[en] This report covers the tenth year of activity in connection with the Eurex project since the signing of Euratom-CNEN Agreement no.001-64-11 RC-II for the construction, operation use for industrial research purposes of the Eurex plant. The report summarizes the contents of three four-monthly reports published during 1974 and presents a summary statement of expenditure. The report contains the following parts: management of the reprocessing division; planning and construction of the plant (modification); cold tests; laboratory and pilot-scale experiments prior to start-up of the plant; industrial operation of the plant

[en] JRC-ETHEL has chosen as the principle objective of its research program the improvement of protection measures in facilities handling large amounts of tritium. Technically, this involves investigating and assessing tritium propagation modes and transfer pathways in materials, components, equipment, and process plants. The experiment research work to be performed in ETHEL will basically aim at investigating: (1) Loss mechanisms by identifying physico-chemical parameters such as adsorption/desorption rates, permeation rates, leakages of materials for fusion reactors and the effects of potential remedies like permeation barriers under process-like conditions. (2) Multiple containment systems and fluid clean-up concepts under normal and accidental conditions. (3) Methods for solid waste handling, treatment, conditioning, and final disposal. (4) Techniques for tritium control, monitoring, and surveillance over the whole concentration range during both normal and accidental conditions and maintenance activities. With the availability of two open-quotes climate chambers,close quotes the small and large caissons of 5 and 350 m³ volume respectively, ETHEL is especially suited for benchmark and scale-up tests of many kinds of large gas volumes treatment system. This will help to close the gap between laboratory-scale results and plant-size design specifications and represents an important source of information for designers (NET, ITER) and regulatory authorities

[en] The R and D activities, carried out in Ispra since 1984, in the field of radioactive waste management,COMB (Technological Development Laboratory) and JRC, concerned final hot tests of the ENEA owned ESTER mini-pilot plant for HLW vitrification, the trasportation of3 of the activeglass? containing crucibles from the Ispra to the Karlsruhe (Transuranium Institute) establishment of JRC for produced by the ESTER plant. While the ESTER experience is being successfully finished, contemporarily the basis has been established to start the PETRA experience, more complete than the previous one, but also more complex. The PETRA experimental infrastructure will be particulary useful for studying, developing and verifying, in real activity conditions, advanced chemical extracting agents and mainly new matrices for the immobilisation of HLWs, or their fractions, and of mixtures of various aqueous waste streams

[en] Studies on the public perception of risks indicate that: (1) The most feared risks, together with open-quotes AIDSclose quotes and open-quotes crimeclose quotes, are those related to open-quotes nuclear accidentsclose quotes and open-quotes nuclear wasteclose quotes, which are almost as dreadful as open-quotes nuclear warclose quotes whereas, open-quotes nuclear power productionclose quotes raises less concern, almost comparable to that of other social risks. (2) Low probability, high consequence risks are those of most concern. Scarce consideration is given by the public to the likelihood of occurrence. The management strategies for fusion waste should be presented taking into account the previous statements. Sensible items are: (1) Planned waste management procedures should be describe in a well detailed fashion, indicating the aim of complying with conservative limits of environmental impact. (2) The build-up of optimistic thoughts on the complete lack of potential hazard in fusion waste as well as the opposite opinion of little or no progress at all by referring to fission waste should be avoided. (3) The hazards related to fission waste should not be over emphasized. It should be stated instead that studies on this waste have produced effective handling and segregation concepts, which are being progressively applied even to conventional waste in order to reduce risks which were not perceived before. This mature and well proven technology will be applied where pertinent to fusion waste too

[en] The number and variety of measurement and monitoring techniques used by inspectors for the implementation of nuclear safeguards has increased much over the last 10 yr. These techniques are also continuously being adapted and improved to profit from new technological developments. Real-scale experiments to test new measurement techniques and systems, their performance assessment, the development of measurement procedures, calibration of instruments, and training of inspectors are often performed in industrial facilities. With the large increase in scale and the change in layout of future facilities, experiments and tests will become nearly impossible and make evident the need to have available specially equipped laboratories and facilities that simulate, to the extent possible, the measurement conditions encountered in reality. 5 refs

[en] On the basis of the laboratory indications presently available, all three proposed partitioning processes appear to be feasible and to have the potential for removing actinides from HAW (TBP, HDEHP, OXAL) up to the necessary level. Additional data on the actinide losses to the various secondary waste streams, on the radiation stability of the chemicals used for stripping operations, as well as on their recycle or removal as wastes still need to be obtained on laboratory scale for fully active HAW. The obtained results will provide the basis for deciding whether to proceed or not to a pilot plant scale experimentation. Until such experiments are performed it will be impossible to demonstrate the overall feasibility of the selected partitioning process with an acceptable degree of reliability. 3 figures, 4 tables

[en] The presence of tritium in tritium-burning devices to be built for large scale research on thermonuclear fusion poses many problems especially in terms of occupational and environmental safety. One of these problems derives from the production of tritiated wastes in gaseous, liquid and solid forms. All these wastes need to be adequately processed and conditioned to minimize tritium releases to an acceptably low occupational and environmental level and consequently to protect workers and the public against the risks of unacceptable doses from exposure to tritium. Since all experimental thermonuclear fusion devices of the Tokomak type to be built and operated in the near future as well as all experimental activities undertaken in tritium laboratories like ETHEL will generate tritiated wastes, current strategies and practices to be applied for the routine management of these wastes need to be defined. Adequate background information is provided through an exhaustive literature survey. In this frame alternative tritiated waste management options so far investigated or currently applied to this end in Europe, USA and Canada have been assessed. The relevance of tritium in waste containing gamma-emitters, originated by the neutron activation of structural materials is assessed in relation to potential final disposal options. Particular importance has been attached to the tritium retention efficiency achievable by the various waste immobilization options. 19 refs., 2 figs., 1 tab

[en] In the frame of JRC's 1984-87 Fusion Technology Programme and with the aim of contributing to tritium R and D for NET, the commission of the European communities decided in July 1985 the construction on the site of the Ispra establishment of a European tritium handling experimental laboratory (ETHEL). ETHEL is a facility designed to handle a maximum inventory of 100 G tritium and to release annually a maximum of some tenths of curies of tritium. Its commissioning is foreseen for end 1989 beginning 1990. Current practices to be applied for routine management of tritiated wasted generated during the normal ETHEL operations as well as objectives of the experimental research that the JRC staff specifically intend to perform in the field of tritiated waste management strategy and technology are briefly illustrated

[en] A flow sheet is presented for the isolation of pure hydrogen from the gas mixture generated by the well known water-gas shift reaction H₂+CO=H₂+CO₂ Separation of the resulting gas mixture is performed on three modified selective zeolite beds. By applying a highly integrated flow pattern with extended recycling of regeneration process streams to the reactor, the only gaseous waste stream generated is CO₂. The preparation of these zeolites as well as other materials is described

[en] Three conceptual processes were studied and investigated for the feasibility of removing actinides from high-activity waste (HAW). Two of the flow-sheets rely completely on counter-current techniques for the actinides separation, namely TBP and HDEHP, whereas the third process, OXAL, uses a precipitation technique in the first instance followed by dissolution of the actinides and rare-earths (RE). From the assessment studies there has emerged the need for R and D work in a number of areas, in particular fast contactor extraction techniques, process safety, control and criticality of the minor actinides, before the processes could be considered feasible and operative in such an advanced fuel-cycle technological industry. On the basis of these results an alternative long-term management scheme is also proposed. The conceptual scheme proposes splitting the actinides and the RE from the fission products (FP) using selected steps of the above processes. The resulting chemical homogeneous fraction of actinides and RE has a considerably lower associated heat output than the original HAW, together with the absence of volatile components. For this group, 'tailor-made' ceramic matrices can be applied for solidification with the aim of obtaining better immobilization since any detrimental effects of the high initial heat generation on the matrix, as well as on the geological formation selected for disposal, are excluded. Detailed risk analysis and cost-benefit investigations are required to place this conceptual scheme on a firm basis. (author)