The Studsvik site was originally a research facility with many different activities going on. During the years some of the work was ended and the buildings and facilities were free released and some of them torn down. Three research reactors, one in Stockholm and two in Studsvik, have been decommissioned and their sites have been released for unrestricted use. The waste produced was included in the Studsvik waste management system. There are today ongoing decommissioning projects in Studsvik. One is the dismantling and free release of the old Active Central Laboratory, ACL, together with its ventilation building, ACF, another project is the decommissioning of the old evaporator facility. A recently completed project is the decommissioning of the Van de Graaff accelerator building in Studsvik. The Van de Graaff accelerator was in use from 1962 to 1989. During 1990–1997 work was performed in the building with the aim to clean-up the building and to radiologically map the building including sampling and decontamination for free release. In 1998 a permit for decommissioning of the free released building was given from SSI and during 1999 the building itself was demolished. Free release of metals have been carried out at Studsvik since 1987 and up to date 5700 tonnes have been melted of which 5000 tonnes have been free released. The aim of melting low-level scrap metals from the nuclear industry is to safely determine the radioactive content of the metals before the material is released for unrestricted reuse. Melting services are performed as a part of the decommissioning of the nuclear power plant Würgassen in Germany. Decontamination for decommissioning, melting and free release of the material has been performed on two steam generators from the shutdown Ågesta PHWR. The project was performed in 1992–93 and has been reported earlier. Studsvik has worked with decommissioning and free release since the 1980’ies. This paper gives some examples on different projects performed during these years. The paper also describes the procedures on how to release both buildings and material from regulatory control as well as pre-treatment method introduced in order to minimise the waste needed to be put into final storage.

Services to the nuclear industry started at the very beginning of the nuclear industry with the construction of the first commercial units. Based on the knowledge developed in laboratories, small units were constructed and gradually extended to the large plants actually in service. Over the past few years, reverse operations have been undertaken with the decommissioning of and the green-field operations for the oldest units. At all stages of this work as well as during the main outage operations, external firms have been involved. Progressively these companies have built up specific knowledge including both the technical requirements identical to those of non-nuclear interventions (mechanical, electrical, engineering…) and the more nuclear-related ones (quality and safety requirements, radwaste management, radiological aspects, alara principles…).

The management of the relatively large volumes of slightly radioactively contaminated material, arising from the decommissioning of nuclear facilities, represents a substantial fraction of the cost of such projects. The recycling of a relevant fraction of this material (or its reuse or disposal) without radiological restrictions, was identified by a Task Group of the OECD/NEA Co-operative Programme on Decommissioning, as a significant means of reducing such costs. The lack of internationally accepted “clearance levels” of radioactivity, at which the material could be utilised without radiological restrictions, seriously limits recycling as a waste management option. The emergence of the NORM/TENORM issue is of great significance for the discussion of clearance regulations. TENORM arisings occur in huge quantities, two to three orders of magnitude larger than those used in European studies on recycling in the nuclear industry, and the activity levels are generally the same as in very low to low-level nuclear waste. The regulation of TENORM is in its early stages. Their occurrence in a large number of industries, as well as their activity levels and quantities, has not been generally appreciated, even by regulatory authorities, until fairly recently. National and international bodies have suggested or are in the process of suggesting regulations for TENORM. The most important development is the publication of the European Commission Directive of May 1996 (ratified in May 2000) laying down basic safety standards for protection against ionising radiation, arising both in the nuclear and non-nuclear industries. The International Atomic Energy Agency has also started looking into this area in connection with the revision of its Safety Series 89 document. Significant to note is that both these bodies suggest release criteria into the general economy that are more relaxed for the radioactive materials from non-nuclear industries than for similarly contaminated material from nuclear industries. This issue is being taken up by several other bodies as well. This paper reviews the current debate and underlines the need for consistency in developing regulations and criteria for exemption and clearance of all radioactive materials regardless of their origin.

Thorium dioxide is an important material for the nuclear industry. In the last decade, there has been a renewal of interest in studying the feasibility of thorium based fuel reactor to decrease the minor actinides production during the burn-up. Furthermore the resistance of the thorium dioxide to aqueous corrosion can make this material attractive for immobilizing tetravalent actinides. Leaching tests of powdered samples of thorium dioxide calcinated at 1300°C showed that the normalized dissolution rate is very low (between 10 −6 and 10 −7 g/(m 2 .d) in acidic media, and 10 −9 –10 −10 g/(m 2 .d) after pH>3 when the formation of colloïdes occurs. Thorium dioxide which is isomorphic with the actinide dioxides such as UO 2 , PuO 2 allows the formation of solid solutions whatever the concentration of the actinide. Several solid solutions Th 1−x U x O 2 were synthesized with mole-ratios Th/(U+Th) ranging from x = 0 to 1. X-ray powder diffraction data allowed to check that the Vegard’s law is respected in all the range, and specific surface area was also measured. The resistance of the solid-solution to aqueous corrosion was measured as a function of several parameters (leaching time, leachate acidity, uranium concentration) and the kinetics of solid solutions dissolution was determined as a function of the uranium concentration. The stoechiometry of the release of both actinides was verified, however due to the oxidization of U (IV) in U (VI) in contact with the leachate, the dissolution rate of both thorium and uranium increases with the thorium substitution in the solid by uranium (TV).

We examined the suitability of taking smear tests (DIN ISO 7503.1) to evaluate surface activity. Special respect was given to mixtures of nuclides typical for the nuclear industry. We analysed the nuclides of samples of different fuel pools to evaluate the smear test measurement technique. Special calculations were carried out to examine the amount of nuclides that are difficult to measure due to their low maximum decay energy of less then 0.15 MeV (e. g. 55 Fe, 63 Ni). The response characteristic, detection sensitivity and the influence of the nuclide used for calibration were examined in detail on the basis of the nuclide vectors of fuel pool samples. We learned that especially the isotope standard used for calibration has a major influence for the suitability of the measurements. Calculation of the amount of these nuclides in relation to the effective dose (inhalation, ingestion) and the skin dose showed that they contribute little to the complete exposition.

Increasing numbers of nuclear power stations are reaching the end of their commercially useful lives. Their sites are being released today from regulatory control either on the basis of case by case criteria or on the basis of national regulations. The central basis for radiation protection of the public in both approaches is the individual dose criteria applied in the consideration of the release of the site. The management of the relatively large quantities of very low level radioactive material that arises during the decommissioning of the nuclear power stations being shut down has become a major subject of discussion, with very significant economic implications. Much of this material can, in an environmentally advantageous manner, be recycled for reuse without radiological restrictions. Much larger quantities — 2–3 orders of magnitude larger — of material, radiologically similar to the candidate material for recycling from the nuclear industry, arises in non-nuclear industries like coal, fertiliser, oil and gas, mining, etc. In such industries, naturally occurring radioactivity is artificially concentrated in products, by-products or waste to form TENORM (Technologically Enhanced Naturally Occurring Radioactive Material). There are, however, many significant strategic issues that need to be discussed and resolved, such as types of risks to be considered, disposal, commercial and other aspects which can and should influence decisions on release criteria. An important issue is to avoid “double” standards that are being proposed by international organisations for the nuclear and non-nuclear industries.

In the United States (U.S.) the nuclear waste issue has plagued the nuclear industry for decades. Originally, spent fuel was to be reprocessed but with the threat of nuclear proliferation, spent fuel reprocessing has been eliminated, at least for now. In 1983, the Nuclear Waste Policy Act of 1982 [1] was established, authorizing development of one or more spent fuel and high-level nuclear waste geological repositories and a consolidated national storage facility, called a “Monitored Retrievable Storage” facility, that could store the spent nuclear fuel until it could be placed into the geological repository. Plans were under way to build a geological repository, Yucca Mountain, but with the decision by President Obama to terminate the development of Yucca Mountain, a consolidated national storage facility that can store spent fuel for an interim period until a new repository is established has become very important. Since reactor sites have not been able to wait for the government to come up with a storage or disposal location, spent fuel remains in wet or dry storage at each nuclear plant. The purpose of this paper is to present a concept developed to address the DOE’s goals stated above. This concept was developed over the past few months by collaboration between the DOE and industry experts that have experience in designing spent nuclear fuel facilities. The paper examines the current spent fuel storage conditions at shutdown reactor sites, operating reactor sites, and the type of storage systems (transportable versus non-transportable, welded or bolted). The concept lays out the basis for a pilot storage facility to house spent fuel from shutdown reactor sites and then how the pilot facility can be enlarged to a larger full scale consolidated interim storage facility.

Radiological characterization of nuclear waste is essential for storage sites management. However, most of Long-Lived RadioNuclides (LLRN), important for long-term management, are difficult to measure since concentration levels are very low and waste matrices generally complex. In an industrial approach, LLRN concentrations are not directly measured in waste samples but assessed from scaling factors with respect to easily measured gamma emitters. Ideally, the key nuclide chosen ( 60 Co, 137 Cs) should be produced by a similar mechanism (fission or activation) as the LLRN of interest and should have similar physicochemical properties. However, the uncertainty on the scaling factors, determined from experimental and/or calculation data, can be quite important. Consequently, studies are performed to develop analytical procedures which would lead to determine precisely the concentration of LLRN in nuclear waste. In this context, the aim of this study was to determine the concentrations of three LLRN: 129 I (T 1/2 = 15.7×10 6 a), 41 Ca (T 1/2 = 9.94×10 4 a) and 10 Be (T 1/2 = 1.387×10 6 a) in spent resins used for primary fluid purification in Pressurized Water Reactors using Accelerator Mass Spectrometry (AMS) for measurement. The AMS technique combined mass spectrometry and nuclear physics to achieve highly efficient molecular and elemental isobars separation. Energies of several Million Electron-Volt transferred to the ions in the first accelerating part of specifically developed tandem accelerators lead to molecular isobars destruction through interaction with the argon gas used to strip the injected negative ions to positive ones. At the exit of the tandem accelerator, the energy acquired in both accelerating parts allows an elemental isobars separation based on their significantly different energy loss (dE) while passing through a thickness of matter dx that is proportional to their atomic number (Z) and inversely proportional to ions velocity (ν) according to the Bethe-Block law (1). (1) d E d x = k * Z 2 ν 2 The use of a particle accelerator in conjunction with a selective ion source, mass and energy filters and a high-performance detector thus allow unambiguously identifying and measuring analyte concentration against much more abundant interfering isobars. The development of AMS and of related applications have recently been extensively reviewed [1–3]. Up to now, the potentialities of the accelerator mass spectrometry technique were explored for the measurement of cosmogenic radionuclides produced in the Earth’s environment either in the atmosphere or in the Earth’s crust (in situ-production). Many applications aiming to date and/or quantify Earth surface processes have been developed in the fields of geology, geomorphology and planetary sciences as well as archeology paleoanthropology and biomedicine. The present study extends the scope of AMS to nuclear industry. Because AMS facilities are not widely accessible and difficult to handle, LLRN concentrations in nuclear waste are usually determined using Inductively Coupled Plasma Mass Spectrometry (ICP-MS) and radiometric techniques. However for the measurement of very low LLRN concentrations, AMS becomes the most effective measurement method with detection limits of 10 5 –10 6 atoms per sample. In this study, AMS measurements were performed using the French AMS national facility ASTER located at the Centre Européen de Recherche et d’Enseignement des Géosciences de l’Environnement (CEREGE). The challenge was to define a chemical treatment procedure allowing the measurement of the three nuclides, 10 Be, 41 Ca and 129 I, by AMS. Each method selection was based on three main requirements: 1) a quantitative recovery in solution of Be, Ca, I and key radionuclides after resin mineralization, 2) a selective extraction from the sample matrix and the separation from β-γ emitters ( 3 H, 14 C, 55 Fe, 59 Ni, 60 Co, 63 Ni, 90 Sr, 125 Sb, 134 Cs, 137 Cs) and isobars, 3) the precipitation of each element under the best suited forms (i.e. AgI, CaF 2 , BeO) for AMS measurements. The chosen methods were optimized on synthetic solutions and finally applied for the determination of the three LLRN concentrations in spent resins from a 900 MWe Nuclear Power Reactor.

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.

Soil is an essential component of all terrestrial ecosystems and is under increasing threat from human activity. Techniques available for removing radioactive contamination from soil and aquatic substrates are limited and often costly to implement; particularly over large areas. Frequently, bulk soil removal, with its attendant consequences, is a significant component of the majority of contamination incidents. Alternative techniques capable of removing contamination or exposure pathways without damaging or removing the soil are therefore of significant interest. An increasing number of old nuclear facilities are entering ‘care and maintenance’, with significant ground contamination issues. Phytoremediation — the use of plants’ natural metabolic processes to remediate contaminated sites is one possible solution. Its key mechanisms include phytoextraction and phytostabilisation. These are analogues of existing remedial techniques. Further, phytoremediation can improve soil quality and stability and restore functionality. Information on the application of phytoremediation in the nuclear industry is widely distributed over an extended period of time and sources. It is therefore difficult to quickly and effectively identify which plants would be most suitable for phytoremediation on a site by site basis. In response, a phytoremediation tool has been developed to address this issue. Existing research and case studies were reviewed to understand the mechanisms of phytoremediation, its effectiveness and the benefits and limitations of implementation. The potential for cost recovery from a phytoremediation system is also briefly considered. An overview of this information is provided here. From this data, a set of matrices was developed to guide potential users through the plant selection process. The matrices take the user through a preliminary screening process to determine whether the contamination present at their site is amenable to phytoremediation, and to give a rough indication as to what plants might be suitable. The second two allow the user to target specific plant species that would be most likely to successfully establish based on prevailing site conditions. The outcome of this study is a phytoremediation tool that can facilitate the development of phytoremediation projects, avoiding the need for in-depth research to identify optimal plant species on a case-by-case basis.

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.

Since the French Atomic Energy Commission (CEA) was founded in 1945 to carry out research programs on use of nuclear, and its application France has set up and run various types of installations : research or prototypes reactors, process study or examination laboratories, pilot installations, accelerators, nuclear power plants and processing facilities. Some of these are currently being dismantled or must be dismantled soon so that the DEN, the Nuclear Energy Division, can construct new equipment and thus have available a range of R&D facilities in line with the issues of the nuclear industry of the future. Since the 1960s and 1970s in all its centers, the CEA has acquired experience and know-how through dismantling various nuclear facilities. The dismantling techniques are nowadays operational, even if sometimes certain specific developments are necessary to reduce the cost of operations. Thanks to availability of techniques and guarantees of dismantling program financing now from two dedicated funds, close to 15 B€ for the next thirty years, for current or projected dismantling operations, the CEA’s Nuclear Energy Division has been able to develop, when necessary, its immediate dismantling strategy. Currently, nearly thirty facilities are being dismantled by the CEA’s Nuclear Energy Division operational units with its industrial partners. Thus the next decade will see completion of the dismantling and radioactive clean-up of the Grenoble site and of the facilities on the Fontenay-aux-Roses site. By 2018, the dismantling of the UP1 plant at Marcoule, the largest dismantling work in France, will be well advanced, with all the process equipment dismantled. After an overview of the French regulatory framework, the paper will describe the DD&R strategy, programme and feedback experience inside the CEA’s Nuclear Energy Division and its progress since ICEM 14 in 2011’s conference in Reims.

Metal surface cleaning appears to be one of the major priorities for industries especially for nuclear industries. The research and the development of a new technology that is able to meet the actual requirements (i.e. waste volume minimization, liquid effluents and chemicals free process…) seems to be the main commitment. Currently, a wide panel of technologies already exists (e.g. blasting, disk sander, electrodecontamination…) but for some of them, the efficiency is limited (e.g, Dry Ice blasting) and for others, the wastes production (liquid and/or solid) remains an important issue. One answer could be the use of a LASER light process. Since a couple of years, the Clean-Up Business Unit of the AREVA group investigates this decontamination technology. Many tests have been already performed in inactive (i.e. on simulants such as paints, inks, resins, metallic oxides) or active conditions (i.e. pieces covered with a thick metallic oxide layer and metallic pieces covered with grease). The paper will describe the results obtained in term of decontamination efficiency during all our validation process. Metallographic characterizations (i.e. SEM, X-ray scattering) and radiological analysis will be provided. We will also focus our paper on the future deployment of the LASER technology and its commercial use at La Hague reprocessing facility in 2013.

Processing liquid wastes frequently generates off gas streams with high humidity and liquid aerosols. Droplet laden air streams can be produced from tank mixing or sparging and processes such as reforming or evaporative volume reduction. Unfortunately these wet air streams represent a genuine threat to HEPA filters. High efficiency mist eliminators (HEME) are one option for removal of liquid aerosols with high dissolved or suspended solids content. HEMEs have been used extensively in industrial applications, however they have not seen widespread use in the nuclear industry. Filtering efficiency data along with loading curves are not readily available for these units and data that exist are not easily translated to operational parameters in liquid waste treatment plants. A specialized test stand has been developed to evaluate the performance of HEME elements under use conditions of a US DOE facility. HEME elements were tested at three volumetric flow rates using aerosols produced from an iron-rich waste surrogate. The challenge aerosol included submicron particles produced from Laskin nozzles and super micron particles produced from a hollow cone spray nozzle. Test conditions included ambient temperature and relative humidities greater than 95%. Data collected during testing HEME elements from three different manufacturers included volumetric flow rate, differential temperature across the filter housing, downstream relative humidity, and differential pressure (dP) across the filter element. Filter challenge was discontinued at three intermediate dPs and the filter to allow determining filter efficiency using dioctyl phthalate and then with dry surrogate aerosols. Filtering efficiencies of the clean HEME, the clean HEME loaded with water, and the HEME at maximum dP were also collected using the two test aerosols. Results of the testing included differential pressure vs. time loading curves for the nine elements tested along with the mass of moisture and solid material on each element at final dP. Plots of overall filtering efficiencies for DOP (spherical aerosol) and dry surrogate (aspherical aerosols) at specified dPs were computed for each filter. Filtering efficiencies were determined as a function of particle size. Curves were also generated showing the most penetrating particle size as a function of dP. A preliminary set of tests was conducted to evaluate spray location, duration, pressure, and wash volume for in-place cleaning the interior surface (reducing dP) of the HEME element. A variety of nozzle designs were evaluated and test results demonstrated the potential to overload the HEME (saturate filter medium) resulting in very high dPs and extensive drain times. At least one combination of spray nozzle design, spray location on the surface of the element, and spray time/pressure was successful in achieving extension of operational life.

Organic compounds of various kinds have been used in the nuclear industry for numerous duties in uranium chemical, metal and ceramic processing plants. In the course of the various operations undertaken, these organic compounds have become contaminated with uranic material, either accidentally or as an inevitable part of the process. Typically, the chemical/physical form and/or concentration of the uranic content of the organics has prevented disposal. In order to address the issue of contaminated liquid organic wastes, the National Nuclear Laboratory (NNL) has developed a suite of treatments designed to recover uranium and to render the waste suitable for disposal. The developed processes are operated at industrial scale via the NNL Preston Laboratory Residue Processing Plant. The Oil Waste Leaching (OWL) Process is a fully industrialised process used for the treatment of contaminated oils with approximately 200 tonnes of uranium contaminated oil being treated to date. The process was originally developed for the treatment of contaminated tributyl phosphate and odourless kerosene which had been adsorbed onto sawdust. However, over the years, the OWL process has been refined for a range of oils including “water emulsifiable” cutting oils, lubricating oils, hydraulic oils/fluids and “Fomblin” (fully fluorinated) oils. Chemically, the OWL process has proved capable of treating solvents as well as oils but the highly volatile/flammable nature of many solvents has required additional precautions compared with those required for oil treatment. These additional precautions led to the development of the Solvent Treatment Advanced Rig (STAR), an installation operated under an inert atmosphere. STAR is a small “module” (100 dm 3 volume) which allows the treatment of both water miscible and immiscible solvents. This paper discusses the challenges associated with the treatment of liquid organic wastes and the process developments which have allowed a wide range of materials to be successfully treated.

The Low Level Waste Repository (LLWR) is the primary facility for disposal of Low Level Waste (LLW) in the United Kingdom (UK), serving the UK nuclear industry and a diverse range of other sectors. Management of LLW in the UK historically was dominated by disposal to the LLWR. The value of the LLWR as a national asset was recognised by the 2007 UK Governmental Policy on management of solid LLW. At this time, analysis of the projected future demand for disposal at LLWR against facility capacity was undertaken identifying a credible risk that the capacity of LLWR would be insufficient to meet future demand if existing waste management practices were perpetuated. To mitigate this risk a National Strategy for the management of LLW in the UK was developed by the Nuclear Decommissioning Authority (NDA), partnered with LLW Repository Ltd. (the organisation established in 2008 to manage the LLWR on behalf of NDA). This strategy was published in 2010 and identified three mechanisms for protection of the capacity of LLWR – application of the Waste Hierarchy by waste producers; optimised use of existing assets for LLW management; and opening of new waste treatment and disposal routes to enable diversion of waste away from the LLWR. Since publication of the National Strategy, there have been significant efforts made by LLWR (on behalf of NDA) to drive strategy implementation and to optimise LLW management practices across the UK. This paper seeks to illustrate the challenges of strategy implementation in the context of the 2010 LLW National Strategy to provide a demonstration that “strategy is a commodity, execution is an art”. The paper describes: the key themes of the National Strategy; the key barriers to strategy implementation; the measures implemented by LLWR since publication of the National Strategy to support transformation of LLW management behaviours, with emphasis on how the initiative supports achieving the behavioural change; and providing a review of the past two years and a look forward to the planned evolution of National Strategy implementation over the next five years.

The impingement of a fluid jet onto a surface has broad applications across many industries. Within the UK nuclear industry, during the final stages of fuel reprocessing, impinging fluid jets are utilised to mobilise settled sludge material within storage tanks in preparation for transfer and ultimate immobilisation through vitrification. Despite the extensive applications of impinging jets within the nuclear and other industries, the study of two-phase, particle -laden, impinging jets is limited, and generally restricted to computational modelling. Surprisingly, very little fundamental understanding of the turbulence structure within such fluid flows through experimental investigation is found within the literature. The physical modelling of impinging jet systems could successfully serve to aid computer model validation, determine operating requirements, evaluate plant throughput requirements, optimise process operations and support design. Within this work a method is considered, capable of exploring the effects of process and material variables on the flow phenomena of impinging jets. This is achieved on a number of experimental test rigs of varying scale employing both intrusive and non-intrusive measurement techniques Particle image velocimetry (PIV), ultrasonic Doppler velocity profiling (UDVP) and high speed imaging, through to visual observations and direct measurements, are all techniques that can be deployed. The influence of a number of parameters on the erosion characteristics of sediment beds following application of an axisymmetric impinging jet is presented in detail. Bed erosion is found to be enhanced as the jet height above the sediment bed is increased, due to greater turbulence development. Different erosion characteristics, as jet outlet velocity increased, were found for the particulates tested; sand, fine Mg(OH) 2 (test simulant representative of waste sludge, has similar particle size to sand, 200–1000μm) and coarse Mg(OH) 2 (1000–2000μm). The crater diameter increased with increasing velocity as expected. However, the effect of the increase in velocity on the crater depth was very different, particularly for the coarse material which was found to re-deposit in the crater when the velocity increased above 1.3 ms −1 , most likely due to enhanced re-circulation at the higher velocities.

As the nuclear industry continues to grow throughout the world, we find that support from government officials, local business leaders and the general public is becoming more and more important. In order to help raise awareness and inform these various publics, AREVA expanded upon a best practice from its worldwide operations and recently established a Community Advisory Council in the United States. The member organizations represent a variety of grassroots and minority organizations from across the United States and are active in various ways in local, state and federal arenas. AREVA’s objective for the Council is simple — listen to concerns, engage in dialogue and raise awareness about the intrinsic link existing between energy, CO2 emissions, global warming, and economic growth, so these same people can make decisions when it comes to energy sources in the future. We want our members to help us better understand their communities, listen to their concerns and answer their questions openly and honestly. AREVA understands that this outreach and education are just the first steps toward helping clean energy sources grow. We must maintain regular dialog and operate in a safe manner, because in the long run, it is these community members who will ensure energy security for the country. And it is only by working together as an industry that we can ensure a safe, clean air future for generations to come, no matter where in the world we live.

Nowadays, nuclear industry is facing a crucial need in establishing radiological characterization for the appraisal and the monitoring of any remediation work. Regarding its experience in this domain, the French Alternative Energies and Atomic Energy Commission (CEA) of Fontenay-aux-Roses, established an important feedback and developed over the last 10 years a sound methodology for radiological characterization. This approach is based on several steps: - historical investigations; - assumption and confirmation of the contamination; - surface characterization; - in-depth characterization; - rehabilitation objectives; - remediation process. The amount of measures, samples and analysis is optimized for data processing using geostatistics. This approach is now used to characterize soils under facilities. The paper presents the radiological characterization of soils under a facility basement. This facility has been built after the first generation of nuclear facilities, replacing a plutonium facility which has been dismantled in 1960. The presentation details the different steps of radiological characterization from historical investigations to optimization of excavation depths, impact studies and contaminated volumes.

The UK nuclear industry has in its inventory legacy waste in the form of complex, polydisperse and “polydense” suspensions, slurries and sludges in a variety of storage and transport vessels. This waste has been difficult to characterise because of radioactivity and limited accessibility, and conditioning and disposal of the waste presents a continuing challenge. In addition, the mechanisms by which very dense particles are transported in pipes are not well understood. Our objectives are to investigate the effect of mono- and bidisperse suspensions with a range of particle sizes and densities on the turbulence characteristics, transport and settling behaviour of slurries that are chosen to be analogues of those found on nuclear sites. Two versatile slurry pipe-flow loops of different diameters have been commissioned which can be operated over a large range of Reynolds numbers and are amenable to ultrasonic measurement methods. Details of the flow loops are presented, including optimisation studies. Results are presented for a variety of particle characterisation studies that have been performed on the particle species that form the suspensions, along with mean and RMS (root mean square) velocity profiles over a range of Reynolds number and particle concentration. In particular, the effect of particle concentration on the formation of settled beds, and mean flow velocity and turbulence characteristics has been investigated.