Reviving a 40-year-old nuclear reactor coolant
The MIT NRL is involved in two collaborative research projects that are aimed at a high-tech revival of a nuclear reactor coolant that has not been used since 1969. The goal of the NRL’s research in this area is to help the development of nuclear reactor materials for a novel fluoride-salt-cooled, high-temperature reactor, the FHR. Work underway at the NRL includes in-core, high-temperature tests of structural materials submerged in the fluoride salt – the first such tests in over 40 years. Currently, the FHR-related experiments at the NRL are the only ones of their kind.
The first collaborative project is in its second three-year period of U.S. Department of Energy-funded research among MIT, the University of Wisconsin-Madison, the University of California at Berkeley, and the University of New Mexico at Albuquerque. The second, newer NRL collaboration is with the Shanghai Institute of Applied Physics, a collaboration enabled by a Memorandum of Understanding between the US Department of Energy and the Chinese Academy of Sciences.
The idea of a salt-cooled reactor is not new, but the current FHR design combines liquid salt and more recently developed technologies with the goal of achieving a new reactor type that has superior economic, safety, waste, nonproliferation, and physical security properties compared to any existing water-cooled power reactor.
The most-recently operated high-temperature, salt-cooled reactor was the Molten-Salt Reactor Experiment (MSRE) that ran from 1965-1969 at the Oak Ridge National Laboratory. The coolant in the MSRE was primarily a mixture of lithium fluoride (LiF) and beryllium fluoride (BeF2), known as flibe. The reactor’s fuel was in the form of uranium tetrafluoride (UF4), and was mixed directly into the liquid coolant. The MSRE operated at a temperature of about 650°C, and successfully demonstrated the feasibility of a high-temperature, salt-cooled reactor.
Flibe is also the coolant proposed for the new FHR designs. Flibe is a good heat transfer fluid, and has a melting temperature of 459°C and a boiling temperature of 1430°C. The vessel for an FHR does not need to pressurized, unlike the vessel for a water or gas-cooled reactor. Flibe is clear when liquid, which is an important benefit over other non-water coolants and allows easier reactor maintenance and instrumentation. It also has very good nuclear properties, as it will not absorb many of the neutrons needed to sustain the nuclear reaction.
The other major technologies that are combined in the current baseline FHR designs are ceramic coated-particle fuels with failure temperatures of about 1650°C, a passively-safe reactor design, and a nuclear air-Brayton combined cycle for electricity generation. The combination of a high-temperature fuel and a high-temperature, low-pressure coolant significantly reduces the possibility of fuel damage, even with the loss of all cooling systems.
The high-efficiency air-Brayton cycle is similar to the systems used in current natural gas combined cycle plants. The FHR could further combine its nuclear-heated base load capability with natural gas-heated co-firing to provide quick on demand peak-load electricity, and can use this unique ability to increase plant revenue by 50% relative to base load nuclear plants with costs similar to or lower than existing nuclear power systems.
The current and planned research at the NRL is designed to test the interaction of liquid flibe and its byproducts with several likely FHR structural materials. Experiments include tests of two metals (316 Stainless Steel, and Hastelloy® N), surrogate ceramic fuel particles, various graphites, carbon-carbon composites, and silicon carbide samples in 700°C flibe, and in the radiation field of the MIT Reactor core. Parallel tests for some of these experiments have been run at UW without radiation exposure.
Once the sample materials have been removed from the experimental capsule, they are initially examined for weight loss or gain, and basic surface characteristics. Scanning electron microscopy (SEM) of the materials will be used to look for corrosion effects such as pitting and crud deposition, as well as structural defects and cracking, and elemental distributions. Materials may be cross-sectioned to examine the depth of attack by corrosion, salt penetration, penetration of corrosion products, and the formation of boundary layers. The solidified flibe from each of the compartments is also tested for components of the structural materials such as nickel and chromium which may have dissolved into the salt.
A second branch of examinations for these experiments is related to the transport and fate of the tritium that is produced when the flibe is irradiated with neutrons. Tritium is a radioactive isotope of hydrogen (hydrogen-3), which decays with a half-life of 12.3 years by emitting a low energy beta-particle. External exposure to tritium is not generally dangerous because its beta particles are unable to penetrate the skin. However, it can form tritiated water (HTO) or other tritiated compounds, so it can pose a radiation hazard when inhaled, ingested via food or water, or absorbed through the skin.
The primary nuclear reaction that produces tritium in the flibe occurs when lithium-6 absorbs a neutron and decays into an alpha particle (helium-4 nucleus) and a tritium nucleus. The flibe used for the tests conducted at the MIT Reactor was originally used as the secondary coolant for the MSRE; it is enriched in the isotope lithium-7 to 99.99% (natural lithium is 92.4% lithium-7 and 7.6% lithium-6). This enrichment greatly reduces the number of neutrons captured by the flibe and the amount of tritium it produces.
At the high temperatures of an FHR, and in the MIT experiment, tritium easily passes through most materials. Prior experiments and data from the MSRE indicate that a large fraction of the tritium may be absorbed by the graphite in the reactor. The partitioning of tritium among potential reactor components, and methods for preventing tritium release to the ambient environment are critical design considerations for any future FHR. Tritium transport, absorption, and release will be measured from the various materials contained in the MIT experiment.
The path forward to either a test or commercially practical FHR is still fairly long and has many challenging components. Current, unique research at the MIT NRL is providing data to address critical FHR materials and design issues that can only be resolved experimentally.
Photo Caption: On the left, a graphite sample holder for the high-temperature, in-core tests was filled with flibe (visible in each of the six sample compartments) in an argon filled glove box; two of the sample compartments are metal lined. The graphite sample holder was sealed into a nickel capsule which included temperature sensors, and gas sampling tubes. In the center; a nickel capsule is loaded into the top of the MIT Reactor’s In-Core Sample Assembly tube; the bottom of this tube sits in the MIT reactor’s core (right-hand image).