This paper addresses the stability of commercial cation exchangers when in contact with nitric acid that are under consideration for UCLan’s alternative reprocessing process. The paper describes the acid tolerance of four commercial ion exchange resins when exposed to 7M nitric acid for one year. The acid degradation was measured by FT-IR and TGA before and post acid exposure. Zirconium uptake was measured pre and post acid exposure as a means of determining degradation of functional sites. The resins undergo a degree of acid degradation depending on their chemical composition i.e., polymeric backbone, in some instances loss of functionality but also the introduction of new functional sites. The likely radiation dose that ion exchange resins would receive from spent nuclear fuel dissolver liquor and from adsorbed radioactive cesium 137 were estimated. It was inferred that the radiation dose the exchangers would receive from adsorbed cesium isotopes would be comparatively low when compared with the radiation background of the spent fuel dissolver liquor.

Due to their acid characteristics and pore structure, which can induce high product selectivity; zeolite catalysts are used extensively in industry to catalyse reactions involving hydrocarbons. However, these catalysts can suffer from deactivation due to cracking reactions that result in the deposition of carbon leading to poisoning of the acid sites and blocking of the pores [1]. Depending upon the reaction and the particular catalyst involved this deactivation may take place over several months or even years but in some cases occurs in minutes. Therefore, zeolite catalysts are frequently reactivated / regenerated. This generally involves a thermal treatment involving air which results in oxidation of the carbon [2]. However, the oxidation of carbon is highly exothermic, and if not carefully controlled, results in the generation of exceedingly high localized temperatures which can destroy the zeolite structure and result in subsequent loss of catalyst activity. More conservative thermal treatments can result in incomplete regeneration and again a catalyst displaying inferior activity. This paper explores the use of non-thermal plasma which had been either generated using microwaves or via a barrier discharge to regenerate spent zeolite catalysts. The catalyst, H-mordenite, was tested for the disproportionation of toluene (Figure 1) using conventional heating. The spent catalyst was then regenerated using a plasma or conventional thermal treatment before having its activity re-evaluated for the toluene disproportionation reaction as previous. Fig. 1. Reaction Scheme for Toluene Disproportionation. Interestingly, not only is plasma regeneration highly effective but also catalysts can be regenerated in greatly reduced times. There is an additional advantage in that plasma regeneration can impart physical properties that result in a zeolite that is resistant to further deactivation. However, the results are highly dependent upon the experimental conditions involved for plasma regeneration. References Wu J, Leu L., Appl. Catal., 1983; 7:283-294. M. Guisnet and P. Magnoux, Deactivation of Zeolites by Coking. Prevention of Deactivation and Regeneration. In: Zeolite Microporous Solids: Synthesis, Structure, and Reactivity. E.G. Derouane, F Lemos, C. Naccache, F. Ramôa Ribeiro, Eds. Pages 437-456. Springer 1992.

14C is an important consideration within safety assessments for proposed geological disposal facilities for radioactive wastes, since it is capable of re-entering the biosphere through the generation of 14C bearing gases. The irradiation of graphite moderators in the UK gas-cooled nuclear power stations has led to the generation of a significant volume of 14C-containing intermediate level wastes. Some of this 14C is present as a carbonaceous deposit on channel wall surfaces. Within this study, the potential of biofilm growth upon irradiated and 13C doped graphite at alkaline pH was investigated. Complex biofilms were established on both active and simulant samples. High throughput sequencing showed the biofilms to be dominated by Alcaligenes sp at pH 9.5 and Dietzia sp at pH 11.0. Surface characterisation revealed that the biofilms were limited to growth upon the graphite surface with no penetration of the deeper porosity. Biofilm formation resulted in the generation of a low porosity surface layer without the removal or modification of the surface deposits or the release of the associated 14C/13C. Our results indicated that biofilm formation upon irradiated graphite is likely to occur at the pH values studied, without any additional release of the associated 14C.

This concept paper, discusses the challenges and opportunities for an extractive chromatographic process for the separation of fission products and minor actinides from uranium and plutonium isotopes in irradiated nuclear fuel. The paper highlights the constraints of the PUREX process, a process that is universally accepted for reprocessing of spent nuclear fuel now and for GEN IV reactor systems. It also identifies the challenges that a new separation process would have to overcome to dislodge its acceptance by both the operators and regulators. Although the concept of using a chromatography technique for this separation is challenging, recent developments of continuous chromatography such as simulated moving bed (SMB) and/or continuous annular chromatography (CAC) provides a degree of encouragement. Equally the development of new stationary phases in particular inorganic exchangers, many of which have not been examined for this application enhances confidence that an alternative to the PUREX process is possible.

The present work has used microwave plasma chemical vapour deposition to generate suitable isotopically labelled carbonaceous deposits on the surface of Pile Grade A graphite for use as surrogates for studying the behaviour of the deposits observed on irradiated graphite extracted from UK Magnox reactors. These deposits have been shown elsewhere to contain an enhanced concentration of 14C compared to the bulk graphite. A combination of Raman spectroscopy, ion beam milling with scanning electron microscopy and secondary ion mass spectrometry were used to determine topography and internal morphology in the formed deposits. Direct comparison was made against deposits found on irradiated graphite samples trepanned from a Magnox reactor core and showed a good similarity in appearance. This work suggests that the microwave plasma chemical vapour deposition technique is of value in producing simulant carbon deposits, being of sufficiently representative morphology for use in non-radioactive surrogate studies of post-disposal behaviour of 14C-containing deposits on some irradiated Magnox reactor graphite.

A four year collaborative European Project ‘Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (Carbowaste)’ was launched in May 2008 under the 7th EURATOM Framework Programme. The aim of the project is to develop best practices in the retrieval, treatment and disposal of irradiated graphite, addressing both existing legacy waste as well as waste from graphite-based nuclear fuel resulting from a new generation of nuclear reactors. This paper covers the activities led by the National Nuclear Laboratory in partnership with the Nuclear Decommissioning Authority in the first year of the project, which includes the lead role in the compilation of a review volume on the extent of irradiated graphite waste globally and the approaches being taken to manage it. An overview is also provided of modelling activities in year two of the project: the application of modelling techniques to the prediction of radiological inventories, to the radiological impact of C-14 and Cl-36 releases on the biosphere and to the decommissioning of Magnox reactor cores.