The practices applied in safeguarding Special Nuclear Materials (SNM) for civil use are outlined together with the basic techniques aiming at an efficient and effective control by the entitled inspectorates The role of the Joint Research Centre (JRC) Ispra is described to support these inspectorates in developing and implementing cost-effective, reliable and user-friendly instrumentation and in improving the performance of the inspections and verifications in field by making available appropriate training facilities and courses.

Management aspects of fusion waste are presented. Conditioning and packaging of tritiated waste must cope with the high mobility of tritium. The first required treatment is immobilization of the waste in solid matrices. Additional barriers against tritium migration are provided by encapsulating layers. Packaging is afforded by single or double-walled structures. Low activation materials (LAMs) are being developed to simplify the management of fusion activated waste. Interim storage of this waste is envisaged at the plant site, then various options are possible, depending on the residual radioactivity levels, i.e. declassification to non-active-waste, recycling, disposal.

Gas chromatography with a modified mordenite column was shown to yield a good hydrogen isotope separation at only 173 K. A linear relationship between peak height and concentration over several orders of magnitude was observed. From an extrapolation of the data it is concluded that all six hydrogen isotopes can be separated in only 13 min. Gas chromatographically separated radioactive hydrogens were detected with a small volume ionization chamber. The detection limit achieved for tritium was found to be about 0.3 ppm.

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.

The present paper compares, in terms of safety and reliability, two different designs for tritium recovery from DEMO Pb-17Li water-cooled blanket.The first one, based on tritium recovery from Pb-17Li outside the blanket units, tritium removal from the purge gas of the extracting system and tritium permeation barriers towards the coolant, is overviewed and analyzed especially with regard compactness and reliability.In the alternative option, tritium generated in the blanket is free to permeate into the cooling water and is recovered by the water detritiation system, avoiding, therefore, the external tritium extractor from Pb-17Li and permeation barriers, the technological feasibility of which is not yet proven on industrial plant scale.A safety analysis of this alternative design has been carried out both for normal operation and accident conditions. The results, here presented and discussed, have shown that this alternative option seems to be feasible and worthy of successive research and investigations.

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 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.

The European Tritium Handling Experimental Laboratory (ETHEL) is a new tritium facility at the Commission of the European Community's Joint Research Centre, Ispra Site. The laboratory, destined to handle multigramme amounts of tritium for safety related R&D purposes, is foreseen to start radioactive operations in late 1992. The general operation and maintenance of laboratory systems and future experiments will generate tritiated wastes in gaseous, liquid and solid forms. The management of such wastes under safe working conditions is a stringent laboratory requirement aimed at minimising the risk of unacceptable tritium exposures to workers and the general public. This paper describes the main systems and facilities installed in ETHEL for the safe management of tritiated wastes.

The European Tritium Handling Experimental Laboratory is committed to investigate the propagation modes of tritium in materials, components, equipment and process loops. As for other radiochemical facilities, a number of safety criteria and principles have been introduced into the design, construction and intended operation of the facility with the aim of optimising the protection of both the workers and the population. This paper highlights a variety of safety aspects applied to the laboratory: confinement barriers, radiological protection, waste handling and, finally, precautions against internal and external events.

For the implementation of safeguards in modern and automated nuclear facilities, a large variety of sophisticated verification techniques are being used. The safeguards Authorities are confronted with the need to test the performance of these techniques, to periodically calibrate instruments and test measurement and inspection procedures and last but not least provide effective training to their inspectors. In this framework, JRC has started several years ago a program for establishing test and training facilities. They are PERLA for non destructive assay, LASCO for containment and surveillance, Tank Measurement Lab (TAME) for volume and weight measurement and PETRA for studies of safeguards for reprocessing plant and active waste. PERLA and LASCO are operational and already used by safeguards inspectors, TAME and PETRA will become operational in 1992. The paper describes the characteristics and functions of the different facilities, their actual use and future plans of development.