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Economics of Nuclear Power
Published in Kenneth D. Kok, Nuclear Engineering Handbook, 2016
The sodium-cooled fast reactor (SFR) uses liquefied sodium as the primary coolant. The major advantage is that the reactor can operate at atmospheric pressure in the primary coolant loop. The ultimate inherent safety of this reactor, when using metallic (not oxide) uranium fuel, was demonstrated in 1986 with the Experimental Breeder Reactor II at the Idaho National Laboratory, under the direction of Argonne National Laboratory.
Supporting Design Analysis of the VTR Using MCNP and TRACE
Published in Nuclear Science and Engineering, 2022
Jack Galloway, Joshua Richard, Cetin Unal
The Versatile Test Reactor (VTR), a sodium-cooled fast reactor (SFR) design, provides an essential function to the U.S. nuclear industry in providing a very high, fast neutron flux for material and reactor designers to design and test new nuclear materials on U.S. soil. Historically, in the United States the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility, along with preceding reactor designs, provided fast flux irradiation facilities to develop new fuel, cladding, and structural materials. The shutdown of all previous fast spectrum reactor irradiation facilities in the United States has caused a significant shortage of potential irradiation sites to test, approve, and refine new nuclear materials, especially those proposed for use in fast neutron spectrums. Since the reactor shutdowns, much of the U.S. industry’s fast spectrum irradiation testing has thus occurred in foreign reactors, such as the BN-600 reactor in Russia. This, however, has put the United States in a precarious position with regard to nuclear technology development, with the bulk of irradiation capabilities needed to satisfy regulatory requirements largely located outside the control of U.S. interests and subject to significant volatility. The VTR will provide the United States with a valuable fast spectrum irradiation capability with highly flexible test chambers for testing fueled and nonfueled experiments, including extended-length test assemblies designed to accommodate the testing of various coolant, structural, clad, and fuel designs in a single location.
Design of an experimental breeder reactor run 138B reactor physics benchmark evaluation management application
Published in Journal of Nuclear Science and Technology, 2020
Ryan Stewart, Edward Lum, Chad Pope
The Experimental Breeder Reactor II (EBR-II) was a sodium-cooled, metal-uranium-fueled fast reactor designed and operated by Argonne National Laboratory (ANL). EBR-II first went critical in 1964, and in 1969 achieved its full power level of 62.5 MWth [1,2]. During the 1960’s, EBR-II’s main goal was to provide groundwork in developing a liquid-metal cooled fast reactor with on-site fuel reprocessing. The 1970’s brought a close to the reprocessing research phase and EBR-II then focused on irradiation capabilities and providing information for fast reactor materials. The 1980’s brought about another major change for the EBR-II facility when it became a test subject for safety analysis in liquid metal cooled reactors. A landmark reactor safety experiment program tested EBR-II’s ability to respond to severe accident conditions such as station blackout without a scram. The most severe tests were conducted during Run 138B and were designed to test a liquid metal cooled reactor’s ability to cope with catastrophic failures in the heat removal systems [3].
Simulation of the Fast Reactor Fuel Assembly Duct-Bowing Reactivity Effect Using Monte Carlo Neutron Transport and Finite Element Analysis
Published in Nuclear Technology, 2021
The Experimental Breeder Reactor II (EBR-II) pioneered many advances in nuclear engineering, including onsite reprocessing of the reactor fuel. These advances have been researched and characterized, but several phenomena have proven extremely difficult to quantify. For example, observation and measurement of EBR-II fuel and reflector assembly bowing as well as attempts to model the behavior and quantify the reactivity effect have been documented.1–5 However, effectively modeling individual fuel assembly duct displacement due to the thermal and mechanical forces generated within the reactor during operation and using the displacement values to calculate the resulting reactivity effect have been elusive.