Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants and in many research and plutonium-producing reactors, in quantities ranging from a few kilograms to more than 3000 tonnes depending on the design. In a number of reactor designs it is also used as a fuel-sleeving material, leading to the generation of large amounts of less-irradiated but still significantly radioactive material. The current resurgence of interest in the high-temperature reactor in certain countries provides a need to demonstrate that the totality of the carbon materials present in their reflectors and in the fuel blocks or pebbles can be appropriately managed throughout the graphite life cycle. Many of the older reactors are now shut down, with more approaching the end of their lives, and in excess of 250,000 tonnes of radioactive graphite ('i-graphite') has now accumulated worldwide. Dismantling old reactors and the management of i-graphite are becoming an increasingly important issue for a number of countries. 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 during the period 2010-2014 and involved 24 organisations from 10 countries. The overall objective of the CRP was to advise experts of the various options currently being researched, to enable them to make an informed decision on the correct policy for their particular situation. The CRP explored innovative and conventional methods for graphite characterisation, retrieval, treatment and conditioning.
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.
Two important initiatives are continuing within the IAEA in the area of i-graphite. The first is a repository for data and reports within the IMMONET knowledge network, accessible via the IAEA ‘Nucleus’ portal. The second is a new project called GRAPA (Graphite Processing Approaches) as part of the IAEA International Predisposal and Decommissioning Networks, in which experts from seven countries are currently participating. This seeks to continue the i-graphite work and to support establishing a demonstration pilot plant for graphite removal and treatment.
Researchers from China, France, Germany, Lithuania, Russia, Spain, Switzerland, Ukraine, the UK and the USA participated in this CRP.
A TECDOC was published in 2016:
For more information, please see the CRP description:
http://www.dgdingfa.net/projects/crp/t21026
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