THORIUM AS AN ENERGY SOURCE - Opportunities for Norway ...

5.3.2.4 Gas Turbine-Modular Helium Reactor (GT-MHR)
Nuclear Reactors for Thorium
The Gas Turbine-Modular Helium Reactor (GT-MHR) was developed in the United States by
General Atomics using a prismatic fuel. The use of helium as a coolant at high temperature and
the relatively small power output per module (600 MWth) permit direct coupling of the reactor to a
gas turbine (a Brayton cycle). This feature results in power generation at 48 % thermal efficiency
(50 % higher than in a conventional nuclear reactor in use today). The GT-MHR core can
accommodate a wide range of fuels, including highly enriched uranium/thorium, U-233/Th, and
Pu/Th. The use of highly enriched uranium/thorium fuel was demonstrated in General Atomics’
Fort St. Vrain reactor in Colorado.
5.3.3 Molten Salt Reactor Experiment (MSRE)
This advanced breeder concept circulates the fuel in molten salt, without any external coolant in
the core. The primary circuit runs through a heat exchanger which transfers the heat of the
fission reaction to a secondary salt circuit for steam generation. It was studied in depth in the
1960s and is now being revived because of the availability of advanced technology for the
materials and the components.
During the 1960s the USA developed the Molten Salt Breeder Reactor (MSBR) as the primary
back-up option for the conventional fast breeder reactor and a small prototype was operated. This
prototype, the Molten Salt Reactor Experiment (MSRE, 8 MWth) was operated at ORNL between
1965 and 1969. Three fuel types were tested: uranium enriched to 30wt% with U-235, pure U-233
and Pu-239. The fuel salt consisted of 66 % LiF, 29 % BeF 2, 5 % ZrF 4 and 0.2%UF 4.In the USA, a
project to produce a molten salt breeder was initiated in 1971. The electricity output was planned
to be 1000 MW. The salt was composed of 71 % LiF, 16 % BeF2, 12 % ThF4 and 0.3 % UF4. The
reactor was graphite moderated and had two core zones: One fissile zone where most of the power
was produced, and one fertile zone where the conversion from Th-232 to U-233 would take place.
The breeding ratio of this reactor was estimated at 1.06, which means that it would produce 6 %
more fissile material than it consumed. The salt reprocessing was quasi-continuous with a
processing time of only ten days, which seems somewhat unrealistic. More recent calculations
have concluded that the proposed reactor concept had slightly positive reactivity feedback
coefficients, which is undesirable for safety reasons. The project was terminated in 1976.
5.3.4 Conclusion
All projects using thorium in their fuel cycles had been terminated by the 1980s (except in India).
The reason for this seems to be threefold: (1) the thorium fuel cycle could not compete
economically with the more well-known uranium cycle, (2) a lack of political support in many
countries for the development of nuclear technology in the aftermath of the Chernobyl accident
and (3) an increased concern worldwide regarding the proliferation risk associated with the
reprocessing of spent fuel.
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