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A URANIUM THERMOCHEMICAL CYCLE FOR HYDROGEN PRODUCTION
Introduction
Thermochemical processes are promising large-scale methods to produce hydrogen because the energy
input is in the form of lower-cost heat rather than more expensive electricity. Nuclear Energy is a
primary candidate to supply that heat (Forsberg, 2009). In a thermochemical process, heat plus water
yields hydrogen and oxygen, all other chemicals within the process being recycled. A thermochemical
cycle should: i) operate at high thermodynamic efficiency; ii) have a small number of process steps;
iii) not have a significant inventory of volatile or hazardous chemicals that can cause off-site risks;
iv) operate at reasonable temperatures. No existing thermochemical cycle meets all of these criteria
(Brown, 2000). There is a need for better thermochemical cycles.
Uranium exists in many valence states, has overlapping 6d and 5f electrons, and a complex
chemistry – exactly the chemistry one would investigate for development of a thermochemical cycle.
However, there has been almost no consideration of using uranium in thermochemical cycles because
it was thought to be expensive or unusable because of its radioactivity. The nuclear fuel cycle involves
mining natural uranium (0.7% 235 U) followed by enrichment of the uranium (3-5% 235 U) in the 235 U
isotope for use in nuclear fuels. The enrichment process has produced as a by-product over a million
tonnes of excess depleted uranium (~0.3% 235 U) available at little or no cost. The depleted uranium is
primarily 238 U with a half-life of several billion years; i.e. its radioactivity is extremely low. The primary
handling hazard is that it is a heavy metal, not that it is radioactive. Based on these considerations,
there are strong incentives to examine possible uranium thermochemical cycles.
An investigation was undertaken at Oak Ridge National Laboratory to find and develop a
thermochemical cycle based on uranium (Ferrada, 2009; US Patent, 2008). This paper describes the
recent development of a new uranium thermochemical cycle (UTC).
Process description
A series of UTC were investigated using a combination of thermodynamic analysis and experiments.
The thermodynamics of alternative cycles were analysed using Outokumpu’s HSC 5.0 Chemistry
software computer code version 5.11 (Outokumpu, 2002).
As reported elsewhere, the experiments were used to demonstrate the major process steps with
recycle of the uranium. Most experiments were conducted in an alumina-lined tube furnace with the
uranium in an alumina boat. Hydrogen gases were produced and measured. Reduction of CuO to
copper metal, determined by weight loss, was the primary method used to measure the amount of
reductant produced. X-ray diffraction analysis was used in most cases to confirm the uranium species.
The results of thermodynamic analysis and experiments led to the UTC shown in Figure 1 that
uses valence state transitions and the formation of oxidised and reduced uranium species to split H 2 O
for the production of H 2 . Some of the chemical reactions are new but most of the chemical reactions
are used industrially within the uranium industry.
In the hydrogen production step, triuranium octoxide (U 3 O 8 ) is initially reacted with sodium
carbonate (Na 2 CO 3 ) and steam to generate hydrogen and sodium diuranate (Na 2 U 2 O 7 ) at 600°C and
above. This can be done under mild pressures (~1 atmosphere) with an excess of carbon dioxide (CO 2 )
in an inert carrier gas, such as argon (Ar). Although not demonstrated in this work, the hydrogen can
be separated from these gases by selective membranes or pressure swing adsorption and/or cryogenic
separation methods. The steam is condensed to recover heat, and the excess CO 2 and carrier gases are
recycled. Only this first step in the UTC process is unique, the remaining two steps are conventional
processes in the uranium processing industry (Edwards, 2000).
In the second step, sodium diuranate (Na 2 U 2 O 7 ) is converted to uranyl ammonium tricarbonate
[UO 2 CO 3·2(NH 4 ) 2 CO 3 ]. The sodium diuranate is dissolved in an alkaline carbonate solution. This uranium
solution is passed through a column of Dowex 1-X resin. The uranium is loaded onto the ion exchange
resin while the sodium passes through the column. The Na 2 CO 3 can be recovered and recycled from
the uranium-free raffinate. The uranium is then stripped from the resin using a solution of ammonium
carbonate [(NH 4 ) 2 CO 3 ]. The resulting uranium solution is dried to produce uranyl ammonium
tricarbonate [UO 2 CO 3·2(NH 4 ) 2 CO 3 ].
454 NUCLEAR PRODUCTION OF HYDROGEN – © OECD/NEA 2010