[en] The scope and objectives of Work Group for heterogeneous waste are presented and the specific objectives for sealed sources concerning safety evaluation, disposal options, categorizing and methodology for derivation of waste acceptance criteria are given. It is proposed to use the ISAM methodology and to identify group of test case for disused and other heterogeneous sources. The report of the work of the sub-groups is given. A test case inventory proposal is proposed, including the use of the real Saratov site inventory

No abstract available

[en] In France, low- and intermediate-level radioactive wastes are disposed in a near-surface facility, at Centre de l'Aube disposal facility. This facility, which was commissioned in 1992, has a disposal capacity of one million cubic meters, and will be operated up to about 2050. It took over the job from Centre de la Manche, which was commissioned in 1969 and shut down in 1994, after having received about 520,000 cubic meters of wastes. The Centre de l'Aube disposal facility is designed to receive a many types of waste produced by nuclear power plants, reprocessing, decommissioning, as well as by the industry, hospitals and armed forces. The limitation of radioactive transfer to man and the limitation of personnel exposure in all situations considered plausible require limiting the total activity of the waste disposed in the facility as well as the activity of each package. The paper presents how ANDRA has derived the activity-related acceptance criteria, based on the safety analysis. In the French methodology, activity is considered as end-point for deriving the concentration limits per package, whereas it is the starting point for deriving the total activity limits. For the concentration limits (called here LMA) the approach consists of five steps: the determination of radionuclides important for safety with regards to operational and long-term safety, the use of relevant safety scenarios as a tool to derive quantitative limits, the setting of dose constraint per situation associated with scenarios, the setting of contribution factor per radionuclide, and the calculation of concentration activity limits. An exhaustive survey has been performed and has shown that the totality of waste packages which should be delivered by waste generators are acceptable in terms of activity limits in the Centre de l'Aube. Examples of concentration activity limits derived from this methodology are presented. Furthermore those limits have been accepted by the French regulatory body and constitute a key point of ANDRA waste acceptance criteria. (author)

[en] With major projects ongoing or being planned, and also with the daily management of radioactive waste from nuclear facilities, the role of transport and/or storage packaging has been often overlooked. Indeed, the packaging development process and transport solutions implemented are a key part of the waste management challenge : protection of people and of the environment. During over four decades, the AREVA Group has developed a complete and coherent system for the transport of waste produced by nuclear industries. The transport solutions integrate the factors to consider, as industrial transportation needs, various waste forms, associated hazards and current regulations. Thus, TN International has designed, licensed and manufactured a large number of different transport, storage and dual purpose cask models for residues and all kinds of radioactive wastes. The present paper proposes to illustrate how a company acting both as a cask designer and a carrier is key to the waste management issue and how it can support the waste management policy of nuclear waste producers through their operational choices. We will focus on the TN International technical solutions implemented to guarantee safe and secure transportation and storage solutions. We will describe different aspects of the cask design process, insisting on how it enables to fulfil both customer needs and regulation requirements. We will also mention the associated services developed by the AREVA Business Unit Logistics (TN International, TRANSNUCLEAR, MAINCO, and LMC) in order to manage transportation of liquid and solid waste towards interim or final storage sites. (authors)

[en] The Centre de l'Aube Disposal Facility (Centre de Stockage de l'Aube) is designed to receive a wide variety of waste produced by nuclear power plants, reprocessing, decommissioning, as well as the industry, hospitals and armed forces. Such a variety of wastes incur highly different risks which must be grasped in the safety analysis of the Centre. This article attempts to show how a number of safety analysis tools are used to meet the highly varied needs of the waste producers and guarantee safe disposal. They involve functional analysis, risk analysis and safety calculations. The paper shows that the most important acceptance criteria for the first containment barrier, namely the waste package, are containment, durability, activity limitation and biological shielding. And a method is proposed to determine some of these criteria from safety scenarios (scenarios of accidents in operation, intrusion in the post-institutional control phase). Over the years, however, the waste producers have asked the Agence Nationale pour la Gestion des Dechets Radioactifs (ANDRA) to accept new types of waste not initially anticipated in the design criteria, and the safety analysis must imagine new scenarios and develop new acceptance criteria. The paper gives the example of sealed sources, closure heads of NPP vessels, racks for fuel elements, contaminated manipulators, irradiating waste, etc, which incur specific risks. In fact, some of this waste represent a source of unusual irradiation, a risk of further contamination in an accidental situation, or simply increase the likelihood of occurrence of certain scenarios, such as retrieval in the post-institutional control phase. The safety analysis must adapt and imagine specific scenarios to judge the acceptability of such waste, and must identify the acceptance criteria commensurate with the risks. The paper offers examples of research, some of it still under way at ANDRA. (author)

[en] The International Atomic Energy Agency has developed an approach to derive quantitative waste acceptance criteria for near surface disposal of radioactive waste. This approach has been successfully used to derive activity limits with regards to post-closure safety assessment. In a second step it has been recognised as necessary to take into account operational safety considerations and to study how this approach could be used for the derivation of operational safety acceptance criteria. The paper presents the work currently being undertaken for the operational phase, following the step by step approach already developed for the post-closure phase, with particular emphasis on the assessment context, system description and the selection of scenarios. (author)

[en] France has chosen a recycling strategy to manage nuclear spent fuels. This strategy relies on high level waste immobilization into a borosilicate glass waste form in order to meet both interim storage and final geological disposal safety requirements. Interim storage facilities at La Hague recycling plant have been designed to last more than an hundred years (concrete structure) factoring the need for a glass waste form durability using for instance natural or forced convection principles. This period of time during interim storage contributes to the final disposal cost optimization by allowing the radioactive decay of the main contributors to thermal power dissipated in the early years after glass production. In parallel to the thermal design of the interim storage, lots of studies have been carried out on the glass thermal and irradiation stabilities. Thermal treatment experiments consolidated by modeling show that the glassy state is expected to be stable during the interim storage period, with no crystallization induced by the glass thermal history. Moreover, the impact of the radiations (beta and alpha decays) expected in interim storage has been studied by external irradiation, glass actinide doping technique and molecular dynamic simulation. The results have demonstrated that the glassy state will not be modified during the interim storage period and the glass will fully preserved its role of conditioning material. (author)