The paper aims to highlight recent developments at the Belgian Nuclear Research Centre SCK•CEN in experimental and numerical study of the coupled physical-chemical behaviour of concrete subject to chemical degradation. The discussion mainly focusses on three interlinked research projects covering novel experimental methods to study the alteration of hydraulic and transport properties during carbonation and calcium leaching, a pore scale numerical model to capture microstructural changes due to the above degradation processes and a generic multiscale model aimed at determining evolution of the properties of a macrostructure over the long term. The paper also describes supplementary continuum scale numerical studies concerning concrete-clay interactions and geochemical impact on the physical structure of concrete. Preliminary findings from these studies show encouraging results such as the development of novel leaching, water permeability and diffusion apparatus, a robust pore scale model based on Lattice-Boltzmann method and a mesoscale study focused on the importance of interfacial transition zones on the effective diffusivity for linear and nonlinear diffusion problems.

SCK•CEN is the Belgian Nuclear Research Centre. Founded in the mid-fifties, it has accumulated experience and know-how in all fields of the nuclear power production: in the neutronics calculation, radiation protection, waste management, fuel performance and analysis, nuclear measurements, radiochemistry, reactor technology, etc. Since 1989, SCK•CEN has launched Decommissioning activities to deal with the Technical Liabilities created by 40 years of operation. The main projects started were: • the dismantling of the BR3 PWR reactor; • the dismantling of active laboratories and the decontamination of buildings for unrestricted reuse; • the management of the waste arising from the refurbishment activities of the BR2, especially the management of the high active beryllium matrix. In 1989, the BR3 reactor, a Pressurized Water Reactor, was selected by the European Commission as one of the four pilot dismantling projects in the framework of the EC five year RTD program on dismantling nuclear installations. Through this project, SCK•CEN has built up a broad know-how in dismantling and decommissioning operations. This know how concerns the decontamination for dose rate reduction and/or free release of materials, teleoperated techniques for cutting highly activated components of a reactor, concrete decontamination techniques, characterization techniques of radioactive waste or for free release of components and development of decommissioning management and recordkeeping of material streams and of nuclear material accountancy. SCK•CEN is now actively involved in other decommissioning projects in Belgium and in expertise abroad. After giving an overview of the main achievements and the perspectives of the decommissioning of the BR3 reactor, the paper intends to present the involvement of SCK•CEN in the other decommissioning projects and to give an overview of our activities and capacities.

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation-analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storage rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage-room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were used, which must be removed in controlled conditions. The ventilation system is considered free from nuclear contamination but it contains asbestos. This paper covers the organization of the dismantling work, the technical execution aspect and conclusions already known (dismantling is ongoing as this is written).

In order to assess the risk of radiological consequences of incidents and accidents in nuclear facilities it is important to contemplate their frequency of occurrence. It has to be shown that incidents and accidents occur sufficiently seldom according to their radiological consequences i. e. the occurrence frequency of radiological doses has to be limited. This is even demanded by the German radiation protection ordinance (StrlSchV), which says that in nuclear facilities other than nuclear power plants (NPP) in operation and for decommissioning, the occurrence frequency of incidents and accidents shall be contemplated in order to prove the design of safety measures and safety installations. Based on the ideas of the ICRP64, we developed a risk based assessment concept for nuclear facilities, which fulfils the requirements of the German regulations concerning dose limits in normal operation and design basis accidents. The general use of the concept is dedicated to nuclear facilities other than nuclear power plants (NPP) in operation and for decommissioning, where the regulation of risk assessment is less sophisticated. The concept specifies occurrence frequency limits for radiation exposure dose ranges, i. e. the occurrence frequency of incidents and accidents has to be limited according to their radiological effects. To apply this concept, scenarios of incidents and accidents are grouped in exposition classes according to their resulting potential effective dose to members of the general public. The occurrence frequencies of the incidents and accidents are summarized in each exposition class whereas the sum must not exceed the frequency limits mentioned above. In the following we introduce the application of this concept in the assessment of the potential radiological consequences of the decommissioning of a nuclear research reactor. We carried out this assessment for the licensing process of the decommissioning on behalf of German authorities.

Concentration of complexing acids solutions (oxalic acid, tartaric acid and citric acid) used in radioactive decontamination by chemical method affects both the mechanical stability and the chemical stability of cement matrix. The paper presents the works performed related to the chemical pretreatment of these organic acids solutions using as neutralizing agent Ca(OH) 2 . In this way it was possible to increase the concentration of organic acids solutions used and the soluble complex radionuclides passing in chemical precipitates, these affecting in a smaller manner the mechanical stability of the cement matrix. The chemical pretreatment the effluents improve the precipitation and conditioning performances by cementation. Were prepared compositions with complexing agents and compositions for oxidative degradation tests to simulate the concentrations of secondary radioactive waste obtained from the primary circuit decontamination of the VVR-S research reactor. It has been studied the influence of chemical pretreatment of complexing acids solutions of different concentrations on the setting time. Also it was determined the compressive strength of mortar samples in which were embedded these solutions of chemically pretreated organic acids. The results shown that an optimum cement–solution ratio doesn’t have a significant impact on the setting time or on the mechanical properties.

The Philippine Nuclear Research Institute (PNRI) in collaboration with the interagency technical committee on radioactive waste has been undertaking a national project to find a final solution to the country’s low to intermediate level radioactive waste. The strategy adopted was to co-locate 2 disposal concepts that will address the types of radioactive waste generated from the use of radioactive materials. This strategy is expected to compensate for the small volumes of waste generated in the Philippines as compared to countries with big nuclear energy programs. It will also take advantage of the benefits of a shared infrastructure and R&D work that accompany such project. The preferred site selected from previous site selection and investigations is underlain by highly fractured “andesitic volcaniclastics” mantled by residual clayey soil which act as the aquifer or water bearing layer. Results of investigation show that the groundwater in the area is relatively dilute and acidic. Springs at the lower elevations of the footprint also indicate acidic waters. The relatively acidic water is attributed to the formation of sulfuric acid by the oxidation of the pyrite in the andesite. A preliminary post closure safety assessment was carried out using the GMS MODFLOW and HYDRUS softwares purchased through the International Atomic Energy Agency (IAEA) technical assistance. Results from MODFLOW modeling show that the radionuclide transport follows the natural gradient from the top of the hill down to the natural discharge zones. The vault dispersion model shows a circular direction from the vaults towards the faults and eventually to the creeks. The contaminant transport from borehole shows at least one confined plume from the borehole towards the creek designated as Repo1 and eventually follows downstream. The influx of surface water and rainfall to the disposal vault was modeled using the HYDRUS software. The pressure head and water content at the base of the foundation layer and the bottom of the concrete is where a significant reduction in water content can be observed. It is also noted that water content and pressure remain constant after one year.

The Spanish experience related to the decommissioning of nuclear facilities includes the decommissioning of the Vandello´s I Nuclear Power Plant, the decommissioning of the CIEMAT Nuclear Research Centre and the decommissioning of the Jose´ Cabrera Nuclear Power Plant. This paper reviews the key aspects of these projects and describes the lessons learned related to preparatory activities, auxiliary facilities, decommissioning technologies, material management and site remediation and release.

The United Kingdom nuclear research programme started in the 1940s. Research Sites Restoration Limited (RSRL) is responsible for the restoration of two sites which were at the forefront of this research, under a programme funded by the UK Nuclear Decommissioning Authority (NDA). These are the 100 hectare Harwell site in Oxfordshire and the 84 hectare Winfrith site on the south coast of England. The work performed on these sites covered a huge range of nuclides, combinations of nuclides, chemical and physical processes, far more complicated than a power station, for example. The sites have a complex history with records of hundreds of buildings, many kilometres of drainage systems, groundwater contamination issues and land areas which require remediation. Formal work towards site release began in the 1990s, but demolition and clearance for re-use started many years earlier. An efficient restoration programme requires appropriate quality data. It is vital to decide what you need to know and how well you need to know it. As part of this, a challenging number of factors need to be considered in its design. This paper discusses these factors using the examples of the approach used at the Harwell and Winfrith sites including: • historical knowledge and associated uncertainties; • relevant clearance criteria; • availability and limitations of surveying equipment; • effective targeted and validation sampling with appropriate analytical methods; • data capture and analysis techniques; • effective communication between RSRL and the relevant technical teams; • mapping technologies (Global Positioning Systems, Geographical Information Systems); • use of Babcock’s IMAGES land quality software tool; • integration of the above over long time scales. The RSRL programme of works at the Harwell and Winfrith Sites is producing large volumes of different types of information from decommissioning, site investigation and remediation projects. This will be required to be accessible and understandable to support the process of site release which will continue over many years. The paper illustrates the methods by which RSRL is using effective knowledge management to compile a verifiable record to support site release as the site restoration works progress.

The government of Iraq, through the Ministry of Science and Technology (MoST) is decommissioning Iraq’s former nuclear facilities. The 18 former facilities at the Al-Tuwaitha Nuclear Research Center near Baghdad include partially destroyed research reactors, a fuel fabrication facility and radioisotope production facilities. These 18 former facilities contain large numbers of silos and drums of uncharacterized radioactive waste and approximately 30 tanks that contain or did contain uncharacterized liquid radioactive wastes. Other key sites outside of Al Tuwaitha include facilities at Jesira (uranium processing and waste storage facility), Rashdiya (centrifuge facility) and Tarmiya (enrichment plant). The newly created Radioactive Waste Treatment Management Directorate (RWTMD) within MoST is responsible for Iraq’s centralized management of radioactive waste, including safe and secure disposal. In addition to being responsible for the uncharacterized wastes at Al Tuwaitha, the RWTMD will be responsible for future decommissioning wastes, approximately 900 disused sealed radioactive sources, and unknown quantities of NORM wastes from oil production in Iraq. This paper presents the challenges and progress that the RWTMD has made in setting-up a radioactive waste management program. The progress includes the establishment of a staffing structure, staff, participation in international training, rehabilitation of portions of the former Radioactive Waste Treatment Station at Al-Tuwaitha and the acquisition of equipment.

The THETIS research reactor on the site of the Nuclear Sciences Institute of the Ghent University has been in operation from 1967 until December 2003. This light-water moderated graphite-reflected low-enriched uranium pool-type reactor has been used for various purposes e.g. the production of radio-isotopes and activation analyses. During the first years its core power was 15 kW. In the early ’70, a core enlargement allowed for operation at typically 150 kW, while the maximum was allowed to be 250 kW. The fuel was 5% enriched uranium cladded with AISI304L stainless steel, with graphite plugs at both ends of the tubes. In order to decommission the reactor, the spent fuel and other nuclear materials present had to be removed from the reactor site. Ghent University entrusted SCK·CEN, the Belgian Nuclear Research Centre, with the study of the further management of the spent fuel. Various options such as reprocessing, intermediate storage awaiting final disposal were investigated. However the characteristics and the small amount of spent fuel (84.64 kg of UO 2 ) made these solutions very expensive. In the meantime ONDRAF/NIRAS, the Belgian radioactive waste management agency, was developing together with Belgoprocess, a solution for final conditioning in 400 liter drums and further intermediate storage of the spent fuel in its nuclear facilities at the BELGOPROCESS site in Dessel. This conditioned waste is foreseen to enter the future geological disposal site after the intermediate storage period only after 2050. Finally SCK·CEN recommended this solution for the back-end of the THETIS spent fuel and Ghent University declared this spent fuel as radioactive waste. Once the feasibility for conditioning and storage was demonstrated, further actions were taken in order to unload the spent fuel out of the reactor and to transport it to the PAMELA-installation at the Belgoprocess site in Dessel. Finally after receiving all necessary licensing authorisations from the FANC/AFCN, the Belgian nuclear safety authority, the operations started at the reactor site beginning of 2010 and the spent fuel was placed into the intermediate storage building after conditioning at the Belgoprocess site at the end of 2010. The paper will focus on: - the inventarisation and characterization of the spent fuel and other nuclear materials; - the operations at Ghent University and Belgoprocess sites; - the conclusions drawn from the operations.

The Government of Iraq (GOI) has undertaken efforts to decommission and dismantle former nuclear facilities. The GOI has only preliminary information on some of the former nuclear facilities. This paper will highlight the challenges involved in conducting inspections of the outlying former nuclear facilities in Iraq and present a brief summary of the results of those inspections. The facilities discussed in this paper are located at various sites throughout Iraq, from locations close to Baghdad to those in the north and far western desert areas. Some of the facilities, such as those at the Al Tuwaitha Nuclear Research Center have been visited and characterized. Other facilities, including the following, have not been visited or thoroughly characterized. • Al Jesira, Uranium feed stock production facility; • Adaya, Burial location for contaminated equipment; • Djerf al Naddah, Spent fuel storage facility; • Rashdiya, Centrifuge development center; • Al Qa’im, Uranium (yellowcake) production facility. The visits were conducted to develop an inventory of the buildings/structures that need to be included in decommissioning/dismantlement efforts. The number of buildings, type of construction, size and general condition of the buildings were noted. In addition, attempts were made to determine contamination levels on surfaces, equipment, rubble, etc. This information will be used to support the Iraqi decommissioning and dismantlement project. Because the facilities are scattered throughout the country of Iraq, significant planning and coordination was required to ensure personnel security. Teams consisting of individuals from the Iraqi Ministry of Science and Technology (MoST) and Americans were under military escort when traveling to and visiting the sites. Because of the security issues, time on the ground at each site was limited. This paper will highlight the challenges involved in conducting the inspections of the outlying former nuclear facilities In Iraq and present a brief summary of the results of those inspections.

CIEMAT, formerly the Nuclear Energy Board (JEN), is the Spanish Centre for Energy-Related, Environmental and Technological Research. Located in Madrid, it used to have more than 60 facilities in operation that allowed a wide range of activities in the nuclear field and in the application of ionising radiations. Particularly significant among these facilities were the research reactors, particle accelerators, hot cells and nuclear fuel manufacturing and processing plants. At present the centre, which is authorised as a single nuclear facility, includes various installations, some of them are now obsolete, shut down and in dismantling phases. In 2000 CIEMAT started the “Integrated plan for the improvement of CIEMAT installations (PIMIC)”, which includes activities for the decontamination, dismantling and rehabilitation of obsolete facilities. This paper will describe the decommissioning process carried out at CIEMAT Research Centre, and will review the key aspects of the project, including site remediation and release.

After more than 55 years of activities of the Nuclear Research Institute Rez (NRI) in the nuclear field, there are some obsolete nuclear facilities that shall be decommissioned. NRI is a leading institution in all areas of nuclear R&D in the Czech Republic. NRI has had a dominant position in the nuclear programme since it was established in 1955 as a state-owned research organization and it has developed to its current status. In December 1992, NRI has been transformed into a joint-stock company. The Institute’s activity encompasses nuclear physics, chemistry, nuclear power, experiments at the research reactor and many other topics. Main issues addressed in NRI in the past decades were concentrated on research, development and services provided to the nuclear power plants operating VVER reactors, development of chemical technologies for fuel cycle and irradiation services to research and development in the industrial sector, agriculture, food processing and medicine. The NRI operates two research nuclear reactors, many facilities as a hot cell facility, research laboratories, technology for radioactive waste (RAW) management, radionuclide irradiators, an electron accelerator, etc. The obsolete facilities to be decommissioned comprise various research facilities and facilities for RAW management. Decommissioning of nuclear facilities NRI is the only ongoing decommissioning project in the Czech Republic. Decommissioning started in 2003 and will be finished in 2014. Some facilities have already been successfully decommissioned.

In 2007, spent nuclear fuel (SNF) from the Nuclear Research Institute Rˇezˇ plc (NRI) was shipped to the Russian Federation for reprocessing. A large amount of SNF of Russian origin has been accumulated after 50 years of research reactor operation. The shipment was realized in the frame of the Russian Research Reactor Fuel Return (RRRFR) program under the US-Russian Global Threat Reduction Initiative (GTRI). SNF shipment from NRI to the Russian Federation represented a very complex and complicated technical, legal and contractual scope of work. The SNF shipment has been realized under specific conditions: 1. High capacity SˇKODA VPVR/M casks were used for transportation for the first time. 2. For the first time, high enriched uranium SNF from a research reactor has been sent to the Russian Federation from a European Union country under the appropriate intergovernmental agreements, legal regulations and conditions. NRI also participates in shipments of SNF from other countries within the framework of the RRRFR program. NRI participated in shipments of SNF from Bulgaria and Hungary in 2008, from Poland in 2009 and 2010, from Ukraine in 2010. Shipments from Belarus and Serbia are planned in 2010. The second shipment of the residue of high enriched SNF from NRI after changeover of the reactor operation to low enriched fuel will be implemented in 2013. The experiences gained during the SNF transportation are described in the paper together with the present and future NRI activities in support of the SNF shipment from other countries.

This paper presents results of a half-scale test performed by ESV EURIDICE, an Economic Interest Grouping between the Belgian Nuclear Research Centre (SCK•CEN) and the Belgian Agency for Radioactive Waste and Enriched Fissile Materials (ONDRAF/NIRAS). The primary objective of the test was to assess the feasibility of constructing the Supercontainer and to provide experimental data to validate modelling calculations obtained using the finite element program HEAT/MLS. The test focused on the early-age behaviour of the concrete matrix materials and the practical aspects of construction. Generally, the results obtained from the half-scale test confirm that it is feasible to construct the Supercontainer with currently available techniques. The results also validate scoping calculations obtained earlier with the finite element model. These findings contribute an important step to demonstrate the feasibility to construct the Supercontainer and to validate the Belgian Supercontainer concept proposed by ONDRAF/NIRAS for disposal of high level waste (HLW) in Belgium.

The Nuclear Research Institute Rez plc (NRI) after 55 years of activities in the nuclear field produced some environmental liabilities that shall be remedied. There are three areas of remediation: (1) decommissioning of old obsolete facilities (e.g. decay tanks, RAW treatment technology, special sewage system), (2) processing of RAW from operation and dismantling of nuclear facilities, and (3) elimination of spent fuel from research nuclear reactors operated by the NRI. The goal is to remedy the environmental liabilities and eliminate the potential negative impact on the environment. Remediation of the environmental liabilities started in 2003 and will be finished in 2014. The character of the environmental liabilities is very specific and requires special remediation procedures. Special technologies are being developed with assistance of external subcontractors. The NRI has gained many experiences in the field of RAW management and decommissioning of nuclear facilities and will use its facilities, experienced staff and all relevant data needed for the successful realization of the remediation. The most significant items of environmental liabilities are described in the paper together with information about the history, the current state, the progress, and the future activities in the field of remediation of environmental liabilities in the NRI.

The Government of Canada’s 1996 Policy Framework for Radioactive Waste Management establishes that waste owners are responsible for the management of their radioactive wastes. This includes the planning, funding, and implementation of long-term waste management initiatives. Within this context, the Government has established three separate programs aimed at addressing the long-term management of radioactive waste for which it has accepted responsibility. The largest of these programs is the Nuclear Legacy Liabilities Program (NLLP). The objective of the NLLP is to address radioactive waste and decommissioning liabilities resulting from 60 years of nuclear research and development at Atomic Energy of Canada Limited (AECL) sites in Canada. In 2005, the Government increased the value of this liability in its Public Accounts based on a new, 70-year long-term strategy and, in 2006, it implemented a $520 million 5-year work plan to initiate the strategy. The cost of implementing the full strategy is estimated at about $7 billion (current-day dollars). Canada’s Historic Waste Program is a second program that is designed to address low-level radioactive wastes across Canada that are not managed in an appropriate manner for the long-term and for which the current owner can not reasonably be held responsible. These wastes mainly emanate from the refining and use of radium in the 1930s and the very early days of the nuclear industry in Canada when radioactive ores were mined and transported long distances for processing. While the Historic Waste Program has been in place since 1982, the Government of Canada launched the Port Hope Area Initiative in 2001 to deal with the bulk of the waste. Finally, the Government of Canada has entered into two agreements with Canadian provincial governments on roles and responsibilities relating to the decommissioning of uranium mine and mill tailings sites. These agreements, one with the Province of Ontario and one with the Province of Saskatchewan, establish the responsibilities of each level of government to address circumstances where further decommissioning work is required and the producer can no longer be held responsible. The paper will provide an overview of these environmental remediation programs for radioactive waste and will describe recent progress and future challenges.

Nuclear legacy liabilities have resulted from 60 years of nuclear research and development carried out on behalf of Canada by the National Research Council (1944 to 1952) and Atomic Energy of Canada Limited (AECL, 1952 to present). These liabilities are located at AECL research and prototype reactor sites, and consist of shutdown reactors, research facilities and associated infrastructure, a wide variety of buried and stored waste, and contaminated lands. In 2006, the Government of Canada adopted a new long-term strategy to deal with the nuclear legacy liabilities and initiated a five-year, $520 million (Canadian dollars) start-up phase, thereby creating the Nuclear Legacy Liabilities Program (NLLP). The objective of the long-term strategy is to safely and cost-effectively reduce risks and liabilities based on sound waste management and environmental principles in the best interests of Canadians. The five-year plan is directed at addressing health, safety and environmental priorities, accelerating the decontamination and demolition of shutdown buildings, and laying the groundwork for future phases of the strategy. It also includes public consultation to inform the further development of the strategy and provides for continued care and maintenance activities at the sites. The NLLP is being implemented through a Memorandum of Understanding between Natural Resources Canada (NRCan) and AECL whereby NRCan is responsible for policy direction and oversight, including control of funding, and AECL is responsible for carrying out the work and holding and administering all licences, facilities and lands. The paper summarizes achievements during the first three years of program implementation in the areas of decommissioning and dismantling; waste recovery and environmental restoration; the construction of enabling facilities to analyze, handle and store the legacy waste; and, planning for the long-term management of the radioactive waste.

Whiteshell Laboratories (WL) is a Nuclear Research and Test Establishment near Winnipeg, Canada, operated by AECL since the early 1960s and now under decommissioning. WL occupies approximately 4400 hectares of land and employed more than 1000 staff up to the late-1990s, when the closure decision was made. Nuclear facilities at WL included a research reactor, hot cell facilities and radiochemical laboratories. Programs carried out at the WL site included high level nuclear fuel waste management research, reactor safety research, nuclear materials research, accelerator technology, biophysics, and industrial radiation applications. In preparation for decommissioning, a comprehensive environmental assessment was successfully completed [1] and the Canadian Nuclear Safety Commission issued a six-year decommissioning licence for WL starting in 2003 — the first decommissioning licence issued for a Nuclear Research and Test Establishment in Canada. This paper describes the progress in this first six-year licence period. A significant development in 2006 was the establishment of the Nuclear Legacy Liabilities Program (NLLP), by the Government of Canada, to safely and cost effectively reduce, and eventually eliminate the nuclear legacy liabilities and associated risks, using sound waste management and environmental principles. The NLLP endorsed an accelerated approach to WL Decommissioning, which meant advancing the full decommissioning of buildings and facilities that had originally been planned to be decontaminated and prepared for storage-with-surveillance. As well the NLLP endorsed the construction of enabling facilities — facilities that employ modern waste handling and storage technology on a scale needed for full decommissioning of the large radiochemical laboratories and other nuclear facilities. The decommissioning work and the design and construction of enabling facilities are fully underway. Several redundant non-nuclear buildings have been removed and redundant nuclear facilities are being decontaminated and prepared for demolition. Along with decommissioning of redundant structures, site utilities are being decommissioned and reconfigured to reduce site operating costs. New waste handling and waste clearance facilities have been commissioned and a large shielded modular above ground storage (SMAGS) structure is in final design in preparation for construction in 2010. The eventual goal is full decommissioning of all facilities and infrastructure and removal of stored wastes from the site.

The ever increasing pressure to reduce the release of radioactive and other toxic substances into environment requires constant improvement/upgrading of processes and technologies for treatment and conditioning of liquid radioactive wastes. To decontaminate liquid nuclear wastes, the coprecipitation process is the most commonly used in nuclear field because it can be applied to any type of aqueous effluents whatever their composition may be. This process deals with the in situ precipitation of solid particles to selectively remove one or more radioelements. In the nuclear research center of CEA (Commissariat a` l’Energie Atomique), the coprecipitation of 90 Sr with barium sulphate is the technique used to treat selectively this radionuclide. In a previous study (1), an elaborated model is presented which predicts the radioactive strontium decontamination factor of nuclear waste solutions which can be achieved by using a coprecipitation process with barium sulphate. The originality of this new approach lies in the possibility to simulate the decontamination process in non equilibrium conditions and at the reactor scale. This modelling combined with the resolution of the population balance, enables to identify the influence of process parameters (flow rates, stirring speed…) on crystal size and ultimately on decontamination. Simulations of the strontium coprecipitation with barium sulphate have been performed in continuous and semi batch reactors. Thanks to these simulations, laws of the treatment efficiency variation as a function of several process parameters (mean residence time, stirring speed, BaSO 4 concentration) have been determined and experimentally verified. This study leads to the determination of optimal treatment conditions. Three apparatus (recycling apparatus, fluidised bed and reactor/settling tank) providing these optimal conditions have been successfully tested and offered significant outlooks for the reduction of the residual sludge volume.