Accepted Manuscript Development of integrated waste management options for irradiated graphite Alan Wareing, Liam Abrahamsen-Mills, PhD, Linda Fowler, Michael Grave, Richard Jarvis, PhD, Martin Metcalfe, PhD, Simon Norris, PhD, Anthony Banford, PhD PII: S1738-5733(16)30354-0 DOI: 10.1016/j.net.2017.03.001 Reference: NET 336 To appear in: Nuclear Engineering and Technology Received Date: 19 December 2016 Accepted Date: 21 March 2017 Please cite this article as: A. Wareing, L. Abrahamsen-Mills, L. Fowler, M. Grave, R. Jarvis, M. Metcalfe, S. Norris, A. Banford, Development of integrated waste management options for irradiated graphite, Nuclear Engineering and Technology (2017), doi: 10.1016/j.net.2017.03.001. This is a PDF file of an unedited manuscript that has been accepted for publication. As a service to our customers we are providing this early version of the manuscript. The manuscript will undergo copyediting, typesetting, and review of the resulting proof before it is published in its final form. Please note that during the production process errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal pertain. ACCEPTED MANUSCRIPT Irradiated graphite management route map The nature of irradiated graphite STEP 1 Problem definition, quantification and Graphite Surface • contamination : radioactivity is identification of constraints surface contamination present as deposits (e.g. Carbon-14 as soot deposits from reactor coolant gas) which may be loose or may be physically or chemically bound to the waste surface. STEP 2 • activation : radioactivity is present within the physical body of the waste i-graphite retrieval as activation products resulting from Activation Contamination penetration irradiation (e.g. Carbon-14 produced into open pore space from activation of Carbon-13 in the product graphite structure). STEP 3 MANUSCRIPTInterim storage and pre-treatment definition Potential STEP 4 STEP 5 Re-use/Recycle i-graphite treatment Re-use / Recycle products STEP 6 Graphite and secondary waste disposal ACCEPTED Paper Category – Original Article Development of integrated waste management options for irradiated graphite Authors Alan Wareing 1; Liam Abrahamsen-Mills, PhD 1; Linda Fowler 1; Michael Grave 2; Richard Jarvis, PhD 1; Martin Metcalfe, PhD 3; Simon Norris, PhD 4; Anthony Banford, PhD 1,5 ACCEPTED MANUSCRIPT 1 National Nuclear Laboratory, 5 th Floor Chadwick House, Warrington Road, Birchwood Park, Warrington, WA3 6AE, United Kingdom 2 Doosan Babcock, Construction House, Kingsway South, Team Valley Trading Estate, Gateshead, NE11 0ED, United Kingdom 3 National Nuclear Laboratory, 102B, Stonehouse Park, Sperry Way, Stonehouse, Gloucestershire, GL10 3UT, United Kingdom 4 Radioactive Waste Management Limited, Building 587, Curie Avenue, Harwell Science and Innovation Campus, Didcot, Oxon, OX11 0RH, United Kingdom 5 School of Chemical Engineering and Analytical Science, University of Manchester, Sackville Street, Manchester, M13 9PL, United Kingdom Running Title: i-Graphite Waste Management Abstract Word Count: 196 Text Word Count: 5094 (excludes References, Acknowledgement, Tables and Figures) Corresponding Author: Anthony William Banford, PhD 1 National Nuclear Laboratory, 5 th Floor Chadwick House, Warrington Road, Birchwood Park, Warrington, WA3 6AE, United Kingdom, Email [email protected] MANUSCRIPT and 5 School of Chemical Engineering and Analytical Science, University of Manchester, Sackville Street, Manchester, M13 9PL, United Kingdom, Email [email protected] Telephone: +44 (0)771 504 3778 First Author: Alan Wareing ACCEPTED 1 Abstract The European CARBOWASTE project sought to develop best practices in the retrieval, treatment and disposal of irradiated graphite (i-graphite) including other i-carbonaceous waste such as structural material made of graphite, non-graphitised carbon bricks and fuel coatings. Emphasis was given to legacy irradiated graphite, as this represents a significant inventory in respective national waste management programmes. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse and disposal of these graphite wastes. By considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programmes currently have, or are in the process of developing, respective approachesACCEPTED to i-graphite MANUSCRIPT management. The output of the CARBOWASTE project is intended to aid these considerations, rather than dictate them. Keywords: Integrated waste management; nuclear reactor decommissioning; irradiated graphite; radioactive waste management; decision framework MANUSCRIPT ACCEPTED 2 1 Introduction The use of graphite in nuclear reactors world-wide as moderator, reflector or operational material results in an accumulation of radioactivity by neutron activation both of the constituent elements of the graphite and of impurities, as well as the potential contamination of its surface. This irradiated graphite (i-graphite) presents a major waste management challenge due to the presence of long-lived radionuclide species such as 14 C and 36 Cl, together with shorter-lived species including 3H and 60 Co, and small quantities of fission products and actinides Over 250,000 tonnes of i-graphite have been accumulated worldwide [1], ranging from countries with a fleet of multiple graphite-moderated power reactors (e.g. UK, France), prototypes and production reactors, to those with a single experimental reactor. Irradiated and contaminated graphite from reactor moderators and reflectors or thermal columns represent theACCEPTED greatest volume MANUSCRIPT of these waste materials. Currently, the majority of this i-graphite is held either in-situ within reactors or in vault/silo storage. Furthermore, Smith and Bredell [2] identify the potential large volumes of irradiated graphite associated with the potential future use of pebble bed modular reactors (PBMRs). There are various options that could be adopted as waste management solutions for i-graphite and many of these have been investigated during the recent European Commission project ‘Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (CARBOWASTE)’ under the 7 th EURATOM Framework Programme [3]. The project was designed to develop best practices in the retrieval, treatment and disposal of i-graphite and to deliver an integrated waste management approach suitable for application by different countries and sites, each with their own particular conditions to meet (e.g. a specific disposal end point or regulatory requirements). However, the purpose was not to dictate a national waste management strategy in relation to i-graphite. The CARBOWASTE project brought together organisations and stakeholders from the nuclear industry and scientific research establishments from European countries, as well as other international partners, to share knowledge and develop methodologies for i-graphite management [4]. 1.1 The i-graphite lifecycle Figure 1 provides a schematic diagram of an example i-graphite lifecycle, showing the principal stages from graphite manufacture to final disposal. During manufacture, impurities becomeMANUSCRIPT associated with the graphite matrix which, along with naturally occurring carbon isotopes, lead to the generation of a range of radionuclides within the graphite matrix during reactor operation (see Section 2.1.1). Following the shut-down of a reactor, the radioactive inventory within the i-graphite will decrease with time, due to radioactive decay. Following a period of in-reactor storage, the point at which the i- graphite is retrieved will influence the radiological hazard posed by the material, since some of the radionuclides present are relatively short-lived. Whilst the radioactivity associated with i-graphite cannot be destroyed, methods of treatment or conditioning can be used to convert it into alternative, more manageable physical and chemical forms. A decontamination process may reduce the radioactivity associated with the bulk graphite matrix but, in doing so, will generate a secondary waste stream which must also be appropriately managed. It may be possible to recycle or re-use i- graphite materials, e.g. the use of decontaminated graphiteACCEPTED in other industrial processes. However, due to the long half-lives of a number of the radionuclides present, e.g. 36 Cl (308,000 years), it is likely 3 that some material will require a disposal method that isolates it from the environment for an extremely long period, e.g. within a geological or engineered (e.g. surface) disposal facility. Figure 1 Schematic of i-graphite lifecycle The identification of potential options for the management of i-graphite that address each stage of the i-graphite lifecycle needs to account for the specific physical, chemical and radiological characteristics and behaviour of the material. These factors will influence the feasibility and performance of processes