Spent energy - reprocessing of spent oxide nuclear fuel

Materials World magazine
,
9 Feb 2013

Dr Carsten Schwandt from the Department of Materials Science and Metallurgy at the University of Cambridge, UK, looks at the reprocessing of spent oxide nuclear fuel.

Nuclear fission for large-scale electricity generation has been an established technology since the mid-1950s, when the first generation of commercial reactors came online. Many industrial countries have, notwithstanding occasional setbacks, voted in favour of the continuation and expansion of nuclear power to meet their growing energy demands with non-carbon technologies. Most of the current R&D activities take place under the auspices of the Generation IV International Forum (GIF, www.gen-4.org), whose 13 member states and federations have set out four overriding objectives for the fourth generation of nuclear energy systems currently being developed:

  • safety and reliability
  • sustainability
  • viable economics
  • proliferation resistance and physical protection


However, nuclear technologies are strongly influenced by political credos. Countries such as Germany, Belgium, Switzerland and Mexico have opted out of the use of nuclear fission, and countries such as Australia and New Zealand have never embarked on it. Yet other countries conduct their development programmes independently and with varying levels of transparency.

Storage solution

Most contemporary nuclear reactors employ the fissile uranium-235 isotope as the fuel, which is enriched from its natural abundance of 0.7% to about 3% in a bulk of the virtually non-fissile uranium-238 isotope. The uranium is generally used in its oxide form (UO2 or U3O8), pressed into cylindrical discs and stacked inside metallic rods. During fuel irradiation, the uranium-235 undergoes fission and forms a variety of lower-mass fission products. Uranium-238 is moderately fertile and so a fraction of it undergoes a breeding process, involving a single neutron capture and a double beta-decay, which leads to fissile plutonium-239. This then yields various fission products. Energy generation becomes inefficient after a small percentage of the uranium has been consumed, which necessitates replacement of the fuel rods, typically after three years. The spent oxide nuclear fuel therefore contains uranium oxide as its majority constituent and plutonium oxide as a significant minority constituent. Also present are the oxides of the fission products – mainly caesium, strontium, zirconium, technetium, iodine, the minor actinides and the lanthanides.

The reprocessing of spent oxide nuclear fuel is imperative in the development of sustainable nuclear energy technologies, as the amount of waste requiring long-term storage in underground repositories must be minimised. Currently, the preferred process is the hydrometallurgical plutonium and uranium recovery by extraction (PUREX), in which the spent fuel is dissolved in aqueous nitric acid and the metals are recovered by organic solvent extraction.

Although the recovered uranium can be recycled back into uranium fuel, economic considerations decide whether this is actually done. The plutonium gained can be processed into a mix of UO2 and PuO2, known as mixed oxide (MOX) fuel, or is added to the existing stockpiles. The plutonium may be suitable for use in weapons – indeed, this was the original motivation for the development of reprocessing routes – but the GIF objective on proliferation resistance mandates a clear paradigm shift. The other actinides can also be recycled into nuclear fuel and thereby close the fuel cycle. Reprocessing spent fuel is expensive but is performed in most countries that exploit nuclear energy – the US is currently a notable exception, despite active research.

The salt route

The conventional pyrometallurgical molten salt route of reprocessing spent oxide nuclear fuel is a multistep one:

1. The spent fuel rods are declad and segmented, volatile constituents evaporated, and the remainder brought into contact with molten lithium chloride (LiCl) containing a quantity of dissolved lithium metal (Li) at 650°C in an inert atmosphere. The various metal oxides (MeOx) in the fuel are chemically reduced by the Li to the corresponding metals (Me), and the lithium oxide (Li2O) by-product remains dissolved in the LiCl flux:

MeOx + 2x Li (in LiCl) = Me + x Li2O (in LiCl)

2. The LiCl melt is regenerated by electrolysing the dissolved Li2O, so that a solution of Li in LiCl is re-formed. The use of a carbon anode leads to carbon dioxide gas as the anode product, while the use of an inert anode provides oxygen:

Li2O (in LiCl) + ½ C = 2 Li (in LiCl) + ½ CO2

or

Li2O (in LiCl) = 2 Li (in LiCl) + ½ O2

3. The individual metals are extracted from the reduced metal mix by electrorefining in a eutectic melt of lithium chloride and potassium chloride (LiCl/KCl) at 500°C.

More recent developments in the pyrometallurgical route are the in situ electrochemical generation of Li in the LiCl flux directly at the metal oxide surface and the combination of oxide reduction and flux regeneration into one step.

Global effort

There are substantial R&D programmes worldwide aimed at advancing the pyrometallurgical reprocessing of spent oxide nuclear fuel. For several years, numerous studies have been carried out by the Argonne National Laboratory (ANL) in the US, the Central Research Institute of Electric Power Industry (CRIEPI) in Japan, the Korea Atomic Energy Research Institute (KAERI) in South Korea, the Idaho National Laboratory (INL) in the US and, more recently, the Indira Gandhi Centre for Atomic Research (IGCAR) in India. These efforts indicate that there is a strong global belief in the practicability of the molten salt route.

In the UK, the REFINE Research Consortium has recently been incorporated, with the intention of providing a coordinated, multidisciplinary R&D programme into the reprocessing of spent nuclear fuel. The Consortium, established with the support of the EPSRC, includes departments from five academic institutions – the Universities of Edinburgh, Manchester, Nottingham, Cambridge and University College London – as well as the UK National Nuclear Laboratory (NNL). It brings together a unique and complementary combination of expertise and infrastructure in the relevant areas of chemistry, materials science, chemical engineering and microfabrication. The Consortium’s main technical objectives are to understand and optimise the fundamental processes required for devising a sustainable closed-loop nuclear energy cycle and constructing viable processing facilities, as well as to educate and train the next generation of researchers in the field.

The R&D scope of the Consortium falls under three distinct themes:

  1. Direct electrochemical reduction of spent oxide nuclear fuel – Traditionally, the molten salt route has focused on the use of LiCl melts, in a two-step process of chemical oxide fuel reduction and electrochemical salt flux regeneration. Fundamental improvements would be reached by harnessing the recently developed FFCCambridge process, which was pioneered by Professor Derek Fray from the University of Cambridge, UK (see October 2012 Materials World). In this process, the oxide-to-metal conversion occurs in a single electrochemical step. This is achieved by making the oxide the cathode in a bath of calcium chloride (CaCl2) and applying a potential that is sufficient to decompose the oxide but insufficient to decompose the electrolyte. The process would render spent fuel reduction more compact and minimise waste. While CaCl2 is generally the preferred electrolyte for the FFC-Cambridge process, it is also feasible with LiCl – the molten salt of choice for the pyrometallurgical route. FFC-type reductions of spent fuel in molten LiCl at pilot plant level have already been reported by KAERI.
  2. Electro-refining of reduced fuel – The recovery of the individual metals from the reduced spent oxide nuclear fuel can be achieved by means of electro-refining. The method is well established for speciation purposes in many areas of metallurgy and is also regarded as the preferred option in the nuclear sector. Optimisation is required concerning the composition of the electrolyte and the dissolution and deposition behaviour of the spent fuel constituents.
  3. Analysis techniques – Processes involving nuclear materials demand reliable monitoring and control systems, to guarantee safe operation and high proliferation resistance. It is expected that such systems can be devised by making use of sensing and spectroscopic techniques. Sensor development focuses on innovative in-line electroanalytical systems with micro-electrodes of adequate detection performance, stability and longevity. Spectroscopic methods, such as nuclear magnetic resonance, X-ray absorption and Raman, are employed for the in situ investigation of processes and the identification of intermediate species.


It is evident that these lines of research will be of huge importance to the further advancement of safe nuclear energy technologies.

Further information

Professor Andrew Mount, University of Edinburgh, Lead Investigator of the REFINE Research Consortium, a.mount@ed.ac.uk

Dr Carsten Schwandt, of the Institute’s Materials Chemistry Committee, cs254@cam.ac.uk