In 2008, following the general French plan for nuclear waste management, Électricité de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage. In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as THermal Organic Reduction, “THOR” (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary waste to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite. In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, ‘Phase I’ of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, ‘Phase II’ testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite. Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides up to 95% for C-14 while minimizing mass loss to <5%, this would minimize the volume of any secondary waste streams and potentially lower the waste class of the larger bulk of graphite. Alternatively, if up to 95% decontamination of C-14 is achieved, the graphite may be completely gasified which could result in lower disposal volumes.

During the operation and dismantling of nuclear plants, a variety of radioactive liquid wastes accumulate, which must be supplied to a disposal chain suitable for final disposal. In the initial conditioning step, with regard to their chemical properties these waste waters are usually adjusted by water treatment systems (predominantly evaporation facilities) permanently installed in nuclear power plants, and then, they are concentrated up to a predefined content of solids. Afterwards, the developing evaporator concentrate is desiccated locally and suitable for final disposal by own permanently installed conditioning installations or by means of mobile plants. For a mobile conditioning of evaporator concentrates, GNS uses the drying plant of the type FAVORIT proven over decades. Each of the mobile FAVORIT processing plants is provided with a handling licence for Germany valid nationwide according to § 7 of the German Radiation Protection Ordinance (StrlSchV).

The study of embedding matrix of radioactive waste is an important issue concerning the need to assure the radiological safety requirements for final disposal. The conditioning of the radioactive wastes by cementation process involves also a structural investigation by X-ray diffraction (XRD) of the samples prepared with cement and complexing agents at different concentration. This paper gives useful information about the influence of complexing agents related to damages produced in concrete microstructures. The experimental data obtained on samples kept in real repository environment for a period up to 15 years, offer structural information regarding the interaction between the different decontamination agents with the major components of the matrix.

The coprecipitation is a robust and inexpensive process for the treatment of important volumes of low and intermediate radioactive level liquid wastes. Its major inconvenient is the huge volume of sludge generated. The purpose of this work is to optimize the industrial coprecipitation continuous process by achieving the following objectives: - maximize the decontamination efficiency; - minimize the volume of sludge generated by the process; - reduce the treatment cost decreasing the installation volume. An innovative reactor with an infinite recycling ratio was therefore designed. It is a multifunctional reactor composed of two zones: a perfectly mixed precipitation zone and a classifier to perform liquid-solid separation. The experiments are focused on the coprecipitation of strontium by barium sulphate. The effluent containing sulphate ions and the barium nitrate solution are injected in the reaction zone where strontium and barium coprecipitate as sulphates. The produced solid phase is returned into the reaction zone by the classifier and goes out slowly from the reactor bottom with a residence time much higher than the liquid phase. This creates both a high concentration of solid phase in the reaction zone and a high efficiency of decontamination. The experimental conditions simulate the industrial effluents. The total treatment flow rate is 17 L/h, with an effluent flow rate of 16 L/h and a reactive flow rate of 1 L/h, hence a mean residence time of 10 minutes. In these experimental conditions, the molar ratio sulphate/barium after mixing corresponds to 4.9. These conditions are used in the reprocessing plant of La Hague. The decontamination factor reached in these experimental conditions is excellent: DF = 1500. The decontamination factor obtained with the classical continuous process is only equal to 60. Different process parameters are studied in order to optimize the reactor/classifier: residence time, barium nitrate flow rate and racking flow rate. The decrease of barium nitrate flow rate reduces the volume of sludge generated by the process keeping a high efficiency of strontium decontamination: DF = 400. An excess of sulphate is necessary to perform an efficient decontamination, but the molar ratio sulphate/barium can be reduced to 3 instead of 4.9 used industrially. The reactor/classifier also represents an efficient device for the coprecipitation process intensification. Indeed, it can sensibly reduce the final installation size while treating important volume of effluents. This innovative reactor optimizes both the decontamination efficiency of radioactive liquid wastes and the reduction of sludge volume. A reduction of sulphate ions in the discharge is also possible, which is environmentally friendly.

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.

Basic characterization of the waste management process during decommissioning of nuclear installations is described in the presented paper. A brief description is given of conditional and unconditional release of materials into the environment. The paper deals also with metal melting as prospective decontamination technique which can significantly reduce metallic radioactive waste. The material and radioactivity flow in the decommissioning process should be followed using the integrated material flow tool that is implemented into the standardized analytical decommissioning parameters calculation code OMEGA. Applying the integrated material flow tool, it is possible to monitor radiological and physical properties of individual material items listed in the nuclear installation input database, from dismantling up to their release into the environment or disposal in repository. Using the OMEGA code and two model databases, several scenarios related to metal melting are evaluated. The impact of applying different input decommissioning parameters on the metal distribution is the main result discussed in the paper.

With the maturing of nuclear industry, there is an added burden on the Back End of fuel cycle. Radioactive facilities are in the need for refurbishment. This paper describes the steps adopted for managing such alpha contaminated unserviceable Glove Boxes. The first step consisted of in-situ encasement of individual Glove Boxes, encountering the challenges of low head room and space congestion in these laboratories with cognizance to regulatory requirement related to radiation safety. The second step was removal, transfer and placement of encased Glove Boxes in a dedicated facility under continuous surveillance. The glove boxes will remain stored in this facility until arrangements are completed for dismantling and volume reduction in another facility dedicated for this purpose. The final step is the development of an appropriate technique for dismantling/cutting of Glove Boxes in an alpha-tight facility constructed to prevent spread of airborne activity, collection of cut pieces, subsequent decontamination of these metallic wastes and disposal. The paper describes the design scheme for dismantling of the glove box. The description also includes hands on evaluation of tools and gadgets in pilot set-up with a view to incorporating the most credible choice in an upcoming active facility.

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.

A reform will be implemented in the helium purification system of the 10MW High Temperature Gas-cooled Test Reactor (HTR-10) in China. The measurement of the γ dose rates of facilities, including valves, pipes, dust filter, etc., in the purification system of the HTR-10, has been performed. The results indicated that most radiation nuclides are concentrated in the dust filter and facilities at the entrance of the helium purification system upstream of the dust filter. Other facilities have the same γ dose rate level as the background. Based on the previous study and experiences in AVR, the measurement results can be understood that the radioactive dust carried by the helium gas was filtered by the dust filter. It provides important insights for the decontamination and decommissioning of facilities in the primary loop, especially in the helium purification system of the HTR-10 as well as the High Temperature Reactor-Pebblebed Modules (HTR-PM).

Reliable methods to determine the contamination depth in nuclear building structures are very much needed for minimizing the radioactive waste volume and the decontamination workload. This paper investigates the geostatistical integration of in situ gamma-ray spectroscopy measurements of different spatial supports. A case study is presented from the BR3 decommissioning project, yielding an estimated reduction of waste volume of ∼35%, and recommendations are made for future application of the proposed methodology.

The UP1 plant in Marcoule reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. During more than 40 years, the decladding operations produced thousands of tons of processed waste, mainly magnesium and graphite fragments. In the absence of a French repository for the graphite waste, the graphite sludge content of the storage pit had to be retrieved and transferred into a newer and safer pit. So, this project consists in the full retrieval and transfer of 15 m3 of water mixed with graphite dust located in the decladding facility, as well as the complete cleanup and decontamination of the pit. The equipment and process necessary for retrieval operations were designed, built and tested. The process is mainly based on the use of two pumps (one to capture and the other one to transfer the sludge) working one after the other and a robotic arm mounted on a telescopic mast. A dedicated process was also set up for the removal of the biggest fragments. In the pit, the sludge retrieval and transfer operations have been almost completed. Most of the non-pumpable graphite fragments has been removed and transferred to a new storage pit. As irradiant fragments have been discovered in the pit, specific studies are in progress in order to remove them to the laboratory for dissolution. This work is expected to 2014.

The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue. In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc. The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing. Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters. This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e. taken along the main steam piping line. Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the components. Moreover the radioactivity in the crud thickness was measured. These values allowed finally to correlate the residence time in the acid attack ponds to the level of the achieved decontamination.

Since 1980, the Agency is responsible by law for the safe management of all radioactive waste produced in Belgium, including decommissioning wastes. By the law of 11 January 1991 and the implementing Royal Decree of 16/10/1991, ONDRAF/NIRAS has been entrusted with a mission concerning the decommissioning of nuclear facilities. This mission involves the collection and assessment of data concerning decommissioning forecasts for nuclear facilities, the approval of facilities’ decommissioning programmes, the establishment — in consultation with operators — of financing conditions for decommissioning, as well as the implementation of these programmes on request by the operator, or in the case of its failure to do so. This is the case for the company Best Medical Belgium SA located at Fleurus (MDS Nordion SA, till April 2011), which produced radioisotopes for medical applications and went bankrupt in 2012. These installations have been entrusted to ONDRAF/NIRAS. A plan of action was developed for taking-over the operations in the framework of remediation and decommissioning. Steps have been taken to integrate his new role as a nuclear operator. The installations of Best Medical Belgium SA are now referred to as the “O/N - Site Fleurus.” Nuclear facility operators, or any person requesting to operate a nuclear facility, are obliged to provide ONDRAF/NIRAS, under their responsibility and in due time, with all the necessary information concerning these facilities’ decommissioning forecasts, the nature, quantities and dates of transfer of the resulting waste, and the financing conditions for these facilities’ decommissioning. In order to make the necessary funds available for decommissioning a nuclear facility when it will be shut down, operators are obliged to establish provisions during the facility’s active life. These provisions are calculated in such a way that the total amount established at the time of the final shutdown covers all costs resulting from the facility’s final phase, namely the preparation of decommissioning, decontamination, more or less long-term maintenance, dismantling, treatment, conditioning and disposal of waste produced during this phase. In order to face to multiple and repeated evaluation processes, the Agency, already in the early 90’s, started with the implementation of its own integrated data processing system, recording the physical and radiological inventories of nuclear facilities, and allowing the evaluation of the quantities of decommissioning materials and wastes as well as of the decommissioning costs of these facilities. The cost evaluations cover all decommissioning activities from final shutdown of the facility until final release from nuclear control, as well as conventional demolition and site restoration if required. Beside the information related to the inventories, the database involves so called auxiliary tables integrating unit prices of the various decontamination and dismantling techniques, unit rates for radioactive waste processing, interim storage and final disposal.

Magnetite dissolution tests using a hydrazine base solution were performed at a temperature range of 90 to 150 °C. The dissolution rate of magnetite increased with [N 2 H 4 ], time, and temperature. The optimum solution pH in the experimental range was 3. The addition of copper ion to the hydrazine base solution greatly increased the magnetite dissolution rate. This was explained by the complex formation between N 2 H 4 and Cu ions, and the reducing power of the hydrazine-Cu complex to the ferric ions of magnetite. The reductive decontamination solution can be applied below 100 °C by the addition of copper ions. The chemical decontamination of a Type 304 stainless steel specimen using a hydrazine base reductive decontamination solution was also performed. The contact dose rate was greatly decreased by the repetitive application of NP and the hydrazine base solution.

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.

After the Fukushima Daiichi nuclear accident, Japan Atomic Energy Agency (JAEA) was chosen by the national government to conduct decontamination pilot projects at selected sites in Fukushima prefecture. Despite tight boundary conditions in terms of timescale and resources, the projects served their primary purpose to develop a knowledge base to support more effective planning and implementation of stepwise regional remediation of the evacuated zone. A range of established, modified and newly developed techniques were tested under realistic field conditions and their performance characteristics were determined. The results of the project can be summarized in terms of site characterization, cleanup and waste management. A range of options were investigated to reduce the volumes of waste produced and to ensure that decontamination water could be cleaned to the extent that it could be discharged to normal drainage. Resultant solid wastes were packaged in standard flexible containers, labelled and stored at the remediation site (temporary storage until central interim storage becomes available). The designs of such temporary storage facilities were tailored to available sites, but all designs included measures to ensure mechanical stability ( e.g. , filling void spaces between containers with sand, graded cover with soil) and prevent releases to groundwater (impermeable base and cap, gravity flow drainage including radiation monitors and catch tanks). Storage site monitoring was also needed to check that storage structures would not be perturbed by external events that could include typhoons, heavy snowfalls, freeze/thaw cycles and earthquakes.

On 26 April 1986, the worst nuclear power plant accident in history occurred at the Chernobyl plant in Ukraine (then part of the Soviet Union). The destruction of Unit 4 sent highly radioactive fallout over Belarus, Russia, Ukraine, and Europe. The object shelter — a containment sarcophagus — was built in November 1986 to limit exposure to radiation. However, it has only a planned 25-year lifespan and would probably not survive even a moderate seismic event in a region that has more than its share of such events. It was time to take action. One of the largest tasks that are in progress is the design and construction of the New Safe Confinement (NSC). The NSC is an engineered enclosure for the entire object shelter that includes a suite of process equipment. The process equipment will be used for the dismantling of the destroyed Chernobyl Nuclear Power Plant Unit. One of the major mechanical handling systems to be installed in the new safe confinement is the Main Cranes System. The planned decontamination and decommissioning or dismantling activities will require the handling of heavily shielded waste disposal casks containing nuclear fuel as well as lifting and transporting extremely large structural elements. These activities, to be performed within the new safe confinement, will require large and sophisticated cranes. The article will focus on the current progress of the new safe confinement and of the main cranes system for the decommissioning or dismantling activities.

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.

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.

Belgoprocess started the industrial decommissioning of the main process building of the former Eurochemic reprocessing plant in 1987. In 2008 the eastern part was torn down and in 2010 the central part of the building was removed. The demolition of the third and last part is planned in 2014. At that point the area will be restored to brown field conditions. During the 25 years of decommissioning the Eurochemic installations a lot of effort has been invested in improving the efficiency. The Belgian decommissioning policy is based on maximizing effort for decontamination with the goal of unconditional release and minimization of nuclear waste. As a result a number of installations have been implemented that reduce the processing costs, increase the output of decontamination and increase the speed of release measurements. • The abrasive blasting installation is used to remove a thin layer from metal components and concrete objects. • The concrete crushing and sampling installation is used as the final measurement before unconditional release of concrete debris. • The concrete spec installation is used to separate the contaminated part from the non contaminated part of concrete debris. As the end of the decommissioning project is approaching the time is right to look back and evaluate the course of the project. Looking back what were the key decisions during the project, what were important milestones and what could or should have been done differently knowing what we know now.