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

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.

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.

Siting a deep geological repository for radioactive waste essentially involves two interrelated steps: deciding on an appropriate geological environment for the underground facilities and selecting a suitable location for the associated surface facility. An acceptable solution is more easily achieved if some flexibility exists for siting the surface facility, irrespective of the exact position of the underground facilities. Such flexibility is available if a ramp is used as the main access route from the surface facility to the underground facilities. Another option is to use a combination of shafts and (sub)horizontal tunnels as the main access route. Both variants include shafts for ventilation, etc. In this paper, the two variants (i) main access via ramp and (ii) main access via shaft are compared in terms of long-term safety. To this end, the entire network of underground tunnels of a deep geological repository is implemented in an analytical resistor network flow model. Radionuclide release through the tunnel system and the host rock is then calculated with a numerical network transport model, using as input the results from the flow model. The results clearly indicate that, even in case of hypothetically deficient horizontal and subhorizontal sealing elements, the choice between ramp and shaft as the main access route is irrelevant to long-term safety.

In large cement-based structures such as a near surface disposal facility for radioactive waste voids and cracks are inevitable. However, the pattern and nature of cracks are very difficult to predict reliably. Cracks facilitate preferential water flow through the facility because their saturated hydraulic conductivity is generally higher than the conductivity of the cementitious matrix. Moreover, sorption within the crack is expected to be lower than in the matrix and hence cracks in engineered barriers can act as a bypass for radionuclides. Consequently, understanding the effects of crack characteristics on contaminant fluxes from the facility is of utmost importance in a safety assessment. In this paper we numerically studied radionuclide leaching from a crack-containing cementitious containment system. First, the effect of cracks on radionuclide fluxes is assessed for a single repository component which contains a radionuclide source (i.e. conditioned radwaste). These analyses reveal the influence of cracks on radionuclide release from the source. The second set of calculations deals with the safety assessment results for the planned near-surface disposal facility for low-level radioactive waste in Dessel (Belgium); our focus is on the analysis of total system behaviour in regards to release of radionuclide fluxes from the facility. Simulation results are interpreted through a complementary safety indicator (radiotoxicity flux). We discuss the possible consequences from different scenarios of cracks and voids.

In 2009, remediation was initiated for a non-operational fuel cycle facility previously used for government contract work located in Windsor, Connecticut, USA. Radiological contaminants consisted primarily of high enriched uranium (HEU). Other radionuclides encountered in relatively minor amounts in certain areas of the clean-up included Co-60, Cs-137, Ra-226, Th-232 and low enriched uranium (LEU). Between 2009 and the spring of 2011, remediation efforts were focused on demolition of contaminated buildings and removal of contaminated soil. In the late spring of 2011, the last phase of remediation commenced involving the removal of contaminated sediments from portions of a 1,200 meter long gaining stream. Planning and preparation for remediation of the stream began in 2009 with submittal of permit applications to undertake construction activities in a wetland area. The permitting process was lengthy and involved securing permits from multiple agencies. However, early and frequent communication with stakeholders played an integral role in efficiently obtaining the permit approvals. Frequent communication with stakeholders throughout the planning and remediation process also proved to be a key factor in timely completion of the project. The remediation of the stream involved the use of temporary bladder berms to divert surface water flow, water diversion piping, a sediment vacuum removal system, excavation of sediments using small front-end loaders, sediment dewatering, and waste packaging, transportation and disposal. Many safeguards were employed to protect several species of concern in the work area, water management during project activities, challenges encountered during the project, methods of Final Status Survey, and stream restoration.

Decommissioning of nuclear power plants generates large volumes of radioactive or potentially radioactive waste. The proper management of the dismantling waste plays an important role for the time needed for the dismantling phase and thus is critical to the decommissioning cost. An efficient and thorough process for inventorying, characterization and categorization of the waste provides a sound basis for the planning process. As part of comprehensive decommissioning studies for Nordic NPPs, Westinghouse has developed the decommissioning inventories that have been used for estimations of the duration of specific work packages and the corresponding costs. As part of creating the design basis for a national repository for decommissioning waste, the total production of different categories of waste packages has also been predicted. Studsvik has developed a risk based concept for categorization and handling of the generated waste using six different categories with a span from extremely small risk for radiological contamination to high level waste. The two companies have recently joined their skills in the area of decommissioning on selected market in a consortium named ndcon to further strengthen the proposed process. Depending on the risk for radiological contamination or the radiological properties and other properties of importance for waste management, treatment routes are proposed with well-defined and proven methods for on-site or off-site treatment, activity determination and conditioning. The system is based on a graded approach philosophy aiming for high confidence and sustainability, aiming for re-use and recycling where found applicable. The objective is to establish a process where all dismantled material has a pre-determined treatment route. These routes should through measurements, categorization, treatment, conditioning, intermediate storage and final disposal be designed to provide a steady, un-disturbed flow of material to avoid interruptions. Bottle-necks in the process causes increased space requirements and will have negative impact on the project schedule, which increases not only the cost but also the dose exposure to personnel. For these reasons it is critical to create a process that transfers material into conditioned waste ready for disposal as quickly as possible. To a certain extent the decommissioning program should be led by the waste management process. With the objective to reduce time for handling of dismantled material at site and to efficiently and environmental-friendly use waste management methods (clearance for re-use followed by clearance for recycling), the costs for the plant decommissioning could be reduced as well as time needed for performing the decommissioning project. Also, risks for delays would be reduced with a well-defined handling scheme which limits surprises. Delays are a major cost driver for decommissioning projects.

The remediation and decommissioning of the Hematite Former Fuel Cycle Facility (FFCF), the Hematite Facility, is currently being carried out by Westinghouse Electric Company LLC under the Hematite Decommissioning Project (HDP). The Hematite Facility is located near the town of Hematite, Missouri, USA. The Hematite Facility consists of 228 acres of land with primary operations historically being conducted within the central portion of the property that is roughly 10 acres including Burial Pits and the Site Pond area. Decommissioning and remediation activities are being performed with the eventual objective of the release of the property. Primary contaminants include the legacy disposal and contamination of natural and enriched uranium from the nuclear fuel cycle, as well as chemicals used during the facility operations. Two major regulatory bodies, the U.S. Nuclear Regulatory Commission (NRC) and the Missouri Department of Natural Resources (MDNR), provide critical roles in the approval and oversight of the current regulatory path to remediation, decommissioning and eventual release. Further, remediation and decommissioning activities are performed under the implementing policies, plans, and procedures under the Hematite Decommissioning Plan (DP) and the Record of Decision (ROD). Remediation and decommissioning tasks at the Hematite Former Fuel Cycle Facility, referred to as the Hematite Facility, are performed against a disciplined decision logic flow that applies accumulated technical and monitoring data to determine each step of the excavation, exhumation, and removal of wastes from the Burial Pits and the remaining Areas of Concern (AOC). Decision flow logic is based upon the nuclear criticality safety controls and threshold conditions, relative level of radioactive and chemical contamination, security protocol, and final waste stream disposition. The end result is to remediate the residual radioactive and chemical contamination to approved dose-based and risk-based cleanup criteria as negotiated with U.S. Federal and State Regulators. The purpose of the paper is to provide a summary of the successful implementation of the decision flow logic to the remediation and decommissioning tasks performed to date.

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.

The U.S. Department of Energy is responsible for risk reduction and cleanup of its nuclear weapons complex. Remediation strategies for some of the contamination may include techniques that mitigate risk, but leave contaminants in place. Monitoring to verify remedy performance and long-term mitigation of risk is key to implementing these strategies and can be a large portion of the total cost of remedy implementation. Especially in these situations, there is a need for innovative monitoring approaches that move away from the cost- and labor-intensive point-source monitoring. In this paper, alternative approaches for monitoring are presented for vadose zone, groundwater, groundwater/surface water interface, and surface water. To illustrate integrated, systems-based monitoring, this paper focuses on vadose zone contaminant remediation to mitigate impact to groundwater. In this context, vadose zone contamination is a source, or potential source, to groundwater plumes. The monitoring design uses a systems-based approach focused on developing a conceptual site model that highlights key features that control contaminant flux to groundwater. These features are derived considering the unsaturated flow and contaminant transport processes in the vadose zone and the nature of the waste discharge. Diagnostic properties and/or parameters related to both short- and long-term contaminant flux to groundwater can be identified and targeted for monitoring. The resolution of monitoring data needed to correspond to a functionally useful indicator of flux to groundwater can be estimated using quantitative analyses and the associated unsaturated flow properties relevant to the targeted site and vadose zone features. This monitoring design approach follows the process of developing a quantitative conceptual model suitable for supporting projections of future flux to groundwater. Support for such projections is important because it is likely that, in many cases, remediation decisions for the vadose zone will need to be made based all or in part on projected impacts to groundwater, and monitoring will then be applied to verify that remedy goals are being met.

Preparation for the decommissioning of the Ignalina Nuclear Power Plant involves multiple problems. Personnel radiation safety during the performance of dismantling activities is one of them. In order to assess the optimal personnel radiation safety, the modelling is performed for large components by the means of computer code “VISIPLAN 3D ALARA Planning tool” developed by SCK CEN (Belgium). Modelling results of radiation doses during the dismantling of the pressurized tank from the emergency core cooling system (ECCS PT) of RBMK-1500 reactor are presented in this paper. The mass of one ECCS PT is approximately 47.6 tons. The radiological surveys indicate that the inner surface of the ECCS PT is contaminated with radioactive products of corrosion and sediments due to the radioactive water. The assessment of workers exposure was performed to comply with ALARA. The effective doses to the workers were modeled for different strategies of ECCS PT dismantling. The impact of dismantling tools and shielding types and extract ventilation flow rate during the dismantling of ECCS PT on effective doses were analyzed. The total effective personnel doses were obtained by summarizing the effective personnel doses from various sources of exposure, i. e., direct radiation from radioactive equipment, internal radiation due to inhalation of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting in premises.

The site descriptive model covering the current status of characteristics of geological environment and the site evolution model for estimation of the long-term evolution of site conditions are used to integrate multi-disciplinary investigation results. It is important to evaluate uncertainties in the models, to specify issues regarding the uncertainties and to prioritize the resolution of specified issues, for the planning of site characterization. There is a large quantity of technical know-how in the modeling process. It is important to record the technical know-how with transparency and traceability, since site characterization projects generally need long duration. The transfer of the technical know-how accumulated in the research and development (R&D) phase to the implementation phase is equally important. The aim of this study is to support the planning of initial surface-based site characterizations based on the technical know-how accumulated from the underground research laboratory projects. These projects are broad scientific studies of the deep geological environment and provide a technical basis for the geological disposal of high-level radioactive wastes. In this study, a comprehensive task flow from acquisition of existing data to planning of field investigations through the modeling has been specified. Specific task flow and decision-making process to perform the tasks have been specified.

The French Act voted in 2006 made the choice of deep geological disposal as the reference option for the long term management of high level (HLW) and intermediate level long-lived waste. The Cige´o repository project aims at avoiding or limiting burden to future generations, which could not be achieved by the extension in time of interim storage. The reversibility as provided by the Act will maintain a liberty of choice for waste management on a duration which is comparable to new storage facility. Interim storage is required to accommodate waste as long as the repository is not available. The commissioning of the repository in 2025 will not suppress needs for interim storage. The paper describes the complementarities between existing and future interim storage facilities and the repository project: repository operational issues and planning, HLW thermal decay, support for the reversibility, etc. It shows opportunities to prepare a global optimization of waste management including the utilization at best of storage capacities and the planning of waste emplacement in the repository in such a way to facilitate operational conditions and to limit cost. Preliminary simulations of storage-disposal scenarios are presented. Thanks to an optimal use of the waste management system, provision can be made for a progressive increase of waste emplacement flow during the first operation phase of the repository. It is then possible to stabilize the industrial activity level of the repository site. An optimal utilization of interim storage can also limit the diversity of waste packages emplaced simultaneously, which facilitates the operation of the repository. 60 years minimum interim storage duration is generally required with respect to HLW thermal output. Extending this interim storage period may reduce the underground footprint of the repository. Regarding reversibility, the capability to manage waste packages potentially retrieved from the repository should be analyzed. The 2006 French Act provides for a research program on interim storage to be carried out along with the repository project development. This program has been guided by the complementarities between interim storage and the repository project. The main research issues address the longevity of storage facilities, up to 100 years, their versatility with regard to waste package types and their modularity to match future needs progressively. In parallel the dialogue between Andra and waste producers will continue to propose optimized waste management scenarios.

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.

Geological hazard assessments are being used to make important decisions relevant to nuclear facilities such as a repository for deep geological disposal of high-level radioactive waste. With respect to such repositories, topographic evolution is a key issue for description of the long-term evolution of a groundwater flow characteristics in time spans of tens to hundreds of thousands of years. The construction of topographic evolution models is complex, involving tacit knowledge and working processes. Therefore, it is important to externalise, that is to explicitly present the tacit knowledge and decision-making processes used by experts in the model building unambiguously, with thorough documentation and to provide key knowledge to support planning and implementation of investigations. In this study, documentation of the technical know-how used for the construction of a topographic evolution model is demonstrated. The process followed in the construction of the model is illustrated using task-flow logic diagrams; the process involves four main tasks with several subtasks. The task-flow followed for an investigation to estimate uplift rates linked to the task-flow for the modelling of topographic evolution is also illustrated. In addition, the decision-making processes in the investigation are expressed in logical IF-THEN format for each task. Based on the documented technical know-how, an IT-based Expert System was constructed. In future work, it is necessary to analyse the knowledge, including the management of uncertainties in the modelling and investigations, and to integrate fundamental ideas for managing uncertainties with expert system.

Svensk Ka¨rnbra¨nslehantering AB (SKB) has performed comprehensive investigations of two candidate sites for a final repository for Sweden’s spent nuclear fuel. In March 2011 SKB decided to submit licence applications for a final repository at Forsmark. Before selection, SKB stated that the site that offers the best prospects for achieving long-term safety in practice would be selected. Based on experiences previous safety assessments, a number of issues related to long-term safety need to be considered in the context of site comparison. The factors include sensitivity to climate change such as periods of permafrost and glaciations, rock mechanics evolution including the potential for thermally induced spalling and sensitivity to potential future earthquakes, current and future groundwater flow, evolution of groundwater composition and proximity to mineral resources. Each of these factors related to long-term safety for the two candidate sites is assessed in a comparative analysis of site characteristics. The assessment also considers differences in biosphere conditions and in the confidence of the site descriptions. The comparison is concluded by an assessment on how the identified differences would affect the estimated radiological risk from a repository located at either of the sites. The assessment concludes that there are a number of safety related site characteristics for which the analyses do not show any decisive differences in terms of implications on safety, between the sites Forsmark and Laxemar. However, the frequency of water conducting fractures at repository depth is much smaller at Forsmark than at Laxemar. This difference, in turn, affects the future stability of the current favourable groundwater composition, which combined with the much higher flows at Laxemar would, for the current repository design, lead to a breach in the safety functions for the buffer and the canister for many more deposition positions at Laxemar than at Forsmark. Thereby the calculated risk for Forsmark will be considerably lower than that for Laxemar. What decided the choice is that Forsmark is thus mainly that it was judged to offer better prospects for achieving long-term safety in the final repository. Other factors considered included implications regarding repository construction and operation, difference in the footprint of the repository, comparisons regarding environment and health as well as social resources and local support.

The UK’s Low Level Waste Repository Ltd submitted an Environmental Safety Case (ESC) for the disposal of low-level waste to the Environment Agency on the 1 st of May 2011. The ESC is a major submission that will decide the future use of the Repository and has major implications for the success of the UK’s LLW Strategy and decommissioning programme. This paper provides an overview of the work that has been carried out to support the submission. Key aspects of this ESC include: • detailed investigations of existing disposals, based on careful examination of existing records and other investigations, including interviews with former operational staff; • analysis of uncertainties in future disposals; • modelling of the biogeochemical evolution of the disposal system, which provides understanding of the evolution of pH, Eh and gas generation and thence underpinning for radionuclide releases in groundwater and gas; • development of a 3-D groundwater flow model, calibrated against observed heads and with a detailed representation of the engineered features; • analysis of coastal erosion and its impacts; • a major focus on optimisation based on detailed technical studies; • a conclusion that existing disposals do not require remediation; • the choice of a concrete vault design with permeable side walls designed to avoid bathtubbing after the end of management control; • a comprehensive set of assessment calculations, including thorough analysis of uncertainties, which demonstrate consistency with the Environment Agency’s risk and dose guidance levels; • revision of the LLWR’s WAC, based in part on the use of the ‘sum of fractions’ approach; • the use of a safety case document structure that emphasises key safety arguments in a Level 1 document and provides supporting evidence in a series of Level 2 documents; • the provision of a Level 2 document that describes in detail how each aspect of the regulatory guidance has been addressed. In the future, the 2011 ESC will be maintained using a formal system of change control. It will be used as a tool for decision making concerning the future development of the LLWR and waste acceptance.

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

Since 1979, Uranium enrichment technology has been researched through the gas centrifuge method, at Ningyo-toge Environmental Engineering Center of Japan Atomic Energy Agency (JAEA). In addition, the Demonstration Plant, that is final stage test facilities, was operating continuously from 1988 to 2001. As a result, a lot of residues accumulated in the plant. Most of this accumulation was found be uranium intermediate fluoride. The basic decommission policy of JAEA is that equipments of gas centrifuge will be decontaminated by sulfuric acid immersion method for clearance and reuse. In our plan, approximately 90% of metals will be cleared and reused, and then the remaining 10% will be disposed of radioactive waste. We propose a combination of sulfuric acid immersion method and the systematic chemical decontamination as an efficient method for decontamination of uranium enrichment facilities. This paper focuses on the method and performance of systematic chemical decontamination using IF 7 gas. The following (Figure 1) shows our decommission policy and position of systematic chemical decontamination by IF 7 gas for uranium enrichment plant. The IF 7 treatment technique belongs to the systematic decontamination technology. It has the high performance decontamination technique for the plant that accumulates the uranium intermediate fluoride, such as UF 4 , UF 5 , U 2 F 9 , and U 4 F 17 , which exist in the uranium enrichment plant through the Gas Centrifuge, called GCF. The one of characteristics of the IF 7 treatment, the secondary waste is just an IF 5 and little residues. In addition, this IF 5 can be reused as materials for making new IF 7 gas. The IF 7 treatment can also be performed in the room temperature and very low pressure like a 10–45hPa. Furthermore, the IF 7 treatment is a simple method using chemical reaction. For this reason, we hardly need to care about secondary reaction with the exception of the reaction with IF 7 gas and the uranium intermediate fluoride. This is a very important feature when applying to a large-scale plant. In order to carry out the IF 7 treatment, we only set up a few equipments in GCF uranium enrichment plant, which were IF 7 feeding equipment and two circulating pumps. IF 7 gas cylinders are seated in IF 7 feeding equipment. This is the only equipment. Figure 2 shows the IF 7 treatment system. We carried out the IF 7 treatment for the four cascades in the uranium enrichment Demonstration Plant. The weights of uranium residue in the cascades were approximately 700kgU per cascade prior to the IF 7 treatment. In the IF 7 treatment, we were able to find the near-optimal processing condition. As a result, we could confirm the IF 7 treatment period for one cascade which was 60 days. The main factor to determine the IF 7 treatment period is the pressure and the flow rate of reaction product gas (UF 6 and IF 5 mixture gas) exhausted from the cascade. Although we carried out the IF 7 treatment with the maximum value of the flow rate, which our facility has, it is possible to further shorten the IF 7 treatment period by setting a higher gas flow rate. Moreover, after the IF 7 treatment, we evaluated the uranium recovery rate for cascades and the residues’ uranium weight in the main equipment of GCF. In addition, in the evaluation of the uranium recovery rate, we enable to confirm the uranium recovery rate of all cascades achieved more tan 98%. Furthermore, the average of uranium recovery rate more than 99% in the cascade that has been processed at the end. As a result, radioactive concentration of uranium in the main parts of the GCF fell to 1.0B q/g and below.

The Low Level Waste Repository (LLWR) is the UK’s principal facility for the disposal of solid low-level radioactive waste and is operated by LLW Repository Limited. Presently, LLWR Ltd is establishing the long-term environmental safety of disposals of solid radioactive waste at the LLWR, through the submission of the 2011 Environmental Safety Case for the LLWR. This Environmental Safety Case addresses the Environment Agency Guidance on Requirements for Authorisation. Aspects of the submission consider improved vault design, closure design, and quantitative assessments. Each of these issues requires an understanding of the movement of water through the facility and the surrounding geology during operations and following facility closure. Groundwater flow modelling has been used extensively in support of the interpretation of field investigations, the development of the engineering design, and an assessment of the groundwater pathway as one of the major pathways by which contaminants may reach the environment. This paper describes these important aspects of the Environmental Safety Case. The geological environment in the region of the LLWR consists of Quaternary age deposits overlying older bedrock. The facility involves shallow excavations into the Quaternary deposits, originally for trenches, with disposals to a vault system beginning in 1988. In the post-closure phase these disposals are covered by a cap and surrounded by a cut-off wall to minimise the water flow around or through the waste. An innovative modelling methodology has been developed to represent the range of scales that have to be considered from the regional groundwater flow patterns over several kilometres, the scale of tens of metres around the immediate site area, and down to about 1 metre for details of flows within the repository itself in three dimensions. Detailed finite-element models of the flow through geological media and the engineered features are used to interpret site data and assess a credible set of post-closure situations and model cases. In the radiological assessment, a more simplified compartment model is used to assess uncertainties in hydrogeological properties and the long-term evolution of the engineered barriers. Together the approach provides flexible tools for understanding and assessing a comprehensive range of aspects including details of flows within the repository, dilution and migration in the external geology, the long-term evolution of the hydrogeological system, the implications of spatial variability and alternative geological models, and effects of uncertainties.