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Processing of Irradiated Graphite: The Outcomes of an IAEA Coordinated Research Project

Published online by Cambridge University Press:  16 February 2017

Michael I. Ojovan  and
Anthony J. Wickham
Michael I. Ojovan*
Affiliation:
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-1400Austria
Anthony J. Wickham
Affiliation:
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, [email protected]
*
*(Email: [email protected])
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Abstract

Dismantling of old reactors and the management of radioactive graphite wastes are becoming increasingly important issues for a number of IAEA Member States. Exchange of information and research cooperation in resolving identical problems between different institutions contributes towards improving waste-management practices, their efficiency, and general safety. The IAEA Coordinated Research Project (CRP) under the title ’Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal’ was conducted during 2010-2014 and has involved 24 organisations from ten Member States [1]. The CRP has explored both innovative and conventional methods for graphite characterisation, retrieval, treatment, and conditioning technologies and produced an IAEA technical document [2] which has identified a number of unresolved scientific and technical issues such as the need to:

  • 1. Improve the scientific understanding required on creation, chemical form, location and release behaviour (transport models) of radionuclides;

  • 2. Improve predictive models of radioisotope behaviour;

  • 3. Ensure that sampling programmes are statistically representative of the totality of the graphite to be disposed of;

  • 4. Establish an accurate radionuclide inventory;

  • 5. Consider novel alternative dismantling and treatment strategies.

The CRP promoted the exchange of technical information on R & D activities and will facilitate practical application for treatment and conditioning of graphite waste. The collaboration continues under the IAEA International Decommissioning and Predisposal Networks (IDN and IPN).

Keywords

radiation effectnuclear materialswaste management
Type
Articles
Information
MRS Advances , Volume 1 , Issue 62: Scientific Basis for Nuclear Waste Management XXXIX , 2016 , pp. 4117 - 4122
Copyright
Copyright © Materials Research Society 2017 

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References

REFERENCES

Ojovan, M.I., Wickham, A.J., ’Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal: Problem and Solutions’, Mat. Res. Soc. Symp. Proc., 1665, (2014), 3 – 12 Google Scholar
International Atomic Energy Agency, ’Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project’, IAEA-TECDOC-in press, (2016).Google Scholar
International Atomic Energy Agency, ’Graphite Moderator Lifecycle Behaviour’, Proceedings of a Specialists Meeting held in Bath, United Kingdom, 24–27 Sept 1995; IAEA-TECDOC-901, Vienna (1996).Google Scholar
International Atomic Energy Agency, ’Nuclear Graphite Waste Management: Technical Committee Meeting on Nuclear Graphite Waste Management’, Manchester UK, 18–20 October 1999, IAEA CD-ROM 01-00120, Vienna (2001).Google Scholar
International Atomic Energy Agency, ’Progress in Radioactive Graphite Waste Management’, IAEA-TECDOC-1647, IAEA, Vienna (2010).Google Scholar
International Atomic Energy Agency, ’Characterization, Treatment and Conditioning of Radioactive Graphite from Decommissioning of Nuclear Reactors’, IAEA-TECDOC-1521, IAEA, Vienna (2006).Google Scholar
White, I.F., Smith, G.M., Saunders, L.J., Kaye, C.J., Martin, T.J., Clarke, G.H. and Wakerley, M.W.; ’Assessment of Management Modes for Graphite from Reactor Decommissioning’, Commission of the European Communities, EUR 9232, Brussels, (1984).Google Scholar
Banford, A., Eccles, H., Graves, M., von Lensa, W., Norris, S.. ’CARBOWASTE – An Integrated Approach to Irradiated Graphite’, Nuclear Future, 4, 268–270 (2008).Google Scholar
International Atomic Energy Agency, ’Nuclear Technology Review’, IAEA, Vienna, (2013)Google Scholar
Nair, S., ‘A Model for Global Dispersion of 14C Released to the Atmosphere as CO2’; J. Soc. Radiological Protection, 3, 16–22 (1983).Google Scholar
Rublevskiy, V.P., ’Commercial Nuclear Reactors and 14C’, Atomic Energy, 113, 143–147 (2012)Google Scholar
Poncet, B., Petit, L., ’Method to Assess the Radionuclide Inventory of Irradiated Graphite Waste from Gas-Cooled Reactors’, J. Radioanal. Nucl. Chem., 296, 3 (2013).Google Scholar
Heasler, P.G., Jensen, L., ’Statistical Evaluations of Current Sampling Procedures and Incomplete Core Recovery’; Pacific Northwest Laboratories Report PNL-9408 (1994)Google Scholar
Dunzik-Gougar, M.L. and Smith, T.E., ’Removal of Carbon-14 from Irradiated Graphite’, J. Nucl. Mater., 451, 328–335 (2014).Google Scholar
Dunzik-Gougar, M.L., Cleaver, J., LaBrier, D., Nelson, K. and Smith, T., ’Chemical Characterisation and Removal of Carbon-14 from Irradiated Graphite - III’; Proc. WM Symp. 2014, Phoenix (February 2014).Google Scholar
McGann, O.J., Ojovan, M.I.. The Synthesis of Graphite-Glass Composites Intended for the Immobilisation of Waste Irradiated Graphite. J. Nucl. Mater., 413, 47–52 (2011).Google Scholar
Mayzan, M.Z.H., Stennett, M.C., Hyatt, N.C., Hand, R.J.. Graphite immobilisation in iron phosphate glass composite materials produced by microwave and conventional sintering routes. J. Nucl. Mater., 454 (1–3), 343–351 (2014).Google Scholar
Heath, P.G., Corkhill, C.L., Stennett, M.C., Hand, R.J., Meyer, W.C.H.M., Hyatt, N.C.. Encapsulation of TRISO particle fuel in durable soda-lime-silicate glasses. J. Nucl. Mater., 436 (1–3), 139–149 (2013).Google Scholar
International Atomic Energy Agency, http://nucleus.iaea.org/sites/nefw-projects/IMMONET/graphite-crp/SitePages/Home.aspx (accessed 11.02.2016).Google Scholar