An international coordinated research project on the treatment of irradiated graphite has identified a number of unresolved scientific and technical issues. By Michael I Ojovan and Anthony J. Wickham

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Dismantling old reactors and the management of radioactive graphite waste are becoming an increasingly important issue for a number of International Atomic Energy Agency (IAEA) member states. Exchanging information and research co-operation can help in resolving identical problems faced by different institutions and improving waste management practices, efficiency, and general safety.

The IAEA Coordinated Research Project (CRP), ‘Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal’, was conducted by IAEA in the period 2010-2014 and involved 24 organisations from 10 member states. The CRP explored innovative and conventional methods for graphite characterisation, retrieval, treatment, and conditioning. The results, summarised in an IAEA technical document to be published soon, have identified a number of unresolved scientific and technical issues. Operators should:

  • Improve scientific understanding of the creation, chemical form, location, and release behaviour of radionuclides;
  • Improve predictive models of radioisotope behaviour;
  • Improve understanding of speciation and its impact on transport models;
  • Ensure that sampling programmes are statistically representative of the totality of the graphite to be disposed;
  • Establish an accurate radionuclide inventory;
  • Consider novel alternative dismantling and treatment strategies.

Graphite is used in nuclear reactors primarily as a neutron moderator and reflector, a structural material, and a fuel-element matrix material. It has been deployed in about 250 uranium-(or UO2)-graphite reactors, including the UK’s magnox and advanced gas-cooled reactors (AGRs), France’s UNGG, a small number of high-temperature reactors (HTRs), the Soviet-era RBMKs and numerous ‘production’ and materials-testing reactors. Many of those reactors have seen more than 40 years’ service, and about half of them are shut down.

Graphite is a porous, chemically inert material, highly conductive and resistant to corrosion, in general retaining its properties after exposure to an intense radiation field and at high temperatures. However, it undergoes structural changes as a result of exposure to fast neutrons, resulting in dimensional changes in components and significant changes in their mechanical and physical properties.
In reactors where the graphite has been exposed to an oxidising coolant, some degree of oxidation of the material induced by the ionising radiation field will have taken place, potentially affecting its strength. More significantly, certain impurities, and the 1.1% of carbon-13 naturally present in the graphite, become activated through interaction with slow neutrons and present a significant radiation hazard post-exposure, which must be accommodated in subsequent dismantling and disposal.

The ability to dismantle and remove graphite stacks has already been demonstrated in a small number of different reactor designs, e.g. Fort St Vrain (a prototype HTR), the air-cooled Brookhaven research reactor (both in the USA), and the GLEEP research reactor and Windscale prototype AGR (both in the UK).

Most of the resulting irradiated graphite waste (i-graphite) awaits acceptable treatment and disposal solutions whilst residing in temporary storage facilities. The exception is material from GLEEP, which was subject to very low irradiation and could be disposed of by an industrial process. The global inventory of i-graphite is currently in excess of 250,000t (see Figure 1).

A clear direction on the disposal of all radioactive wastes is now accepted as a necessity to underwrite the case for further development of nuclear power operations worldwide. In a ‘greener’ world, the need for baseload generation and backup for renewables clearly marks out a role for nuclear power, and it is urgently necessary to reverse its decline (currently 13% overall worldwide). Thus, work on all aspects of radioactive waste is viewed by IAEA as important and timely.

When the IAEA initiated the graphite CRP in 2010, it aimed to complement the EU Carbowaste Project and to provide a platform for a number of the participants who are common to both projects to continue their studies after Carbowaste funding finished. In addition, the CRP involved member states and organisations, which were not part of Carbowaste – notably research groups from Russia, Ukraine and China. A total of 24 organisations participated from ten member states: China, France, Germany, Lithuania, Russian Federation, Spain, Switzerland, Ukraine, the UK and the USA.

The project included: alternative techniques for removing the graphite from the pressure vessel or containment; direct chemical or physical treatments, including pre-treatment to reduce the radioisotope content and improve the economics and radiological safety of the following process operations; and treatment of the products of processes intended to improve radiological safety or economics (such as separation and recycling of useful isotopes for the nuclear and/or medical industries). Issues also included evaluating the possibilities for disposal of 14CO2 (produced by the oxidation of reactor graphite) by dilution either in the atmosphere, or with carbon-14 depleted CO2 from burning fossil fuels prior to sequestering in exhausted gas fields and oil shale deposits (known as carbon capture and storage).

The research programme aimed to identify and evaluate, by practical means where possible, handling and treatment technologies that may be of assistance in graphite-disposal strategies, complementary to work done under Carbowaste. This included all stages of the decommissioning process: removing graphite from the reactors; potential processes and treatments; disposing of the solid graphite; and disposal using innovative methods such as ‘disperse and contain’ in the form of carbon dioxide.

The participating member states’ organisations had differing interests, broadly classified into four groups:

  • characterisation of the i-graphite (for radioisotope distribution and content and physical/mechanical properties);
  • processing (i.e. leaching, washing and heat treatments for isotope mobilisation and recovery, and alternative removal procedures such as ‘nibble and vacuum’, which crumbles the graphite in situ and then moves it out of the reactor vessel by vacuum transport rather than the more usual assumption of removal as intact blocks);
  • immobilisation (of isotopes within the solid material and of the solids themselves);
  • disposal (shallow and deep repository options, plus more radical alternatives such as CO2 sequestering).

A schematic of the CRP work and the interactions between various groups and topics is shown in Figure 2.

The overall objective of the CRP was to advise member states of the various options currently being researched (Figure 3), to enable them to make an informed decision on the correct policy for their situation. The aim was not to recommend specific methodologies on a global basis, since very different conditions and constraints apply in different member states. As examples, Andra in France is obliged by law to find a solution to the disposal of graphite wastes within the next few months, whereas in the UK the policy is to leave reactors intact in ‘safe storage’ for up to 130 years before eventual dismantling. Germany has severe constraints on the quantities of carbon-14 which may be disposed of in the identified repository (Konrad), while other member state authorities view the presence of chlorine-36 to be a much more significant issue than carbon-14.

The many delays and complications surrounding suitable deep and surface or near-surface repository sites in a number of member states are leading to temporary storage facilities being established, where the benefits of making progress must be weighed against the short-term safety issues. They are also the catalyst for investigation of alternative options for handling the graphite. These include, in some cases:

  • re-establishing investigations into methods which have previously been studied (e.g. incineration);
  • extending the remit and conditions suitable for including i-graphite into existing processes (e.g. pyrolysis, Thor);
  • conversion of the graphite into CO2, followed by admixing with fossil-fuel emissions and re-injection into exhausted hydrocarbon reservoirs in suitable geological strata, exploiting the philosophy of ‘dilute and contain’.

A major outcome of the CRP was recognition that i-graphite is a unique waste form, which deserves more comprehensive consideration than mere classification as (usually) intermediate-level waste (ILW). The various treatments and potential processes under consideration may offer considerable savings in final waste volume, in ultimate disposal costs and (where deemed desirable) in timescale.

It was also recognised that i-graphite from different sources (different reactor types and different reactors of the same type) had very different characteristics in terms of radioisotope content. This distribution was very variable between different components in the same reactor and even spatially within a single component. Equally important is the understanding that handling during dismantling may not be straightforward, for example where components have been distorted due to irradiation damage by fast neutrons; or where radiolytic oxidation in the gas coolant has led to distributions of mass loss. This is now approaching 50% locally in the oldest UK AGRs.

More work is needed to establish the size and nature of sampling programmes to ensure that data of sufficient accuracy are available to waste authorities on the isotope content and other parameters relevant to handling. The reliability of calculation routes towards isotope inventories is also related to the required size of sampling programmes. Some excellent work has been done in this area, but it needs further attention.

A number of important issues were raised with regard to long-lived beta-emitting isotopes, which are present in graphite. Other gaps in the scientific understanding of radioisotope formation and behaviour considered to need further attention included the potential role of microbiological processes in the capture by i-graphite of potential radioisotope precursors and in the potential release of those radioisotopes.

The CRP recommended that there should be better coordination between laboratory studies and sampling and measurement programmes. In addition, given the almost universal delays in establishing geological disposal facilities, attention should continue to be given to developing alternative dismantling and treatment strategies, so the best available ‘toolbox’ for handling i-graphite is available to member state waste authorities.

Two important initiatives are continuing within IAEA in the area of i-graphite. The first is a repository for data and reports within the Immonet knowledge network, accessible via an IAEA ‘Nucleus’ registration. The second is a new project ‘Grapa’ (‘Disposal of i-Graphite: Joint Project on Industrial Applications’) as part of the IAEA International Predisposal and Decommissioning Networks, in which seven member states are currently participating. This seeks to continue the i-graphite work and to support establishing a demonstration pilot plant for graphite removal and treatment, possibly in collaboration with the EU Eighth Euratom Research Framework.

Other proposed activities include widening the scope of investigations of:

  • graphite removal (e.g. water-jet cutting as a possible alternative to ‘nibble and vacuum’, especially where underwater dismantling is envisaged, such as France);
  • heat treatments (e.g. using microwave and electric-arc heating);
  • optimising packaging and transport containers;
  • evaluating of the unique issues of disposing of Chernobyl graphite to a repository sited within the exclusion zone; and
  • collaborative evaluation of issues specific to certain reactor designs, such as the Wigner energy present in some low-temperature-irradiated components in RBMK reactors.

Consideration of the behaviour of weak beta emitters will also continue alongside the current EU ‘Carbon-14 Source Term’ (Cast) programme. Characterisation data will also continue to be acquired. It is hoped that this new venture will receive formal approval to proceed in February 2016, with oversight of i-graphite-management activities internationally continuing indefinitely through the IDN and Immonet. ¦

About the authors

Michael I. Ojovan, Waste Technology Section, Division of Nuclear Fuel Cycle and Waste Technology, Department of Nuclear Energy, International Atomic Energy Agency, PO Box 100, Wagramerstrasse 5, Vienna, A-1400 Austria

Anthony J. Wickham, Nuclear Technology Consultancy, Cwmchwefru, Llanafanfawr, Builth Wells, LD2 3PW, UK, and School of Mechanical, Aerospace and Civil Engineering, The University of Manchester, Manchester M13 9PL, UK