Smith C.F.,Naval Postgraduate School, Monterey |
Takahashi M.,Tokyo Institute of Technology
Progress in Nuclear Energy | Year: 2014
The lead-cooled fast reactor (LFR) features a fast neutron spectrum, high temperature operation, and cooling by molten lead or lead-bismuth eutectic (LBE), low-pressure, chemically inert liquids with very good thermodynamic properties. It would have multiple applications including production of electricity, hydrogen and process heat. System concepts represented in plans of the Generation IV International Forum (GIF) System Research Plan (SRP) are based on Europe's ELFR lead-cooled system, Russia's BREST-OD-300 and the SSTAR system concept designed in the US. The LFR has excellent materials management capabilities since it operates in the fast neutron spectrum and uses a closed fuel cycle enhanced by the fertile conversion of uranium. It can also be used as a burner to consume actinides from spent LWR fuel and as a burner/breeder with thorium matrices. An important feature of the LFR is the enhanced safety that results from the choice of molten lead as a chemically inert and low-pressure coolant. In terms of sustainability, lead is abundant and hence available, even in case of deployment of a large number of reactors. More importantly, as with other fast systems, fuel sustainability is greatly enhanced by the conversion capabilities of the LFR fuel cycle. Because they incorporate a liquid coolant with a very high margin to boiling and benign interaction with air or water, LFR concepts offer substantial potential in terms of safety, design simplification, proliferation resistance and the resulting economic performance. An important factor is the potential for benign end state to severe accidents. The LFR has development needs in the areas of fuels, materials performance, and corrosion control. During the next 5 years progress is expected on materials, system design, and operating parameters. Significant test and demonstration activities are underway and planned during this time frame. © 2013 Elsevier Ltd. All rights reserved.
Giancarli L.M.,ITER Organization |
Abdou M.,University of California at Los Angeles |
Campbell D.J.,ITER Organization |
Chuyanov V.A.,ITER Organization |
And 10 more authors.
Fusion Engineering and Design | Year: 2012
The objective of the ITER TBM Program is to provide the first experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. Such information is essential to design and predict the performance of DEMO and future fusion reactors. It foresees to test six mock-ups of breeding blankets, called Test Blanket Module (TBM), in three dedicated ITER equatorial ports from the beginning of the ITER operation. The TBM and its associated ancillary systems, including cooling system and tritium extraction system, forms the Test Blanket System (TBS) that will be fully integrated in the ITER machine and buildings. This paper describes the main features of the six TBSs that are presently planned for installation and operation in ITER, the main interfaces with other ITER systems and the main aspects of the TBM Program management. © 2011 Elsevier B.V. All rights reserved.