Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or negligible porosity and a water impermeable structure. Structural analysis shows that the glass in the composite has replaced the pores in the graphite structure. The typical pore volume of a graphite material is in the range of 20 vol.%. Therefore no volume increase will occur in comparison with the former graphite material. This IGM material will allow the encapsulation of i-graphite with package densities larger than 1.5 ton per cubic meter. Therefore a huge volume saving can be achieved by such an alternative encapsulation method. Disposal performance is also enhanced since little or no leaching of radionuclides is observed due to the impermeability of the material. NNL and FNAG have proved that IGM can be produced by hot isostatic pressing (HIP) which has several advantages for radioactive materials over the HVP process. • The sealed HIP container avoids the release of any radionuclides. • The outside of the waste package is not contaminated. • The HIP process time is shorter than the HVP process time. • The isostatic press avoids anisotropic density distributions. • Simple filling of the HIP container has advantages over the filling of an axial die.

Within the EURATOM FP7 Collaborative Project “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)”, a State-of-the-Art Report was prepared. The fast / instant release fraction (IRF) is defined as a fraction of the inventory of radionuclides that may be rapidly released from the fuel and fuel assembly materials at the time of canister breaching. In the context of safety analysis for a repository, the time span for mobilization of this fraction can be considered instantaneously, even if the process takes some time in experiments. Radionuclides contributing to the fast release are fission gases (Kr and Xe), easily soluble elements such as cesium and iodine, and other elements which are hardly incorporated in the UO 2 crystal lattice. The present contribution summarizes the results obtained from published studies focused on rapid release experiments carried out with different spent nuclear fuel (SNF), samples, sizes, techniques (batch and flow-through), and durations. A total of 80 experiments cover the study of UO 2 SNF from pressure water reactors (PWR) of different initial enrichments and burn-up, while 20 experiments were performed with UO 2 SNF from boiling water reactors (BWR) and 8 with MOX fuel.

Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the 14 N(n,p) 14 C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated by ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types × 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation. The total C-14 inventory was estimated as 4.46×10 14 Bq, consisting of 2.58×10 14 Bq for BWRs and 1.87×10 14 Bq for PWRs, and is consistent with the safety assessment of 4.4×10 14 Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72×10 13 Bq (13% of the total) in the previous assessment to 1.30×10 13 Bq (2.9% of the total; consisting of 1.48×10 12 for BWRs and 1.15×10 13 for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible.

There are approximately 56 million gallons (212km 3 ) of high level waste (HLW) at the U.S. Department of Energy (DOE) Hanford Site. It is scheduled that by the year 2040, the HLW is to be completely transferred to secure double-shell tanks (DST) from the leaking single-tanks (SST) via transfer pipeline system. Blockages have formed inside the pipes during transport because of the variety in composition and characteristics of the waste. These full and partial plugs delay waste transfers and require manual intervention to repair, therefore are extremely expensive, consuming millions of dollars and further threatening the environment. To successfully continue the transfer of waste through the pipelines, DOE site engineers are in need of a technology that can accurately locate the blockages and unplug the pipelines. In this study, the proposed solution to remediate blockages formed in pipelines is the use of a peristaltic crawler: a pneumatically/hydraulically operated device that propels itself in a worm-like motion through sequential fluctuations of pressure in its air cavities. The crawler is also equipped with a high-pressure water nozzle used to clear blockages inside the pipelines. The crawler is now in its third generation. Previous generations showed limitations in its durability, speed, and maneuverability. Latest improvements include an automation of sequence that prevents kickback, a front-mounted inspection camera for visual feedback, and a thinner wall outer bellow for improved maneuverability. Different experimental tests were conducted to evaluate the improvements of crawler relative to its predecessors using a pipeline test-bed assembly. Anchor force tests, unplugging tests, and fatigue testing for both the bellow and rubber rims have yet to be conducted and thus results are not presented in this research. Experiments tested bellow force and response, cornering maneuverability, and straight line navigational speed. The design concept and experimental test results are reported.

The great amount of water used for cooling the stricken power reactors at Fukushima Dai-ichi following the earthquake and tsunami of 11 March 2011 has resulted in accumulation of cooling water so-called the remaining water in some buildings. As the cooling water is subsequently contaminated by fission products (FPs) and some other radioactive substances, it is necessary to decontaminate this ‘cooling water’ to reduce the volume of liquid radioactive waste and to reuse it again for cooling the affected reactors. Some methods are applied to remove the radioactive substances from the cooling water. However, after treatments of water, there arises a secondary radioactive waste, the sludge. Steps are now taken to immobilize this sludge. In this paper, BaSO 4 , as one of main constituents of the sludge, was chosen as an immobilizing target substance. The appropriate manufacturing condition of glass waste form for loading the sludge (BaSO 4 ) was studied and the chemical durability was evaluated by measuring the dissolution rate. For this experiment, the iron phosphate glass (IPG), which is known to possess a large loading capacity for a variety of chemical substances, was employed as the glass medium. Based on experimental results, it is evident that BaSO 4 can be loaded into the IPG medium when it possesses the appropriate composition and melting temperature. During loading, BaSO 4 converted into BaO, acting as a network modifier, which leads to enhanced stability of IPG.

The economic, environmental, social and societal constraints have pushed the nuclear industry to develop new dismantling and decommissioning (D&D) techniques in order to meet the current industrial requirements. Since 5 years, AREVA, in partnership with Air Liquide, has been working on a new cryogenic solution that can be used to achieve the main important D&D operations (i.e. cutting, surface decontamination and concrete scabbling). This solution, called NiThrow™, is based on the Nitrocision LLC concept and consists in spraying highly pressurized liquid nitrogen (3500 bar) at a very low temperature (−140°C) onto a surface. The main advantage in using liquid nitrogen instead of water is due to its rapid conversion into gaseous nitrogen that will avoid the generation of liquid effluent. The preliminary trials made in France at Saint Ouen l’Aumône and at the AREVA SICN facility by using NiThrow™ appeared to be very interesting and promising in terms of effectiveness, reduction of the dosimetry (i.e. ALARA principle) and adaptability to the different types of surfaces and materials to be treated. On an other hand, they also highlighted a couple of defects like a lack of reliability that is not compatible with a safety use in a nuclear environment. In 2009, with the aim of fixing all these issues, AREVA decided to start an extensive R&D program. This work has been essentially focused on the lancing tool, the insulation and the flexibility of the liquid nitrogen distribution network, but also on the design and the manufacturing of a waste collection unit and a dedicated carrier. Today, these R&D efforts have been totally completed. and resulted to the publication of 3 patents. Also, the collaboration, with Air Liquide and the American company Conco allows AREVA to propose to its customers a safe and reliable solution for their cleaning operations in France and all over the world.

The Pile Fuel Storage Pond (PFSP) was built in 1948/50 to treat materials from the Windscale Piles. Multiple operational regimes over the intervening 60 years have resulted in a complex inventory of spent nuclear fuels, solid and liquid intermediate level wastes. A coordinated programme of work, designed to retrieve and safely dispose of the pond contents, has been implemented to enable the decommissioning of the facility. The long period of passive storage operations which preceded the implementation of the programme meant that the operator was faced with a dual challenge of providing new technical capability and changing a working culture that was inappropriate for the dynamic environment required to successfully deliver the programme. It was recognised that the nature of the programme meant that implementing a standard manufacturing approach to operations would not be appropriate. In order to create a dynamic retrievals focussed working culture, the operator has vigorously embraced change programmes aimed at improving a number of working practices including encouraging innovation, managing integrated but flexible production schedules, and encouraging workface problem solving. The combined impact of beginning to resolve the technical challenges and focussing on the delivery culture has resulted in the facility making a step change towards becoming fully retrievals operations focussed.

Knowledge of the radiological state of processes and equipment of a nuclear facility is essential to supervise a wide variety of sensitive tasks: building of intervention scenarios in order to optimize maintenance or dismantling operations, optimization of waste categorization, monitoring the effectiveness of decontamination processes, monitoring of nuclear facility decommissioning, etc. In order to meet the diversity of the issues involved, the CEA has developed in situ radiological characterization methods and techniques to acquire reliable radiological data. The data gathered is necessary to build robust radiological models which can be used as input data for dismantling studies. Over the last 30 years, the main nuclear measurement techniques, such as gamma imaging and gamma spectrometry, have been widely deployed by the CEA on many facilities under dismantling and more recently, on the Phénix nuclear power plant. Phénix was a small-scale prototype of a sodium-cooled fast breeder reactor, located at the Marcoule nuclear site. These techniques have been implemented on this reactor in order to meet the increased need for radiological knowledge to prepare for future dismantling operations following its final shutdown in 2009. This paper will focus on the description of three radiological characterization methods which take advantage of advanced nuclear measurement techniques. For each method, an example of a specific application on the Phénix reactor will be presented. Firstly, the so-called “gamma scanning” method will be explained. The objective of this method is to determine the activity profile of equipment based on collimated gamma spectrometry measurements with compact probes like CdZnTe. This method was applied to a neutron shielding of the reactor core to estimate the 60 Co activity profile. Then the measured activities helped to validate the theoretical activities resulting from neutron activation calculations. Secondly, this paper will focus on the interest of combining different measurement techniques such as gamma imaging, gamma spectrometry and collimated/uncollimated dose rate mapping to characterize complex equipment or processes. In this case, a specific methodology was developed to define the radiological state of a shielded cell used for the processing of irradiated nuclear fuels. Finally, an isotopic characterization technique using a high purity germanium detector will be discussed. This technique was applied to a non-irradiated fertile fuel sub-assembly in order to determine the level of uranium enrichment. The processing was carried out by three types of analysis: two automated, with the MGA-U and IGA software, and one manual.

This paper presents aspects of site decommissioning and clearance of a former fuel fabrication facility (development and production of fuel assemblies for research reactors and HTR) at Hanau (Germany). The main pathways for environmental contamination were deposition on soil surface and topsoil and pollution of deep soil and the aquifer by waste water channel leakage. Soil excavation could be done by classical excavator techniques. An effective removal of material from the saturated zone was possible by using advanced drilling techniques. A large amount of demolished building structure and excavated soil had to be classified. Therefore the use of conveyor detector was necessary. Nearly 100000 Mg of material (excavated soil and demolished building material) were disposed of at an underground mine. A remaining volume of 700 m 3 was classified as radioactive waste. Site clearance started in 2006. Groundwater remediation and monitoring is still ongoing, but has already provided excellent results by reducing the remaining Uranium considerably.

Belgium can rely on significant experience in the field of decontamination and decommissioning of nuclear facilities. Several projects are ongoing and include research reactors (BR3, Thetis), uranium and MOX fuel fabrication plants (FBFC International, Belgonucleaire), fuel reprocessing facilities (Eurochemic) and radwaste processing facilities (Belgoprocess). Additional projects are expected in the coming years with the planned final shutdown of the oldest nuclear power reactor units. Two national authorities are involved in the decontamination and decommissioning process of nuclear facilities. The FANC (together with its subsidiary Bel V) is concerned for all matters related to nuclear safety and radiation protection, while NIRAS/ONDRAF is concerned for all matters related to radioactive waste and fuel management and financial provisions. These attributions ensure that all safety and material concerns are addressed and that both the licensees and the national authorities bear their own responsibilities. They rely on an existing regulatory framework covering both the procedural and the technical aspects of the decontamination and decommissioning activities. However, opportunities for regulatory improvement were raised after some recent events in Belgium, among which the bankruptcy of a nuclear company producing radioisotopes, involving numerous additional interested parties in a complex judiciary context. Amendments in the current regulations are considered to increase the prerogatives of the authorities regarding the management of radioactive waste by a licensee, the transfer of an operating license from an operator to another, and the general decommissioning strategy of a facility. Furthermore, a dedicated “waste and decommissioning” working group within WENRA defined new reference levels applying to the decontamination and decommissioning of nuclear facilities. Belgium committed to include these requirements explicitly in its national legislation, even though most of them were already included in the existing decommissioning authorizations. Amendments will cover the safety provisions inherent to the decontamination and decommissioning phase, such as the decommissioning strategy, the qualification of techniques, the experience feedback valorization, the periodic safety reviews, the radioactive waste management, or the final characterization of the sites. Additionally, requirements regarding the authorization process will be detailed, such as the content of the licensee’s application file or the structure of the safety report covering the decontamination and decommissioning phase. These changes will contribute to a more secure regulatory framework for all interested parties.

The ability to monitor critical environment parameters of nuclear plants at all times, particularly during and after a disruptive accident, is vital for the safety of plant personnel, rescue and recovery crews, and the surrounding communities. Conventional hard-wired assets that depend on supplied power may be decimated as a result of such events, as witnessed in the Japanese Fukushima nuclear power plant in March 2011. Self-powered monitoring devices operating on a wireless platform, on the other hand, may survive such calamity and remain functional. The devices would be prepositioned at strategic locations, particularly where the dangerous build-up of contamination and radiation may preclude subsequent manned entrance and surveillance. Equipped with sensors for β-γ radiation, neutrons, hydrogen gas, temperature, humidity, pressure, and water level, as well as with criticality alarms and imaging equipment for heat, video, and other capabilities, these devices can provide vital surveillance information for assessing the extent of plant damage, mandating responses (e.g., evacuation before impending hydrogen explosion), and enabling overall safe and efficient recovery in a disaster. A radio frequency identification (RFID)-based system — called ARG-US — may be modified and adapted for this task. Developed by Argonne for DOE, ARG-US (meaning “watchful guardian”) has been used successfully to monitor and track sensitive nuclear materials packages at DOE sites. It utilizes sensors in the tags to continuously monitor the state of health of the packaging and promptly disseminates alarms to authorized users when any of the preset sensor thresholds is violated. By adding plant-specific monitoring sensors to the already strong sensor suite and adopting modular hardware, firmware, and software subsystems that are tailored for specific subsystems of a plant, a Remote Area Modular Monitoring (RAMM) system, built on a wireless sensor network (WSN) platform, is being developed by Argonne National Laboratory. ARG-US RAMM, powered by on-board battery, can sustain extended autonomous surveillance operation during and following an incident. The benefits could be invaluable to such critical facilities as nuclear power plants, research and test reactors, fuel cycle manufacturing centers, spent-fuel dry-cask storage facilities, and other nuclear installations.

Decommissioning of nuclear facilities is a complex process involving operations such as detailed surveys, decontamination and dismantling of equipment’s, demolition of buildings and management of resulting waste and nuclear materials if any. This process takes place in a well-developed legal framework and is controlled and followed-up by stakeholders like the Safety Authority, the Radwaste management Agency and the Safeguards Organism. In the framework of its nuclear waste and decommissioning program and more specifically the decommissioning of the BR3 reactor, SCK•CEN has developed different software tools to secure the waste and material traceability, to support the sound management of the decommissioning project and to facilitate the control and the follow-up by the stakeholders. In the case of Belgium, it concerns the Federal Agency for Nuclear Control, the National Agency for radioactive waste management and fissile material and EURATOM and IAEA. In 2005, BELGONUCLEAIRE decided to shutdown her Dessel MOX fuel fabrication plant and the production stopped in 2006. According to the final decommissioning plan (“PDF”) approved by NIRAS, the decommissioning works should start in 2008 at the earliest. In 2006, the management of BELGONUCLEAIRE identified the need for an integrated database and decided to entrust SCK•CEN with its development, because SCK•CEN relies on previous experience in comparable applications namely already approved by authorities such as NIRAS, FANC and EURATOM. The main objectives of this integrated software tool are: • simplified and updated safeguards • waste & material traceability • computerized documentation • support to project management • periodic & final reporting to waste and safety authorities. The software called DASAO (Database for Safeguards, Waste and Decommissioning) was successfully commissioned in 2008 and extensively used from 2009 to the satisfaction of BELGONUCLEAIRE and the stakeholders. SCK•CEN is now implementing a simplified release of the software for the management of the decommissioning of the THETIS reactor. Its decommissioning will start in March 2013 and will be completed by the end of 2014.

The Springfields facility manufactures nuclear fuel products for the UK’s nuclear power stations and for international customers. Fuel manufacture is scheduled to continue into the future. In addition to fuel manufacture, Springfields is also undertaking decommissioning activities. Today it is run and operated by Springfields Fuels Limited, under the management of Westinghouse Electric UK Limited. The site has been operating since 1946 manufacturing nuclear fuel. As part of the decommissioning activities, there was a need was to quantify contamination in a large redundant building. This building had been used to process uranium derived from uranium ore concentrate but had also processed a limited quantity of recycled uranium. The major non-uranic contaminant was Tc-99. The aim was to be able to identify any areas where the bulk activity exceeded 0.4 Bq/g Tc-99 as this would preclude the demolition rubble being sent to the local disposal facility. The problems associated with this project were the presence of significant uranium contamination, the realisation that both the Tc-99 and the uranium had diffused into the brickwork to a significant depth and the relatively low beta energy of Tc-99. The uranium was accompanied by Pa-234m, an energetic beta emitter. The concentration/depth profile was determined for several areas on the plant for Tc-99 and for uranium. The radiochemical analysis was performed locally but the performance of the local laboratory was checked during the initial investigation by splitting samples three ways and having confirmation analyses performed by 2 other laboratories. The results showed surprisingly consistent concentration gradients for Tc-99 and for uranium across the samples. Using that information, the instrument response was calculated for Tc-99 using the observed diffusion gradient and averaged through the full 225 mm of brick wall, as agreed by the regulator. The Tc-99 and uranium contributions to the detector signal were separated using a simple absorber, which essentially eliminated the Tc-99 count rate and reduced the uranium contribution only marginally. The outcome of the project was that it was possible to demonstrate that the complete building met the criterion for acceptance at the local waste facility.

In 2007, Cegelec has been attributed a contract by ANDRA under the ESDRED program (Engineering Studies and Demonstrations of Repository Designs) financed by 13 organisations across Europe, for the design, manufacturing and trial test of a full scale industrial demonstrator of HL-LL waste canister transfer system. Completed in 2009, this project gives a good synthesis of solutions the industry can offer in order to manage HLW transportation with regard to deep geological disposal. This paper highlights solutions developed and results achieved with numerous pictures of the actual system.

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.

A Deep Geologic Repository (DGR) for low and intermediate level radioactive waste has been proposed by Ontario Power Generation for the Bruce nuclear site in Ontario, Canada. As proposed the DGR would be constructed at a depth of about 680 m below ground surface within the argillaceous Ordovician limestone of the Cobourg Formation. This paper describes the hydrogeology of the DGR site developed through both site characterization studies and regional-scale numerical modelling analysis. The analysis provides a framework for the assembly and integration of the site-specific geoscientific data and examines the factors that influence the predicted long-term performance of the geosphere barrier. Flow system evolution was accomplished using both the density-dependent FRAC3DVS-OPG flow and transport model and the two-phase gas and water flow computational model TOUGH2-MP. In the geologic framework of the Province of Ontario, the DGR is located on the eastern flank of the Michigan Basin. Borehole logs covering Southern Ontario combined with site-specific data from 6 deep boreholes have been used to define the structural contours and hydrogeologic properties at the regional-scale of the modelled 31 sedimentary strata that may be partially present above the Precambrian crystalline basement rock. The regional-scale domain encompasses an approximately 18500km 2 region extending from Lake Huron to Georgian Bay. The groundwater zone below the Devonian includes units containing stagnant water having high concentrations of total dissolved solids that can exceed 300g/L. The Ordovician sediments are significantly under-pressured. The horizontal hydraulic conductivity for the Cobourg limestone is estimated to be 2 × 10 −14 m/s based on straddle-packer hydraulic tests. The low advective velocities in the Cobourg and other Ordovician units result in solute transport that is diffusion dominant with Peclet numbers less than 0.003 for a characteristic length of unity. Long-term simulations that consider future glaciation scenarios include the impact of ice thickness and permafrost. Solute transport in the Ordovician limestone and shale was diffusion dominant in all simulations. The Salina formations of the Upper Silurian prevented the deeper penetration of basal meltwater.

Paper presents the overall scope and actual results of the project for evaluation of representative scenarios for reuse of conditionally released materials from decommissioning. Aim of the project is to evaluate the possibilities of reuse of conditionally released steels and concrete in technical constructions which guarantee the long-term preservation of design properties over periods of 50–100 years. Interaction of conditionally released materials with public is limited and predictable due to design and purpose of selected constructions and due to fact that in many scenarios these materials are embedded in non-radioactive materials such as bars in reinforced concrete. Worker’s scenarios for preparation, operation and maintenance of these constructions are analysed in detail including the manufacturing of elements for these constructions. Project aims to evaluate the scenarios of reuse of conditionally released materials in a complex way in order to develop the data for designers of scenarios and to evaluate the volumes of conditionally released materials based on facility (to be decommissioned) inventory data. The long-term constructions considered are bridges, tunnels, roads, railway constructions, industrial buildings, power industry equipment and others. Evaluation covers following areas: • Analysis of activities for manufacturing of reinforcement bars, rolled steel sheets and other steel elements and analysis of activities for construction of evaluated scenarios in order to evaluate the external exposure of professionals performing these activities; • Analysis of external exposure of professionals involved in operation and maintenance of the long-term constructions; analysis of external exposure of public groups which are exposed to evaluated constructions; • Analysis of internal exposure of public groups from the radionuclides released from the evaluated scenarios based on models for migration of radionuclides from the long-term constructions to critical groups of public; • Based on evaluation of external and internal exposure both for public and workers, maximal concentration of individual radionuclides is defined for construction elements of evaluated scenarios, manufactured from conditionally released materials; • Evaluation of volumes of steels and concrete in the frame of a decommissioning project which fulfil the defined radioactivity concentration limits. Visiplan 4.0 3D ALARA software is used for evaluation of external exposure of professionals and public, GOLDSIM software for evaluation of internal exposure of public and OMEGA code for evaluation of volumes of conditionally releasable materials. Several other parallel papers proposed for ICEM 11 are presenting selected details of the project.

Decontamination facilities use various techniques to decontaminate solid substrates. The aim of these facilities is either to recover the substrate for a future second life, or to sufficiently lower the radioactivity level in order to reduce the final volume of high activity waste. One of these techniques remains in aqueous bathes under ultrasonic agitation. This technique is very suitable for the decontamination of small metallic pieces. Most of those pieces are covered with various greases or organic oils very resistant to classical aqueous washes. Thus, this oily layer contains some unfixed radionuclides that must be removed to reach the aimed decontamination factor. This urged decontamination facilities operating staff to consider additive molecules necessary to render aqueous washes consistent with such a contamination. These molecules, namely surfactants, act on the liquid surface to increase affinity between aqueous and oily phases. The surfactant formulations commonly used in French decontamination facilities are standard industrial formulations (F1, F2). Those formulations are generally designed by international manufacturers to be consistent with several applications. Thus, there isn’t any specific formulation fitted to nuclear applications. Today, according to potential modification of the final wastes conditioning matrix (for instance glass matrix instead of bitumen matrix), this standard formulations use could move to the use of some dedicated formulations for nuclear use. The glass matrix and related effluents concentration process required effluents with controlled chemical composition with regard to the presence of radionuclides. However, the composition of the industrial formulations used until today are not precisely known (according to trade secret), and sometimes can contain molecules unfitted to the final waste vitrification process. For instance, some sodium silicates are used to bring basic properties to the surfactant formulations whereas this kind of compounds is quite inappropriate to the concentration process (possible formation of gel in nitric acid with high concentration of silicates). The present study deals with the potential replacement of these industrial formulations by other ones with controlled chemical composition and concentration. Each new formulation consists of an assembly of well-known surfactants. The first stage of this study involves the degreasing power evaluation for every formulation. Solid and liquid greases detachment is assessed thanks to different techniques. For instance, the study of liquid greases detachment is achieved thanks to a CCD camera. The contact angle between the solid substrate and the grease is followed up during the degreasing step. This allows us to draw degreasing kinetics so as to compare the formulation effectiveness including industrial ones. The second stage of this work presents decontamination factor achieved during real operations with the new designed formulations. These tests were performed in an industrial decontamination facility and the efficiency of the new formulations is compared to the standard industrial ones.

The project include systematic scenarios analysis of conditionally released materials from the decommissioning of nuclear installations and the creation of new knowledge in this field, which will be used for implementing projects for reuse of these materials. New knowledge includes data about materials from the decommissioning (types of materials and radiological data on the basis of analysis of various scenarios). Scenarios contain information about conditionally released materials, data of the external exposure of personnel who will assemble those structures and who will be use the constructions up to the target scenario. Scenarios assume guarantee that the final products will be placed on the current position for a very long period from 50 to 100 years. The paper presents the review of activities for manufacturing of various steel construction elements made of conditionally released steels and activities for realisation of selected scenarios for reuse of construction elements. The ingots after melting of decommissioned radioactive steel materials are as the starting material for manufacturing of steel components. Ingots from the controlled area will be melted into induction furnace and the mixture of liquid steel will be alloyed for achieve of required chemical parameters. Typical steel products suitable for established scenarios are steel rebar of concrete, steel profiles of various forms, railway rails and rolled steel sheets. Target scenarios include an analysis of staff exposure during installation of steel constructions as well as exposure of individual from critical groups of population during their exploitation.

The research and development work presented in this paper was initialized by Andra in 2007. The work necessary for manufacturing and testing a full scale demonstrator is presently implemented. The case story is twofold. The first part is related to the initial development of a high performance concrete formulation used for fabricating concrete storage containers (containing Intermediate Level and Long Lived Waste primary canisters) to be stacked and emplaced into 400-m long concrete lined horizontal disposal vaults (also called cavern), excavated in the Callovo-Oxfordian clay host formation at a 550 to 600-m depth, with an inside diameter of approximately 8-m. The fabrication of the concrete boxes is illustrated. The second part presents the outcome at the end of the detailed design phase, for a system which is now being manufactured (for further test and assembly), for the emplacement of the concrete containers inside the vault. The application was engineered for remote emplacing a pile of 2 concrete containers (the containers are preliminarily stacked in a pile of 2, inside a hot cell, thanks to a ground travelling gantry crane). The emplacement process is justified and the related emplacement synoptic is illustrated. The test campaign is scheduled in 2011–2012. The successful completion of the technical trials is mandatory to confirm the mechanical feasibility of remotely emplacing concrete containers into large horizontal disposal caverns over long distances. The later display of the machinery at work in Andra’s showroom will be instrumental for the confidence building process involving the various stakeholders concerned by the public enquiry period (mid-2013) preceding the deep geological repository license application (2014–2015).