The most important lessons learned in recent years from decommissioning are organized and examined to enable the importance of this aspect of planning for new nuclear facilities to be gauged. Policy, strategy, licensing and regulatory aspects are examined. The design of physical features and methods of operation are discussed in order to make recommendations to facilitate decommissioning. The issues are presented in a concise and systematic manner, along with practical and thought-provoking examples.

Since 2001 the International Atomic Energy Agency has championed the concept and use of Networks to advance radioactive waste management across the globe. At the present time there are four Networks managed on behalf of Member States by the IAEA and a fifth one is currently being implemented. The scopes of interest covered by the Networks encompass near-surface and deep geological disposal, the decommissioning of nuclear facilities, the environmental remediation of sites contaminated with radioactive materials and the characterisation of low- and intermediate-level radioactive wastes. To date over 100 organisations from more than 40 Member States are involved in the Networks. Many of these Network participants generously donate resources, time and effort to support Network activities, while others with nascent or otherwise less well developed programmes are still in the process of acquiring experience, capabilities and know-how. Regardless of the stage of development, all Network participants share in the mutual benefits that arise from improved communications with sister organisations and the sharing of experience and knowledge.The universal Goal of the Networks is the promotion of methods and technologies that will enhance the safety and sustainability of radioactive waste management practices and facilities. This Goal is being achieved through continuous improvements in communication and knowledge sharing between Network participants and the provision of enhanced opportunities for training, involvement in demonstration projects and the development of novel technologies and methodologies. We recognise that interdisciplinary understanding and the coordination of efforts at key interfaces at the back-end of the fuel cycle are critical aspects for achieving the Network Goal efficiently and effectively. Consequently, the IAEA Networks that will be operational by the end of 2010 are themselves are being molded into an organic "Network of Networks" where the use of new electronic media and the possibilities presented by enhanced communication channels will be exploited.Here we provide an overview of the IAEA Networks in radioactive waste management and present a new tool that is under development, an internet-based portal for enhanced communications and the provision of improved training opportunities.

Au cours des dix dernières années, l’AIEA a poursuivi et amplifié ses efforts visant à mieux répondre aux besoins des Etats-membres en matière de démantèlement d’installations. Ces besoins croissants sont le résultat d’un grand nombre d’installations nucléaires déjà arrivées ou qui parviendront bientôt au terme de leur exploitation, et se traduisent par des demandes d’appui technique toujours plus nombreuses reçues par l’AIEA.Les objectifs du programme de l’AIEA relatif au démantèlement des installations sont les suivants :• Sensibiliser les Etats-membres à leurs responsabilités vis-à-vis du démantèlement des installations et de la réhabilitation de sites contaminés par des résidus radioactifs et les conseiller en la matière ;• Faciliter et promouvoir l’échange d’informations et de retour d’expérience et le transfert de savoir-faire sur les stratégies, méthodes, état de l’art et technologies ;• Aider les organismes en charge du démantèlement dans les Etats-membres en voie de développement à améliorer leurs moyens et compétences par le biais du programme de coopération technique.Le rôle de l’AIEA dans la promotion de la coopération internationale entre ses États-membres est mentionné à l’article VIII.C de son Statut, qui prévoit que l’AIEA " prend des mesures positives pour encourager l’échange, entre ses membres, de renseignements sur la nature et l’utilisation de l’énergie atomique à des fins pacifiques, et, à cet effet, sert d’intermédiaire entre ses membres ". Ce rôle a été confirmé par les Etats-membres lors de Conférence Générale de 2007, comme indiqué dans les paragraphes 46 et 47 de la résolution GC(51)/RES/11A (2007), où les Etatsmembres se félicitent “des résultats de la Conférence internationale d’Athènes sur le démantèlement des installations nucléaires” et “du lancement du réseau international de démantèlement (IDN) qui servira de plateforme pour l’échange d’informations et les formations techniques par le biais de démonstrations pratiques sur le terrain”.Les activités en cours à l’AIEA dans le domaine du démantèlement d’installations sont très variées et comprennent notamment :• La mise en oeuvre du Plan d’Action International en matière de démantèlement approuvé par le Conseil des Gouverneurs de l’Agence et la Conférence Générale ;• L’élaboration de normes internationales de sûreté et la publication des documents techniques associés ;• La mise en place et l’animation de réseaux tels que l’IDN, proposant des activités de formation au niveau régional ou international et des exercices d’inter-comparaison ;• La mise à disposition des Etats-membres d’experts et la fourniture d’équipements ;• Des évaluations de sûreté, expertises techniques, revues de pairs de projets ou d’activités de démantèlement ;• La coopération et l’échange d’information avec des organismes internationaux et l’organisation de conférences et de séminaires.La mise en oeuvre des programmes de l’AIEA en matière de démantèlement impose une coordination étroite entre les Départements d’Energie Nucléaire, de Sûreté et de Sécurité Nucléaires et de Coopération Technique. Les Départements de Sûreté et de Sécurité Nucléaires et d’Energie Nucléaire apportent l’appui technique nécessaire à la bonne exécution des diverses activités, tandis que celui de la Coopération Technique se charge du pilotage des projets de coopération technique dont bénéficient une centaine d’Etats-membres sur les cent quarante-cinq que compte l’AIEA.Plusieurs exemples d’activités en cours et d’initiatives récentes illustrent la diversité et complexité du programme de l’AIEA en matière de démantèlement, ainsi le démantèlement complet des installations nucléaires et la réhabilitation des sites contaminés en Iraq, le projet-pilote de démantèlement de réacteurs de recherche, la mise en place du Réseau International de Démantèlement (IDN) et le projet international d’inter-comparaison DeSa visant a harmoniser les démonstrations de sûreté pour les projets de démantèlement. Les standards de sûreté et documents techniques publiés récemment sont également présentés.

Within the framework of the Safety and environmental assessment of fusion power (SEAFP) project, design features of tokamak power reactors and ancillary systems likely to ensure public acceptance and take into account the operating requirements of utilities have been analyzed. The present paper describes the development of fuel cycle systems with tritium inventories minimized to diminish the tritium release potential under accident conditions. Priority also has been given to designs with inherently low rates of waste production. This favours design options involving flow-through processing techniques (such as mechanical pumps, permeators, catalytic reactors) rather than techniques involving the accumulation of the processed gases (such as cryopumps, molecular sieves, getter-based techniques). In addition, processes which are undergoing validation within technology programmes, or which are already implemented on an industrial scale and where upgrading to achieve tritium compatibility is considered feasible, have been employed to produce a robust, reliable fuel cycle loop design.

This novel pumping concept consists of a 20 K stage to increase gas density followed by a mechanical pump operating at the same temperature. Advantages of the concept include order-of-magnitude reductions in size, weight, stored energy, eddy-current rotor heating, and sudden venting thrust when compared to a conventional turbomolecular pump scaled-up to perform the same duty. The device offers extremely low tritium inventory and near steady-state operation. Critical design and development issues include minimizing backstreaming in the mechanical portion of the pump, and eddy-current heating of the rotor.

This paper summarizes Fuel Cycle options with emphasis on recent innovations to reduce tritium irwentory. Tritium system design concepts described in the NET Pre-design Report [Toschi et al, 1993], and ITER-CDA Fuel Cycle Design Report [Leger et al., 1991] are supplemented by designs which address the high-throughput requirements currently under discussion for ITER (EDA) and reactors (SEAFP). The report provides a basis for selecting reference Fuel Cycle concepts for the EDA-ITER and SEAFP Reactor as details of these designs evolve.

As part of the development of a generic power reactor plant model to be used as the basis for the study of the Safety and Environmental effects of Fusion Power, a power reactor fuel cycle involving enhancements to the NET fuel cycle design has been defined. To facilitate the application of NET type dose targets in a reactor context, a fuel cycle able to cope with up to 4x higher fusion power without increasing either local or global inventories of hydrogen isotopes is required. To achieve such low vulnerable inventories, attention is given to advanced design concepts for the main potential sources of vulnerable inventory in the NET design. Implications for overall fuel cycle inventories and reactor design choices are summarized.

The NET Predesign Phase and the ITER Conceptual Design Activity resulted in documented conceptual designs. Assessment of these reports led to the quantification of a reasonable range of values for the tritium inventories of the following systems or plant areas: fuelling, plasma vacuum pumping, fuel purification, blanket tritium recovery, fuel cycle auxiliaries (including isotopic separation system), first wall and divertor, dust produced inside the plasma chamber, blanket, and coolant water of first wall, divertor and blanket. For realistic dose calculations source terms are derived from inventory estimates. Release limitation by process design features and confinement strategy following specific accident scenarios have been considered. The potential to reduce source terms by design improvements and alternative confinement strategies are evaluated.

This paper presents a summary of the results of a detailed design study performed for the Fuel Processing Loop of the Next European Torus (NET). The design is based on the concept of adsorption of impurities on molecular sieve at liquid nitrogen temperature, followed by catalytic oxidation of the regenerated impurities to water, and subsequent reduction of the water by electrolysis. The design study has shown that the process can be engineered using, mostly, available components, and can be designed to be safe. Special design features to make the process passively safe are described. The results of a preliminary safety analysis are summarized. Tritium inventories in the various sections of the process loops are shown. A cost estimate for the overall process system is also presented.

The ITER Conceptual Design Activity (CDA) was a three-year, 400 professional-year effort to design a next step tokamak. The activity was conducted under the auspices of the IAEA jointly by EURATOM, Japan, the USSR and USA. The main ITER parameters are summarized in the paper. An engineering design phase (EDA) lasting 5-6 years is planned to begin in 1992. Fuel Cycle design studies carried out as part of the CDA concluded that suitable options existed or could be developed to satisfy all tritium-handling requirements for the machine within the EDA time and resource framework. During the EDA, special emphasis will be required on design integration and optimisation.

Inter-connected elements are required to carry out fuelling of the torus, torus vacuum pumping, processing of (exhaust) fuel recovery of tritium from the breeder blanket and test sectors, and several "common" processes such as hydrogen isotope separation, storage and management of fuel gases, and treatment of solid, liquid and gaseous tritiated wastes. Results of the International Tokamak Experimental Reactor - Conceptual Design Activity (ITER-CDA) design efforts are summarized, and technical and organizational proposals made to facilitate design optimization during the next phase of the Project.

This paper presents integrated hydrogen Isotope Separation System (ISS) designs for ITER based on requirements for plasma exhaust processing, neutral beam injection deuterium cleanup, pellet injector propellant detritiation, waste water detritiation, and breeding blanket detritiation. Specific ISS designs are developed for a machine with an aqueous lithium salt blanket (ALSB) and a machine with a solid ceramic breeding blanket (SBB). The differences in the ISS designs arising from the different blanket concepts are highlighted. It is found that the ISS designs for the two blanket concepts considered are very similar with the only major difference being the requirement for an additional large water distillation column for ALSB water detritiation.

Plans for next-step fusion engineering devices involve the design of vacuum systems with many unique features. The vacuum vessel in which the fusion reaction takes place must be designed to demanding requirements with respect to size, leak-tightness, thermal and mechanical stresses, nuclear shielding, remote maintenance access and electrical conductivity. The vessel contains complex components designed to receive the particle and energy loads from the plasma, extract heat (in excess of 1000 MW), breed tritium fuel, and shield the superconducting magnets of the machine. These components are contained in a volume of the order of 2000 m3, and involve surface areas of >104m2. The torus evacuation system must provide, in addition to pump-down and conditioning of this vessel, a continuous evacution of He and impurities to permit the thermonuclear burn to be sustained. A total installed pump speed of∼ 1500 m3 s−1 requires innovative solutions.

Joint work on the design of the International Thermonuclear Experimental Reactor (ITER) has provided an opportunity for engineers and physicists from Japan, the EC, USA and Soviet Union to compare concepts and identify critical data needs which must be evolved over the next few years to support the design of the next-step machine for magnetic-confinement fusion. In the fuel cycle area, well-developed national programs exist, and the challenging task of the Fuel Cycle Design Unit in ITER is to promote a harmonisation of these tasks to minimise duplication and ensure gaps which would affect the development of satisfactory design concepts do not occur. The principle mechanisms for this involves: •- identification of a subset of R&D task where results are especially needed for design,•- establishment of task and subtask definitions and program milestones to facilitate task monitoring,•- reporting and information exchange as results become available.

The increasing use of mammography to screen asymptomatic women makes it important to know the risk of breast cancer associated with exposure to low levels of ionizing radiation. We examined the mortality from breast cancer in a cohort of 31,710 women who had been treated for tuberculosis at Canadian sanatoriums between 1930 and 1952. A substantial proportion (26.4 percent) had received radiation doses to the breast of 10 cGy or more from repeated fluoroscopic examinations during therapeutic pneumothoraxes. Women exposed to greater than or equal to 10 cGy of radiation had a relative risk of death from breast cancer of 1.36, as compared with those exposed to less than 10 cGy (95 percent confidence interval, 1.11 to 1.67; P = 0.001). The data were most consistent with a linear dose-response relation. The risk was greatest among women who had been exposed to radiation when they were between 10 and 14 years of age; they had a relative risk of 4.5 per gray, and an additive risk of 6.1 per 10(4) person-years per gray. With increasing age at first exposure, there was substantially less excess risk, and the radiation effect appeared to peak approximately 25 to 34 years after the first exposure. Our additive model for lifetime risk predicts that exposure to 1 cGy at the age of 40 increases the number of deaths from breast cancer by 42 per million women. We conclude that the risk of breast cancer associated with radiation decreases sharply with increasing age at exposure and that even a small benefit to women of screening mammography would outweigh any possible risk of radiation-induced breast cancer.