backstreet: Molten-Salt Reactor Experiment

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MSRE plant diagram
The Molten-Salt Reactor Experiment (MSRE) was an experimental molten-salt reactor at the Oak Ridge National Laboratory (ORNL); researching this technology through the 1960s. The MSRE was a 7.4MWth test reactor simulating the neutronic "kernel" of an inherently safe epithermal thorium breeder reactor. It used three fuels: plutonium-239, uranium-235 and uranium-233. The last, 233UF4 was the result of breeding from thorium.
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 650 core was shed via a cooling system using air blowers and radiators. It is thought similar reactors could power high-efficiency heat engines such as gas turbines.
The MSRE was located at ORNL. Its piping, core vat and structural components were made from Hastelloy-N and its moderator was a pyrolytic graphite core. It went critical in 1965 and ran for four years. The fuel for the MSRE was LiF-BeF2-ZrF4-UF4 (65-30-5-0.1), the graphite core moderated it, and its secondary coolant was FLiBe (2LiF-BeF2), it operated as hot as 650 and operated for the equivalent of about 1.5years of full power operation.
The result promised to be a simple, reliable reactor. The molten salt reactor experiment was needed to prove the technology. The purpose of the Molten-Salt Reactor Experiment was to demonstrate that some of the 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 10MW(t) or less, with heat rejection to the air via a secondary (fuel-free) salt.
Contents
1 Plant description
1.1 Core
1.2 Fuel/primary coolant
1.3 Secondary coolant
1.4 Pump
1.5 Air-cooled heat exchangers
1.6 Neutronics and thermal-hydraulics
1.7 Building grounds
1.8 Structural alloy Hastelloy-N
1.9 Development and construction timeline
2 Operation
3 Results
4 Decommissioning
5 References
6 See also
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Plant description

MSRE reactor room, top down view
Core

Graphite MSRE 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 micrometre. 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.[1]
Fuel/primary coolant
The fuel was LiF-BeF2-ZrF4-UF4 (64-30-5-1 mole%).
This reactor could breed more of its 233U fuel from thorium. Thorium is at least four times as abundant as uranium in the Earth's crust and at least 500 times as abundant as uranium-235. Compared to conventional light-water reactors, this breeding had the potential to eliminate the difficulty and expense of uranium enrichment, as well as the need for fast breeder reactors to make plutonium fuel from 238U.
Keeping all the breeding and fuel as fluoride salts made it theoretically possible to combine the reactor core and breeding blanket in one fluid, by sculpting the moderator rods. Further, it appeared that the fluid salt would permit on-site chemical separation of the fuel and wastes.
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. Low pressure stability requires less robust reactor vessels and increased reliability. The high reactivity of fluorine traps most fission reaction byproducts.
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-HF to remove oxide, F2 to remove uranium as UF6. Haubenreich and Engel,[2] Robertson,[3] and Lindauer[4] provide more detailed descriptions of the reactor and processing plant.
Secondary coolant

Molten FLiBe
The secondary salt was LiF-BeF2 (6634 mole%).
Pump
The bowl of the fuel pump was the surge space for the circulating loop, and here about 50gal/min of fuel was sprayed into the gas space to allow xenon and krypton to escape from the salt. Removing xenon made the reactor safer and...(and so on)