The Molten-Salt Reactor Experiment (MSRE) was an experimental molten salt reactor research reactor at the Oak Ridge National Laboratory (ORNL). This technology was researched through the 1960s, the reactor was constructed by 1964, it went critical in 1965, and was operated until 1969. The costs of a cleanup project were estimated at about $130 million.
The MSRE was a 7.4 MWth test reactor simulating the neutronic "kernel" of a type of inherently safer epithermal thorium breeder reactor called the liquid fluoride thorium reactor. It primarily used two fuels: first uranium-235 and later uranium-233. The latter 233UF4 was the result of breeding from thorium in other reactors. Since this was an engineering test, the large, expensive breeding blanket of thorium salt was omitted in favor of neutron measurements.
In the MSRE, the heat from the reactor core was shed via a cooling system using air blown over radiators. It is thought similar reactors could power high-efficiency heat engines such as closed-cycle gas turbines.
The MSRE's piping, core vat and structural components were made from Hastelloy-N and its moderator was a pyrolytic graphite core. The fuel for the MSRE was LiF-BeF2-ZrF4-UF4 (65-29.1-5-0.9 mole %), the graphite core moderated it, and its secondary coolant was FLiBe (2LiF-BeF2), it operated as hot as 650 °C and operated for the equivalent of about 1.5 years of full power operation.
The result promised to be a simple, reliable reactor. The purpose of the Molten-Salt Reactor Experiment was to demonstrate that some key features of the proposed molten-salt power reactors could be embodied in a practical reactor that could be operated safely and reliably and be maintained without excessive difficulty. For simplicity, it was to be a fairly small, one-fluid (i.e. non-breeding) reactor operating at 10 MWth or less, with heat rejection to the air via a secondary (fuel-free) salt.
Reactor description
Core
The pyrolytic graphite core, grade CGB, also served as the moderator.
Before the MSRE development began, tests had shown that salt would not permeate graphite in which the pores were on the order of a micrometer. Graphite with the desired pore structure was available only in small, experimentally prepared pieces, however, and when a manufacturer set out to produce a new grade (CGB) to meet the MSRE requirements, difficulties were encountered.
Fuel/primary coolant
The fuel was 7LiF-BeF2-ZrF4-UF4 (65-29.1-5-0.9 mole %).
The first fuel was 33% 235U, later a smaller amount of 233UF4 was used.
By 1960 a better understanding of fluoride salt based molten-salt reactors had emerged due to earlier molten salt reactor research for the Aircraft Reactor Experiment.
Fluoride salts are strongly ionic, and when melted, are stable at high temperatures, low pressures, and high radiation fluxes. Stability at low pressure permits less robust reactor vessels and increases reliability. The high reactivity of fluorine traps most fission reaction byproducts.
It appeared that the fluid salt would permit on-site chemical separation of the fuel and wastes.
The fuel system was located in sealed cells, laid out for maintenance with long-handled tools through openings in the top shielding. A tank of LiF-BeF2 salt was used to flush the fuel circulating system before and after maintenance. In a cell adjacent to the reactor was a simple facility for bubbling gas through the fuel or flush salt: H2-hydrogen fluoride to remove oxide, fluorine to remove uranium as uranium hexafluoride. Haubenreich and Engel, Robertson, and Lindauer provide more detailed descriptions of the reactor and processing plant.
Secondary coolant
The secondary salt was LiF-BeF2 (66–34 mole %).
Pump
The bowl of the fuel pump was the surge space for the circulating loop, and here about of fuel was sprayed into the gas space to allow xenon and krypton to escape from the salt. Removing the most significant neutron poison xenon-135 made the reactor safer and easier to restart. In solid-fuel reactors, on restart the 135Xe in the fuel absorbs neutrons, followed by a sudden jump in reactivity as the 135Xe is burned out. Conventional reactors may have to wait hours until xenon-135 decays after shutting down and not immediately restarting (so-called iodine pit).
Also in the pump bowl was a port through which salt samples could be taken or capsules of concentrated fuel-enriching salt (UF4-LiF or PuF3) could be introduced.