Irradiated graphite: what are the options?

Above: Quarter Scale Rig at Jacobs’ Technology and Innovation Centre in Birchwood, near Warrington, which measures distortion due to brick cracking and validates the brick modelling software that underpins current and future safety case claim

How much irradiated graphite (i-graphite) is there and where is it?

There are approximately 240,000 metric tonnes of i-graphite around the world. The UK has the largest amount with approximately 70,000 metric tonnes (mainly from its Advanced Gas-cooled Reactors [AGRs] and Magnox reactors), followed by Russia/Ukraine/Lithuania (from its RBMKs), France (from its UNGGs), USA (from its Hanford reactors and two High Temperature Reactors [HTRs]), Germany (from its two HTRs) plus Japan, Italy, Spain (one graphite power reactor each) and a few smaller amounts elsewhere in research reactors. Almost all of the graphite is still in the reactor pressure vessels (RPVs), although discharged AGR and Magnox fuel sleeves are stored in silos. The Windscale AGR (WAGR) in the UK and the Fort St. Vrain HTR (USA) are the only decommissioned cores of any significant size where the graphite has been removed from the pressure vessels. The graphite blocks from WAGR are stored above ground in concrete boxes and those from Fort St. Vrain, which still have the fuel inside them, are stored above ground in steel canisters which are continuously cooled by a natural circulation of air. In Germany, however, instead of removing the graphite from the reactor vessel of the Arbeitsgemeinschaft Versuchs-Reaktor (AVR) as initially planned, it was eventually decided to fill the reactor vessel with low-density cellular concrete. This allowed the complete RPV to be removed and transported horizontally to a nearby interim store, where it remains to this day.

What are the challenges of dealing with this material?

Most of the graphite activity will have decayed away within 60-70 years after a nuclear power plant is permanently shut down, but the long-term challenge is carbon-14 (C-14), as it is a radiological hazard and can be released from graphite waste in gaseous form on leaching. It is therefore classed as intermediate level waste (ILW), although its activity level will be closer to low level waste (LLW) after 60-70 years.

The principal challenge is to find a way to dispose of the material which is safe but also cost-effective. There are a few ways of removing some or most of the C-14 absorbed in the pore structure of the graphite, but these have only been trialled at laboratory scale. However, there is no certainty that these processes will reduce the C-14 inventory to a sufficient extent to reduce the overall ILW waste volumes of the graphite by changing it into LLW. In addition, they have not been trialled on the industrial scale needed to deal with the quantity of graphite in a typical power reactor core (>1,000 metric tonnes).

Above: Leningrad 1, was the first of Russia’s 15 RMBK-1000 reactors to permanently shut down in December 2018 (Photo credit: Rosenergoatom)

What experience does the industry have of dismantling graphite reactors/treatment of i-graphite?

As mentioned, this is limited to WAGR and Fort St Vrain and also a few small low power reactors such as JASON and the Graphite Low Energy Experimental Pile (GLEEP) reactor, both in the UK, and the Brookhaven Research Reactor in the USA. However, extensive studies for dismantling a number of graphite cores have been carried out in the past such as the Windscale Piles and the Tokai Mura reactor in Japan.

What are the treatment options for i-graphite?

The most important treatment option for graphite from a long-term disposal point of view is to try to remove as much of the adsorbed C-14 in the pore structure as possible. Combinations of heat treatment and ‘flushing’ the pores using liquids have been trialled, but these have only been done at laboratory scale. Sealing the outside of individual blocks using resin coatings or similar, which would effectively lock the C-14 in, has also been trialled but the long-term behaviour of the sealant over many thousands of years is not known. Some graphite, for example the Windscale Pile graphite, has Wigner energy stored in it, so that would also need releasing before putting it in a concrete box, for example, for long-term disposal. Other ways of processing the graphite are being considered, such as pyrolysis, which effectively converts the graphite into a slurry, but as with all these options, this results in a much larger volume of waste to deal with.

Where are the key areas that need further R&D/study/demonstration?

The prime areas for further research should be targeted to allow informed decisions on graphite waste management to be made through a deeper understanding of the role of C-14. This should include a better understanding of the specific activity of deposits in the porosity of the graphite, which are removable, compared to the C-14 in the bulk material, which is not. The C-14 removal process should be optimised together with the ways of dealing with the associated waste produced.

What work is Jacobs doing in this area?

Jacobs has unique graphite knowledge, gained from its heritage as designer and architect engineer of the Magnox and AGR fleets. For EDF, operators of the UK’s AGR fleet, it supports lifetime safety cases with inspection data analysis and stress and damage tolerance code development. In the UK, Jacobs also operates a suite of graphite-related rigs, including a quarter scale core rig, at its Technology and Innovation Centre in Birchwood, and its Waste Management team at Harwell draws on this long history of working with graphite to support technical programmes for safe and timely characterisation, retrieval, waste treatment, storage and geological disposal.

Since 2003, Jacobs has been measuring releases of C-14 from graphite samples on leaching with a particular focus on the potential for gaseous release. Samples have been studied from WAGR, the British Experimental Pile 0 (BEPO) and the Oldbury Magnox reactors. Jacobs has also supported assessments of releases of radioactive gases from graphite and other waste materials in a geological disposal facility and continues to work on improving our understanding of the options for disposal of graphite.

Biography

Mike Davies, a nuclear engineer for more than 39 years, is an internationally recognised specialist on graphite core structures. He provides expertise and support to the Graphite Core Project Team in EDF and has been involved with programmes relating to the development of a European HTR and the Gemini Plus HTR project. He was a member of the Gen IV VHTR Graphite Working Group, a member of the ASME committee responsible for developing a design code for the graphite cores in future HTRs, and was recently appointed as the deputy UK member of the GenIV VHTR System Steering Committee.