The United States Environmental Protection Agency (U.S. EPA) supports a number of local community initiatives to encourage public involvement in decisions regarding environmental waste management and remediation. Native American tribal communities, in most cases, operate as sovereign nations, and thus have jurisdiction over environmental management on their lands. This paper provides examples of initiatives addressing Native American concerns about past radioactive waste management practices — one addresses uranium mining wastes in the Western United States and the other, environmental contamination in Alaska. These two projects involve the community in radioactive waste management decision-making by encouraging them to articulate their concerns and observations; soliciting their recommended solutions; and facilitating leadership within the community by involving local tribal governments, individuals, scientists and educators in the project. Frequently, a community organization, such as a local college or Native American organization, is selected to manage the project due to their cultural knowledge and acceptance within the community. It should be noted that U.S. EPA, consistent with Federal requirements, respects Indian tribal self-government and supports tribal sovereignty and self-determination. For this reason, in the projects and initiatives described in the presentation, the U.S. EPA is involved at the behest and approval of Native American tribal governments and community organizations. Objectives of the activities described in this presentation are to equip Native American communities with the skills and resources to assess and resolve environmental problems on their lands. Some of the key outcomes of these projects include: • Training teachers of Navajo Indian students to provide lessons about radiation and uranium mining in their communities. Teachers will use problem-based education, which allows students to connect the subject of learning with real-world issues and concerns of their community. Teachers are encouraged to utilize members of the community and to conduct field trips to make the material as relevant to the students. • Creating an interactive database that combines scientific and technical data from peer-reviewed literature along with complementary Native American community environmental observations. • Developing educational materials that meet the national science standards for education and also incorporate Native American culture, language, and history. The use of both Native American and Western (Euro-American) educational concepts serve to reinforce learning and support cultural identity. The two projects adopt approaches that are tailored to encourage the participation of, and leadership from, Native American communities to guide environmental waste management and remediation on their lands. These initiatives are consistent with the government-to-government relationship between Native American tribes and the U.S. government and support the principle that tribes are empowered to exercise their own decision-making authority with respect to their lands.

In most of the failed nuclear objects sitting processes the main barriers were not the technical or safety questions, but the problem of communication and reaching the public acceptance about decision. The paper presents one of the possible approaches that could contribute to the more community sensitive, transparent, communicative and finally successful siting process. The computer-supported approach for assessment of LILRW disposal impacts on regional and urban development is one of several modules loosely connected to a spatial simulation model in GIS. The results are shown as cartographic representation of impacts on different land uses and a possible future land use pattern. This model will be used for evaluation of alternative sites from regional and urban development impact aspects and for negotiating compensation schemes for local communities. One of the very important findings of this work is that the modeling outcomes depend more on value assumptions, according to which the input parameters are chosen, than on accuracy of computations. Therefore any conclusion based on the model outcomes should be verified in a wide panel of experts, interest groups and also wider public.

One from the problems connected to financial maintenance of the civil-law responsibility behind a harm and damage caused person and an environment, caused to the person, and radiation accident, has become definition of emergencies, with allowance for of possible consequences. By us was the made attempt partially to decide this problem on an example of an evaluation of the loss, which can be caused to health and enclosing natural medium owing to radiation accident with sources of an ionizing radiation during a realization of activity with them. The degree of risk connected to radiation accident was determined by a radiation source, that is sources of an ionizing radiation (SIR). Necessity of legal settlement of the relations in an orb of use of a nuclear energy stipulated by potential danger, with which connected activity in this area. After a proclamation of independence of Ukraine for seven years the nuclear legislation of Ukraine because of Law of Ukraine On use of a nuclear energy and radiation safety (1995) was created. The decree of the supreme Body of Ukraine On introduction in an operation of the Law of Ukraine About use of a nuclear energy and radiation safety provided acceptance of other acts in this area, namely: about a radiation guard of the population; about a civil liability for nuclear harms; about export and import of nuclear materials, nuclear installations special non-nuclear materials and services; about safe transportation of radioactive substances; about promptly of activity in an orb of use of a nuclear energy; About a physical guard of a nuclear material, nuclear installations and radioactive scraps; about production and use of uranium ores. July 12, 1996 the association of Ukraine to the Viennese Convention on a civil liability for nuclear harms - following pitch in directions of settlement of the relations, further development of co-operation between the states and creation of a reliable international mode of the nuclear responsibility was held. It provided development of the legal and financial mechanism of the responsibility rather promptly of activity on a principle it of the interdiction, for want of use of a nuclear material, without submission in competent organs of the warranties of financial maintenance of reimbursement of damage from radiation (nuclear) accident. The law of Ukraine On use of a nuclear energy and radiation safety makes the responsibility of the licensee - legal or natural person, which has the sanction to a realization of activity in an orb of use of a nuclear energy given when due thereunder to have financial possibilities of compensation for damages from accidents by an own means or at the expense of a means of the insurance companies. All SIR, the circulation with which provides availability of the licenses: the closed sources of ionizing radiation’s (CSIR), open sources of ionizing radiation’s (OSIR) and system generating an ionizing radiation. In structure CSIR I In entered: irradiating installations, flaw detectors, radioisotope gears, radioisotope thermoelectric generators. The fulfillment of work was conducted in two stages: I stage - creation of the list of sources of ionizing radiation; II stages - definition of possible radiation accidents with SIR. At the first stage of work, it was necessary to make the list of sources of ionizing radiation’s, with which the emergencies were considered. In report was represented categories SIR according to area of their application in the economic and scientific purposes in Ukraine. On an indication of possible influence on staff, population and environment, SIR were divided into three groups: 1 group SIR - during accident happens an external exposure of staff; 2 groups SIR - during accident happen a combined exposure of staff (external and internal); 3 groups SIR - during accident happen radiation contamination of an environment to potential danger of an internal exposure of staff and separate persons from the population. At the same time, for want of creation of the list SIR, was taken into account: • Potential danger SIR, connected with their kind, type and technology of the circulation with them; • Probability of origin of radiation accidents depends on a kind of activity with SIR. The review of the analyzed radiation situations has shown that them the violation of radiation safety has become causal. There are emergencies connected to detection not taken into account not discounted) SIR (more often on dustbin) less often. Probably, it is stipulated by absence of means on a burial place SIR, term of which use has terminated. To check such violation after its realization it is very complicated, but to provide, it is possible and it is necessary. On warning of such violations the creation of a state system of the account (record keeping) and monitoring of sources of an ionizing radiation in Ukraine - State Register SIR is directed, which introduction in operation is planned not later, 2002. The following, behind frequency of cases, violation - accidents connected to loss of monitoring above SIR during their operation depressurization. The reason of these violations again consists in absence of necessary financing for the contents in proper condition; installations (gears) completed SIR, in the correspondence with requests of radiation safety. Such situation can be under monitoring, can be stipulated and in time is eliminated. Less widespread, the SIR in a diverse kind. Such rough violation is connected to violation of the rules of the circulation with SIR during their storage (in storehouses or during transportation). It testifies to bad organization of activity with SIR by administration of establishment, where the accident was held. Among the least met violations - reasons of radiation accidents: loss SIR during transportation, breakaway of a shell in a chisel slit during geophysical work and going depressurization SIR. Summing up outcomes of work, the analytical evaluation of possible influence of radiation accidents with SIR on staff, population and environment for definition of sizes of radiation consequences was conducted. The analysis conducted a research has shown, that all radiation accidents, which was registered in the boundaries of Ukraine and Russia, become a corollary of violation of the rules of radiation safety. It allows to make the following conclusion: taking into account on danger, with which the connected activity with SIR, is necessary to save necessary means concerning guarantee of fulfillment of the rules of radiation safety according to requests acting in the field of a nuclear energy.

Bituminization technology in Jaslovské Bohunice is described. Practical experiences in bituminization of A-1 (GCHW type reactor), V-1 and V-2 (VVER type reactor) concentrates are presented. Main accent is given on fire-risk prevention, analytical control of inlet concentrates (concentration ofand), chemical and radiochemical control of bituminized product. Recomendations and outcomes for future operation are presented.

The testing begun in the framework of the CEC project ECP-4 “Decontamination technologies and strategies” have allowed to develop and to test new technology of the polluted soils decontamination by removal of the thin turf layer by the vibrating blade of the special machine (Turf-Cutter). The experiments were conducted at the radioactively contaminated soils of Ukraine and Belarus during 1992–2000. The machine “TURF HARVESTER” (USA) was used in the experiment. The first testing of the method was conducted on the well turfed radioecological polygon “Buryakovka”, 4 km from the Chernobyl NPP, with levels of contamination: 100 Ci/km2 by Cs-137, 80 Ci/km2 by Sr-90, 7 Ci/km2 by Pu-239. As the preliminary researches have shown, about 95% of the radionuclides were concentrated in the upper layer of the unploughed soil. In an outcome of tests on a selected plot the decontamination factor (Fd) 25–40 for different radionuclides was obtained. After removal of turf and opening the soil surface, the wind soil erosion and secondary resuspension the radionuclides was expect. It has not taken place, as special researches on an evaluation of the wind resuspension of radionuclides by the soil particles after the turf harvesting. This can be explained as follows. The vibrating blade does not decondence and decompose the soil layer remaining. At the same time, the thin turf and soil layer removal saves the vegetation regenerating organs and roots, which allows the grass restoration and surface fixation within one month after the experiment. The second test of a method was conducted on a polygon “Chistogalovka”, 3 km of the NPP. A high level of the radioactive contamination (150 Ci/km2 by Cs-137) and the weak turf cover of the rugged sand surface characterized the polygon. The turf removal at this polygon has allowed to receive Df = 10–15. Another testing was made at the Belorussian part of the Zone, which have demonstrated the possibility of the selected turf removal under the spotty radioactive contamination. The field gamma-spectrometer “Corad”, produced by the Kurchatov Institute (Russia), was used for the operative definition the highly contaminated spots. The selected removal of the mostly contaminated spots decreased the mass of the turf removed by 70%, obtaining the Df = 5–7. Next testing was conducted at the village Miliach (Rivne Province, Dubrovitsa district, Ukraine) at the pasture “Stav” with the contamination level by Cs-137 about 5 Ci/km2. This pasture was not influenced by any post-accident countermeasures. After the radioactive turf removal (Df = 15–20), the fodder grass was sow. The grass contamination was 15 times less, comparing to the control. The experimental fattening of 10 cows by a grass, skewed on the decontaminated plot, within 10 days, was carried on. A comparison the contamination of the milk from the experimental cows, which were fed by a grass of the turf-harvested plot, and the milk of the control cows, has shown the milk Df about 11 in 1993. The data obtained show high efficiency of the decontamination technology for the polluted soils based on the turf removal by the vibrating blade. Decontamination factor about 7–15 for the sandy and dusty-sandy soils with a weak turf layer up to 20–40 for the organic and wet silty soils with a strong turf layer was obtained. Important thing is, the best Dfs were obtained for the soils, which are critical on the intensity of the root uptake of the radionuclides. The high ecological and radioecological safety of the Turf-Cutter technology of the soil decontamination is also to be considered. The thin turf and soil layer removal does not deteriorate dramatically the migration situation and at the same time does not avoid the damaged ecosystem self-restoration. The volume of the matter harvested is comparatively low, because of the thin cutting. Being stored in the walls 2,5 m height, it occupies less 5% of the territory decontaminated, and the risk of migration the radionuclides outside the storage sites is comparative to those of the primary soil layouts. The field testing of the Turf-Cutter technology show correlation of its efficacy to the soil types, vegetation cover and the landscape conditions of the contaminated territory. It allowed, using some elements of the GIS-technologies and cartographic modeling, to prepare special evaluation and zonification the territories contaminated on the efficacy of the Turf-Cutter technology, and to identify the areas best for it’s mostly effective application. Following investigations confirm stable, long-term character of the improvements carried out. The sampling of 2000 at Miliach experimental plot shows the decontamination factor 10–11 for the grass and about 8 for milk. Moreover, as the Cs-137 still remains at the upper part of the soil profile, the Turf-Cutter technology is still actual for the territories of the post-Chernobyl radioactive contamination. Obviously, it can be suitable also for the removal of any other surface pollutant from the soil.

The Supercontainer (SC) design is the preferred option for the final disposal of high level nuclear waste and spent fuel in Belgium. The SC consists of a carbon steel overpack surrounded by a very thick concrete buffer, contained within a stainless steel envelope. In this highly alkaline environment of the annulus and under normal conditions (without the ingress of aggressive species), the carbon steel overpack will be protected by a passive oxide film, which is believed to result in very low and almost negligible uniform corrosion rates. This paper discusses the modeling efforts simulating the evolution of various parameters that can potentially influence the corrosion processes (temperature, pH, saturation degree and concentration of aggressive species). The outcome of the Belgian experimental program to study the passive uniform corrosion of carbon steel in concrete is also addressed. Long-term uniform corrosion rates are determined using different independent techniques (hydrogen gas measurements by means of manometric gas cells or pressure transducers, long-term passive current density measurements).

Large scale demonstration experiments in underground research laboratories (both onsite and off-site) are currently undertaken by most high level radioactive waste management organisations. The decision to plan and implement prototype experiments, which might have a life of several decades, has both important strategic and budgetary consequences for the organisation. Careful definition of experimental objectives based on the design and safety requirements is critical. The implementation requires the involvement of many parties and needs flexible but consequent management as, for example, additional goals for the experiments, identified in the course of the implementation, might jeopardise initial primary goals. The outcomes of an international workshop in which European and Japanese implementers (SKB, Posiva, Andra, ONDRAF, NUMO and Nagra) but also certain research organisations (JAEA, RWMC) participated identified which experiments are likely to be needed depending on the progress in implementing a disposal programme. Already earlier in a programme, large scale demonstrations are generally performed aiming at reducing uncertainties identified during the safety case development such as thermo-hydraulic-mechanical process validation in the engineered barrier system and target host rock. Also feasibility testing of underground construction in a potential host rock at relevant depth might be required. Later in a programme, i.e., closer to the license application, large scale experiments aim largely at demonstrating engineering feasibility and performance confirmation of complete repository components. Ultimately, before licensing repository operation, 1:1 scale commissioning testing will be required. Factors contributing to the successful completion of large scale demonstration experiments in terms of planning, defining the objectives, optimising results and main lessons learned over the last 30 years are being discussed. The need for international coordination in defining the objectives of new large scale demonstration experiments is addressed. The paper is expected to provide guidance to implementing organisations (especially those in their early stages of the programme), considering participating in and/or or conducting on their own large scale experiments in the near future.

On March 11, 2011, now two years ago, the magnitude 9.0 Great East Japan earth quake, Tohoku, hit off the Fukushima coast of Japan. This was one of the most powerful earthquakes in recorded history and the most powerful one known to have hit Japan. The ensuing tsunami devastated a huge area resulting in some 25,000 persons confirmed dead or missing. The perfect storm was complete when the tsunami then found the four-reactor, Fukushima-Daiichi Nuclear Station directly in its destructive path. Some 2 million people were evacuated from a fifty mile radius of the area and evaluation and cleanup began. A tremendous effort has been made, by many nationalities, since this time to restore this damaged plant and surrounding area and to return a great deal of the residents to their homes and farm lands. While most of the outcome of this unprecedented natural and manmade disaster was negative, both in Japan and worldwide, there have been some extremely valuable lessons learned and new emergency recovery technologies and systems developed to cope with the aftermath of this disaster. This paper describes new technology developed to selectively remove radioactive materials dangerous to workers, local citizens, and the natural environment from seawater used to cool the damaged reactors at Fukushima. As always, the mother of invention is necessity.

In order to offer a more cost effective, safer and efficient Intermediate Level Waste (ILW) management service, Energy Solutions EU Ltd. and Gesellschaft für Nuklear-Service mbH (GNS) have been engaged in the development of integrated radioactive waste retrieval, packaging and conditioning solutions in the UK. Recognising the challenges surrounding regulatory endorsement and on-site implementation in particular, this has resulted in an alternative approach to meeting customer, safety regulator and disposability requirements. By working closely with waste producers and the organisation(s) responsible for endorsing radioactive waste management operations in the UK, our proposed solutions are now being implemented. By combining GNS’ off-the-shelf, proven Ductile Cast Iron Containers (DCICs) and water removal technologies, with Energy Solutions EU Ltd.’s experience and expertise in waste retrieval, safety case development and disposability submissions, a fully integrated service offering has been developed. This has involved significant effort to overcome technical challenges such as on-site equipment deployment, active commissioning, conditioning success criteria and disposability acceptance. Our experience in developing such integrated solutions has highlighted the importance of working in collaboration with all parties to achieve a successful and viable outcome. Ultimately, the goal is to ensure reliable, safe and effective delivery of waste management solutions.

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.

In the UK the Office for Nuclear Regulation and the Environment Agency developed the Generic Design Assessment process in response to a request from the UK Government. The process allows the regulators to jointly assess new nuclear reactor designs, in advance of any site-specific proposals to build a nuclear power station. Two reactor types are currently being assessed within Generic Design Assessment: • AREVA and Electricite´ de France’s UK EPR ® ; • Westinghouse Electric Company’s AP1000 ® . This paper will present the outcome of the assessment of radioactive waste management within the Generic Design Assessment process. One aspect of particular interest is the management of spent fuel from proposed new reactors as the assessment is based on an assumption that it will be sent for disposal. Therefore the paper will specifically consider the management of spent fuel and how this affects the regulatory decisions. The paper will look at four aspects. The first of these is to give a short overview of the Generic Design Assessment process. This will be followed by a summary of the Generic Design Assessment Radioactive Waste Management assessment on the acceptability of: • The types of waste and spent fuel. • The plans for conditioning of the wastes. • The safety issues associated with short-term storage. • The safety issues associated with long-term storage. • The issues associated with the disposal of the wastes. • The safety issues associated with decommissioning the reactors. The third aspect will be to look at the work commissioned by the Office of Nuclear Regulation in support of the Generic Design Assessment of radioactive waste management and how this has affected the regulatory decisions. This work has looked at the long-term stability of spent fuel in storage and the potential faults associated with the storage and handling of the spent fuel. The paper will end with the main conclusions of the radioactive waste management assessment within Generic Design Assessment. Looking at how storage of spent fuel can affect transport, disposal and decommissioning and how work by licensees could alter these conclusions.

Like in many countries, polluted industrial sites also exist in Belgium. Although the contamination is purely chemical in most cases, they may also contain a radioactive component. For chemically contaminated sites, extensive regulations and methodologies were already developed and applied by the different regional authorities. However and essentially because radioactivity is a federal competence, there was also a necessity for developing a legal federal framework (including an ER-methodology [1]) for remediation of radioactive contaminated sites. Most of the so-called radioactive contaminated sites are exhibiting a mixed contamination (chemical and radiological), and hence the development of such methodology had to be in line with the existing (regional) ones concerning chemical contamination. Each authority having their own responsibilities with regard to the type of contamination, this makes it more complicated and time-consuming finding the best solution satisfying all involved parties. To overcome these difficulties the legal framework and methodology — including the necessary involvement of the stakeholders and delineation of each party’s responsibilities — has to be transparent, clear and unambiguous. Once the methodology is developed as such and approved, the application of it is expected to be more or less easy, logic and straightforward. But is this really true? The aim of this document is to investigate as well the impact of factors such as the type of radioactive contamination — levels of contamination, related to NORM activity or not, homogeneous or heterogeneous, the differences in licensing procedures,… — on the application of the developed methodology and what could be the consequences in the long run on the remediation process. Two existing case studies in Belgium will be presented ([2]). The first case deals with a historical radium contaminated site, the second one with a phosphate processing facility still in operation, both with (very) low levels of radioactivity but containing very large volumes of contaminated materials. These case studies will demonstrate that, although the applied methodology will be the same in both cases, the impact of e.g. sampling strategy, scenario definitions, modelisations, final destination of the land, presence of chemotoxic components, dose or risk assessments, uncertainties, derivation of clean-up radionuclide guidelines, stakeholder involvement and waste treatment could be important on licensing, cost-estimate, planning and final outcome of the environmental remediation activities to be executed.

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.

Nuvia Limited was contracted to design, build and operate a waste treatment plant to stabilise a quantity of about 300m 3 of active sludge stored in the External Active Storage Tanks (EAST) at the former United Kingdom Atomic Energy Authority (UKAEA) research site at Winfrith, UK. At the end of this process both the old and new plants are to be decommissioned and demolished with the minimisation of waste material volumes. The sludge was produced in the Steam Generating Heavy Water Reactor (SGHWR), which is now in the early stages of decommissioning. As part of the reorganisation of UKAEA, responsibility for the site now lies with RSRL (Research Sites Restoration Limited) with funding provided by the Nuclear Decommissioning Authority (NDA). The process of stabilization of the SGHWR sludge from the EAST tanks within 500 litre stainless steel drums in the newly constructed Winfrith EAST Treatment Plant (WETP) was completed in March 2010. The plant will now enter a full decommissioning phase since no further work on any other waste materials has been identified by the client. This paper describes the development of a decommissioning plan for the mixture of old and new structures such that this next process can be undertaken without delay. This involved the completion of a decommissioning safety case together with supporting documents to demonstrate best practice and close attention to ALARP issues. The potential techniques for decommissioning are discussed leading to the final processes selected for this project. The challenges to be overcome, particularly with the older plant items where local contamination of their structures is suspected, are also identified and addressed. A process of recovery of concrete core samples for radiochemical analysis and their outcomes will be included. In contrast, the means by which the newer WETP facility was shown to be capable of unrestricted demolition as a result of the care taken during the design and construction phase will also be explained in the context of the practicalities, minimisation of the costs and timescale of such operations. Finally, issues concerning the minimisation of waste volumes, particularly where tritium contamination is present, are addressed within a UK-based disposal environment. The link between these outcomes and the planning and execution of the various structure surveillance tests will also be described and potential waste volumes identified. These considerations may be of interest to other groups undertaking similar projects on mixtures of older and more recent plants in varying states of contamination.

The IAEA Coordinated Research Project (CRP) on cementitious materials for radioactive waste management was launched in 2007 [1, 2]. The objective of CRP was to investigate the behaviour and performance of cementitious materials used in radioactive waste management system with various purposes and included waste packages, wasteforms and backfills as well as investigation of interactions and interdependencies of these individual elements during long term storage and disposal. The specific research topics considered were: (i) cementitious materials for radioactive waste packaging: including radioactive waste immobilization into a solid waste form, (ii) waste backfilling and containers; (iii) emerging and alternative cementitious systems; (iv) physical-chemical processes occurring during the hydration and ageing of cement matrices and their influence on the cement matrix quality; (v) methods of production of cementitious materials for: immobilization into wasteform, backfills and containers; (vi) conditions envisaged in the disposal environment for packages (physical and chemical conditions, temperature variations, groundwater, radiation fields); (vii) testing and nondestructive monitoring techniques for quality assurance of cementitious materials; (viii) waste acceptance criteria for waste packages, waste forms and backfills; transport, long term storage and disposal requirements; and finally (ix) modelling or simulation of long term behaviours of cementations materials used for packaging, waste immobilization and backfilling, especially in the post-closure phase. The CRP has gathered overall 26 research organizations from 22 Member States aiming to share their research and practices on the use of cementitious materials [2]. The main research outcomes of the CRP were summarized in a summary report currently under preparation to be published by IAEA. The generic topical sections covered by report are: a) conventional cementitious systems; b) novel cementitious materials and technologies; c) testing and waste acceptance criteria; and d) modelling long term behaviour. These themes as well as all contributions of participating organizations were further developed in the individual reports to be presented in the IAEA publication. The CRP facilitated the exchange of information and research co-operation in resolving similar problems between different institutions and contributed towards improving waste management practices, their efficiency and general enhancement of safety.

Transport and processing of nuclear waste for treatment and storage can involve unique and complex thermal and fluid dynamic conditions that pose potential for safety risk and/or design uncertainty and also are likely to be subjected to more precise performance requirements than in other industries. From an engineering analysis perspective, certainty of outcome is essential. Advanced robust methods for engineering analysis and simulation of critical processes can help reduce risk of design uncertainty and help mitigate or reduce the amount of expensive full-scale demonstration testing. This paper will discuss experience gained in applying computational fluid dynamics models to key processes for mixing, transporting, and thermal treatment of nuclear waste as part of designing a massive vitrification process plant that will convert high and low level nuclear waste into glass for permanent storage. Examples from industrial scale simulations will be presented. The computational models have shown promise in replicating several complex physical processes such as solid-liquid flows in suspension, blending of slurries, and cooling of materials at extremely high temperature. Knowledge gained from applying simulation has provided detailed insight into determining the most critical aspects of these complex processes that can ultimately be used to help guide the optimum design of waste handling equipment based on credible calculations while ensuring risk of design uncertainty is minimized.

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).

The development of scenarios for quantitative or qualitative analysis is a key element of the assessment of the safety of geological disposal systems. As an outcome of an international workshop attended by European and the Japanese implementers, a number of features common to current methodologies could be identified, as well as trends in their evolution over time. In the late nineties, scenario development was often described as a bottom-up process, whereby scenarios were said to be developed in essence from FEP databases. Nowadays, it is recognised that, in practice, the approaches actually adopted are better described as top-down or “hybrid”, taking as their starting point an integrated (top-down) understanding of the system under consideration including uncertainties in initial state, sometimes assisted by the development of “storyboards”. A bottom-up element remains (hence the term “hybrid”) to the extent that FEP databases or FEP catalogues (including interactions) are still used, but the focus is generally on completeness checking, which occurs parallel to the main assessment process. Recent advances focus on the consistent treatment of uncertainties throughout the safety assessment and on the integration of operational safety and long term safety.

The THOR ® fluidized bed steam reforming process has been successfully operated for more than 10 years in the United States for the treatment of low- and intermediate-level radioactive wastes generated by commercial nuclear power plants. The principle waste stream that has been treated is ion exchange resins (IER) and Dry Active Waste (DAW), but various liquids, sludges, and solid organic wastes have also been treated. The principle advantages of the THOR ® process include: (a) high volume reduction on the order of 5:1 to 10:1 for IER and up to 50:1 for high plastic content DAW streams depending on the waste type and waste characteristics, (b) environmentally compliant off-gas emissions, (c) reliable conversion of wastes into mineralized products that are durable and leach-resistant, and (d) no liquid effluents for treatment of most radioactive wastes. Over the past ten years, the THOR ® process has been adapted for the treatment of more complex wastes including historic defense wastes, reprocessing wastes, and other wastes associated with the fuel cycle. As part of the U.S. Department of Energy (DOE) environmental remediation activities, the THOR ® dual bed steam reforming process has successfully processed: (a) Idaho National Laboratory (INL) Sodium-Bearing Waste (SBW), (b) Savannah River Tank 48 High Level Waste (HLW), (c) Hanford Low Activity Waste (LAW), and (d) Hanford Waste Treatment Plant Secondary Waste (WTP-SW) liquid slurry simulants. The THOR ® process has been shown in pilot plant operations to successfully process various simulated liquid, radioactive, nitrate-containing wastes into environmentally safe, leach-resistant, solid mineralized products. These mineralized products incorporate normally soluble ions (e.g. - Na, K, Cs, Tc), sulfates, chloride salts, and fluoride salts into an alkali alumino-silicate mineral matrix that inhibits the leaching of those ions into the environment. The solid mineralized products produced by the THOR ® process exhibit durability and leach resistance characteristics superior to borosilicate waste glasses. As a result of this work, a full-scale THOR ® process facility is currently under construction at the DOE’s Idaho site for the treatment of SBW and a full-scale facility is in the final design stage for the DOE’s Savannah River Site for the treatment of Tank 48 high level waste. Recent work has focused on the development of new monolithic waste formulations, the extension of the THOR ® process to new waste streams, and the development of modular THOR ® processes for niche waste treatment applications. This paper will provide an overview of current THOR ® projects and summarize the processes and outcomes of the regulatory and safety reviews that have been necessary for the THOR ® process to gain acceptance in the USA.

The paper describes the public participation from the viewpoint of a stakeholder and member of the public. The dialogue between various members of the Dounreay Stakeholders Committee vary widely and do not always seem to represent the views of the wider public. Whilst great care has been taken to select various options for the ultimate condition of the Dounreay site and these have been discussed at great length and the preferred option selected by consensus, there still appears to be some conflict within the local community. It is probable that if the local population had to vote on the options for the future of the Dounreay nuclear site the outcome would be vastly different from that of the Stakeholders Committee. Whilst the politicians have been elected by the people, they represent a distinctly anti-nuclear view (even to the extent of decommissioning) whereas many local people (especially the workers on the Dounreay site) would prefer to see a continuation of nuclear activity at Dounreay. The problem is not only with local politicians but at national level in Scotland itself where the Scottish National Party has formed a coalition with the Green Party on condition all nuclear activities are phased out.