This title appears in the Scientific Report : 2013 

Elicitation of dissolution rate data for potential wasteform types for plutonium under repository conditions
Deissmann, Guido (Corresponding author)
Neumeier, Stefan / Brandt, Felix / Modolo, Giuseppe / Bosbach, Dirk
Nukleare Entsorgung und Reaktorsicherheit; IEK-6
Migration 2013, Brighton (UK), 2013-09-08 - 2013-09-13
Safety Research for Nuclear Waste Disposal
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Various plutonium isotopes are generated during the operation of nuclear reactors from the uranium present in the nuclear fuels through capture of neutrons. The plutonium contained within spent nuclear fuels can be recovered during reprocessing. At present, separated stocks of UK civil plutonium (about 90 tHM as PuO2) are held in storage as zero value asset, since there is no final decision on plutonium disposition in the UK. Recently, the UK government stated that its preliminary preferred policy on the long-term management of plutonium is reuse as MOX fuel, but consideration of disposal options will continue [1]. However, irrespective of future UK government strategies for plutonium disposition, at least a portion of the UK plutonium inventory (i.e. some tonnes) is likely to be designated for geological disposal.This paper will describe outcome and conclusions of an elicitation exercise regarding the dissolution rates of, and the plutonium release from, potential wasteforms for plutonium, performed on behalf of the NDA RWMD. The elicited distributions provide a 'non-conservative lower bound', a 'conservative upper bound' and where possible a 'best current estimate' for the dissolution rates and the plutonium release rates. These data can be used in Post-Closure Safety Assessment Models to calculate performance measures such as annual individual risk related to plutonium disposal with time. Generic candidate wasteform types included in this study were borosilicate glasses and phosphate glasses, ceramic wasteforms, and low-specification "storage" MOX. Due to the character of the current UK disposal programme (‘generic stage’, in which no preferred disposal concept has yet been selected), a range of possible environmental conditions in the repository near-field, including effects of an alkaline plume potentially arising from a co-located cementitious LILW-repository module, and various disposal scenarios were considered.Experimental data on the durability of plutonium wasteforms under repository-relevant conditions are generally limited or in many cases absent (e.g. [2, 3]). Due to this limited knowledge base, the initial part of the work comprised the development of conceptual approaches to evaluate the plutonium release rates from various wasteforms. Following the compilation and analysis of relevant data and information, the respective distributions of the wasteform dissolution rates and plutonium release rates for the conditions expected in a geological repository in the UK were derived. The information basis regarding wasteform durability and leaching resistance is rather diverse for the different matrices. Information on the performance of plutonium-bearing glasses is rather limited to date. However, considerable knowledge exists from laboratory and in-situ studies about borosilicate-based nuclear waste glasses for disposal of reprocessing wastes and their long-term performance. In contrast, the database regarding the long-term leaching behaviour of phosphate-based glasses was found to be comparatively small. Data with respect to the performance of ceramic wasteforms in the repository environment are rather scarce, and a systematic approach to an understanding of the aqueous durability of the various ceramic matrices (regarding, e.g., the dissolution behaviour as function of the crystallographic structure, chemical composition, lattice substitutions, radiation damage, etc.) is still lacking. Furthermore, even for nominally similar matrices, the data can show a considerable spread and are often difficult to compare, due to different processing and fabrication routes employed, different experimental conditions, as well as the usage of plutonium surrogates in some experiments. Experimental investigations on the dissolution and long-term performance of storage MOX and/or calcined PuO2 are also rather limited to date. Thus the assessments and the elicitation of the dissolution rates of this wasteform was based on experiments and modelling studies related to the matrix dissolution of spent nuclear fuels (i.e. UOX and MOX), and on the understanding of relevant processes affecting their long-term behaviour in a repository, which has been significantly expanded throughout the last decades. The dissolution rates derived for the different (generic) plutonium wasteforms under conditions relevant for a UK geological disposal facility are summarised in table 1. Table 1. Elicited dissolution rates (in g m-2 d-1) for potential plutonium wasteforms under conditions relevant for a UK geological disposal facilityWasteform Lower bound Best estimate Upper boundBorosilicate glass <10-4 10-4 … 10-2 100Phosphate glass <10-5 10-5 … 10-2 10Ceramic wasteforms 10-7 10-5 … 10-4 0.5Storage MOX 10-7 5•10-6 10-2 [1] Department of Energy & Climate Change, Management of the UK’s Plutonium Stocks, London (2011).[2] E.M. Pierce et al., Appl. Geochem. 22: 1841–1859 (2007).[3] G. Deissmann et al., Min. Mag. 76: 2911–2918 (2012).