375089 Development of on-Line Pyroprocessing for Liquid Thorium Fueled Reactors

Monday, November 17, 2014: 4:50 PM
M103 (Marriott Marquis Atlanta)
Milan Stika, Metallurgical Engineering, University of Utah, Salt Lake City, UT

The Molten Salt Reactor (MSR) is a high temperature, liquid fuel, fluoride salt based nuclear reactor concept. It employs a thorium-uranium fuel cycle and is designed to sustain its energy production via breeding the fissile U-233 from fertile Th-232. The MSR was initially developed and successfully operated at Oak Ridge National Laboratory (ORNL) from the 1960s to the early 1970s. The program was cancelled in 1972 in favor of focusing support onto the liquid sodium cooled Integral Fast Reactor (IFR), development of which was eventually stopped in the mid-90’s due to proliferation concerns and poor economic viability. While sodium cooled fast reactors such as the IFR are still being considered for future development by the Department of Energy, the MSR concept has also gained renewed interest from a variety of areas.  Notably, this includes the commercial electric utility sector.

Many people would agree that a departure from conventional nuclear power technology is needed to address problems with waste management, safety, non-proliferation, and resource sustainability.  The MSR is experiencing the revival of interest because it holds the promise of delivering in all these areas.  Relative to uranium, thorium is an abundant and cheap nuclear fuel. Given that there is substantially more recoverable fertile thorium (practically monoisotopic Th-232) in the earth than U-235, the MSR would be much more sustainable than current U-235 fission systems.

Another advantage is improved safety from low pressure operations and from the fact that the molten salt is simultaneously a fuel carrier and a coolant. The core runs at a higher temperature which enables the substitution of a traditional steam turbine with a higher-efficiency helium turbine. This results both in increased electricity generation efficiency and in inherently safer system, since less thermal power is present and required to be cooled down.

The system is relatively proliferation resistant due to the hard gamma radiation that accompanies the fissile U-233.  And, very importantly, there is much reduced generation of plutonium and other transuranics which minimizes the demands on geologic repository performance. Ideally, MSRs could produce waste that would not require building of a deep geological repository, since the waste would mainly comprise of fission products and its activity would be back to background levels in about 300 years.

All these features are enabled if the reactor’s fuel is constantly being reprocessed by various pyrochemical and pyrometallurgical processes (on-line pyroprocessing). There are several methods suitable for on-line pyroprocessing. This includes the fluorination of salt containing UF4 to obtain gaseous UF6 (fluoride volatility process). Other methods include the salt/bismuth reductive extraction, electrochemical separation, and vacuum distillation.

There are two main goals of on-line pyroprocessing – to clean the fuel of neutron poisons (lanthanides); and to isolate protactinium – the newly bred precursor to fissile uranium. A more subtle objective of pyroprocessing is to return actinides back into reactor for incineration and to prevent them from reaching waste streams designated for lanthanides (fission products).

In order to assess the whole processing facility, mass flow rates for all unit operations under different configurations were estimated. An integrated process model is being developed that includes both reactor core performance and salt recycling steps to accurately capture the performance of the facility and to identify bottlenecks and other problematic areas. Preliminary calculations from this model will be presented and discussed.  Knowledge gaps for improving the model’s predictive capability will also be discussed.

In the experimental domain, one particular unit operation – the reductive extraction – is being investigated. This step involves separation of small amounts of uranium and protactinium from large amounts of thorium in molten LiF-BeF2. Either lithium or thorium metal can act as a reductant. Electrochemical sensors are proposed to monitor the composition of the molten salt in this and other unit operations in the system. Laboratory scale experiments have been setup, and preliminary voltammetry data in LiF-CaF2 system will be presented.


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