bluuebunny/crossref_metadata_2025 huggingface dataset:
Regulations for the Disposal of Radioactive Waste in the Konrad Repository
Authors: Jung Hagen G. , Bandt Gabriele
Published: August 2012

In Germany low / medium level waste, which is classified here as radioactive waste with negligible heat generation, will be disposed of in the Konrad underground repository. The construction and the operation of this nuclear facility required authorization by different fields of law, i.e., by nuclear law, mining law and water law. Whereas the nuclear law considers solely radiological aspects, the relevant permit issued according to the water law considers the impact of radioactive as well as non-radioactive harmful substances. The Federal Office for Radiation Protection (BfS) as operator of the repository and permit holder has (a) to record the disposed of radioactive and non-radioactive harmful substances and (b) to balance them. To meet these requirements BfS has developed a concept, which led to a site specific solution. Threshold values were defined for recording and for balancing the harmful substances. It had to be verified that by disposal of radioactive waste packages according to these values an adverse effect on the near-surface groundwater can be excluded. The Lower Saxony Water Management, Coastal Protection and Nature Conservation Agency (NLWKN) as the responsible water law regulatory authority approved the operator’s concept as appropriate to comply with the requirements of the Water Law Permit. Nonetheless, collateral clauses were imposed to assure this.
Status of the Development of In-Tank/At-Tank Separations Technologies for High-Level Waste Processing for the U.S. Department of Energy
Authors: Wilmarth William R. , Machara Nicholas P. , Peterson Reid A. , Bush Sheryl R.
Published: August 2012

Within the U.S. Department of Energy’s (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in “tank farms”). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are “first-of-a-kind” and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant reengineering to adapt to DOE’s specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford’s Waste Treatment and Immobilization Plant (WTP) or Savannah River’s Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R&D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the salt and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, wile treatment of posphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.
Remediation and Assessment of the National Radioactive Waste Storage and Disposal Site in Tajikistan
Authors: Buriev Nazirzhon T. , Abdushukurov Dzhamshed A. , Vandergraaf Tjalle T.
Published: August 2012

The National Radioactive Waste Storage and Disposal Site was established in 1959 in the Faizabad region approximately 50 km east of the capital, Dushanbe. The site is located on the southern flank of the Fan Mountains facing the Gissar Valley in a sparsely populated agricultural area, with the nearest villages located a few km from the site. The site was initially designed to accept a wide range of contaminated materials, including obsolete smoke detectors, sealed radioactive sources, waste from medical institutions, and radioactive liquids. Between 1962 and 1976, 363 tonnes and 1146 litres of material, contaminated with a range of radionuclides were shipped to the site. Between 1972–1980 and 1985–1991, ∼4.8 × 1014 and 2 × 1013 Bq, respectively, were shipped to the site. An additional 7 × 1014 Bq was shipped to the site in 1996. Partly as a result of the dissolution of the former Soviet Union, the disposal site had fallen into disrepair and currently presents both an environmental hazard and a potential for the proliferation of radionuclides that could potentially be used for illicit purposes. Remediation of the disposal site was started in 2005. New security fences were erected and a new superstructure over an in-ground storage site constructed. A central alarm monitoring and observation station has been constructed and is now operational. The geology, flora, and fauna of the region have been documented. Radiation surveys of the buildings and the storage and disposal sites have been carried out. Samples of soil, surface water and vegetation have been taken and analyzed by gamma spectrometry. Results show a slight extent of contamination of soils near the filling ports of the underground liquid storage container where a Cs-137 concentration of 2.3 × 104 Bq/kg was obtained. Similar values were obtained for Ra-226. Radiation fields of the in-ground storage site were generally <1 μSv/h with 8% exceeding this value. Neutron radiation levels at the same location were also low with a few readings exceeding 10 μSv/h. The volume of liquid waste is estimated to be ∼140 m3. Most of the activity appears to be associated with the sediments in the tank. Suggestions are presented for the immobilization of the liquid wastes and long-term monitoring of the site and the surrounding territory.
Discrimination Monitors for Various Kinds of Waste to Be Down Graded
Authors: Naito Susumu , Yamamoto Shuji , Izumi Mikio , Obata Masamichi , Yoshimura Yukio , Sakurai Jiro , Sakai Hitoshi
Published: August 2012

During operation and maintenance, or decommissioning of nuclear power plant, various kind of waste should be treated, and exposure control is also required. These wastes have a wide range of contamination, different composition of nuclides, and a different shape, so each measurement instrument would be optimized for its use especially for very low level radioactivity measurement. TOSHIBA provides appropriate equipment for any needs to discriminate the very low and non radioactive waste to save cost of waste disposal, based on our original and innovative technology. For alpha emitting nuclides, we are ready to supply instruments based on ionized particle measurement technology. For beta, gamma-emitting nuclides, we are ready to customize a shape of detector based on our original plastic scintillation material. Some examples are introduced.
Implications of Co-Existing National and Multinational Geological Repository Development Programmes in Europe
Authors: Chapman Neil , McCombie Charles , Verhoef Edwoud
Published: August 2012

The aim of this paper is to identify interactions between national and multi-national geological repository programmes that are potentially either beneficial or problematic, so as to avoid any unintended negative impacts on national or multinational programmes and to maximise mutual benefits wherever possible. Regional, multi-national cooperation on geological disposal is a topical issue at present at the IAEA and also in the EU, which has produced a new proposed Directive that explicitly allows agreements on sharing. Two extended tables in the paper identify and comment briefly on historic and current positive and negative perceptions of co-existing national and multinational geological disposal programmes. It would, we believe, be a positive message to the global nuclear community to know that leading national disposal programmes are interacting constructively with regional initiatives. The extensive experience already gained in national programmes could be very beneficial to new nuclear countries. In particular, it is crucial for all nuclear programmes that the new entrants develop credible, long-term waste management programmes, whether these be national, regional or dual-track. For global safety and security reasons, the nuclear community must move towards a situation where both national and shared programmes will co-exist and interact symbiotically.
Performance of a DrumScan® HRGS Solo Scanner for the Assay of Legacy Waste at the Belgoprocess Site
Authors: Parvin Daniel F. , Huys Thomas
Published: August 2012

On the sites of Belgoprocess several thousands of drums containing conditioned legacy waste are stored. A significant number of these waste packages are 220 litre drums containing radioactive waste embedded into inactive bitumen. Most of the radioactive waste in these drums was generated during the development and production of MOX-fuels and the operation of the Eurochemic reprocessing plant. The current state of a number of these packages is no longer acceptable for long term storage. In order to make the waste packages acceptable for interim storage a repackaging process was developed. The process involves the repackaging of the waste items into 400 or 700 litre waste drums and a non-destructive gamma-ray assay (NDA) measurement performed on the new package. The aim of the NDA measurement is to detect significant quantities of fissile material in order to demonstrate compliance with the operational limits of the storage building. Since the waste items are destined for geological disposal, there is no specific need for a detection limit in the order of milligrams of plutonium as required for surface disposal. To meet this NDA requirement Babcock International Group supplied, calibrated and commissioned an open geometry system from its HRGS product range. The DrumScan® HRGS Solo assay system was delivered to the Belgoprocess site in 2009 after completing a series of factory acceptance tests performed in the UK. In May 2009 after successful completion of the site acceptance tests performed in Belgium, the system has been undergoing extensive testing and validation by Belgoprocess in order to demonstrate acceptance and compliance to the Belgian Radioactive Waste Agency, NIRAS/ONDRAF. After a careful evaluation of the qualification file, NIRAS/ONDRAF approved the system for operational measurements at the end of 2010. This paper provides a detailed description of the NDA requirement, calibration methodology, system validation tests and overall measurement performance of the system.
Gamma-Ray Imaging for Generating 3D Dose Rate Maps
Authors: Hughes Karl , Cracknell Edmund
Published: August 2012

The use of gamma-ray imaging has become established in the nuclear industry, especially in the fields of decommissioning and clean-up, for identifying the origins of elevated gamma dose rates. Since their first use in the mid 1990s, gamma-ray imaging devices, such as the Babcock RadScan, have been used to produce colour overlay plots that indicate, in two dimensions, the locations and distribution of radioactive contaminants. This information is invaluable in helping project managers to plan clean-up and shielding activities, ensuring that all work carried out is cost effective and ALARP. Recent work undertaken by Babcock demonstrates the capability to generate 3D maps of dose rate fields from the output of gamma-ray imaging work. The combination of gamma-ray imaging survey and resultant dose map is a very powerful tool for planning decommissioning. The conventional gamma-ray image provides an unambiguous identification of the origins of the dose rates present whilst the 3D dose map allows the dose uptake to personnel to be determined. Furthermore, the ability to quantify the effect of clean-up or shielding on the dose rates is possible, providing project teams with a metric for determining the best option available. A simple procedure is followed to generate 3D dose maps from gamma-ray imaging data. Firstly a model of the plant area is constructed. This model can be generated from existing plant drawings, from laser scan surveys, or from simple physical measurements taken on plant. The model includes information about the shielding properties of the plant structures, and can be easily modified to demonstrate the effect of adding more shielding, or of reducing any of the source terms. The data from a gamma-ray imager, such as RadScan, are analysed to generate distributions of radionuclide specific activities. These activities are entered into the model and form the source term. The advantage of using a gamma ray imager to generate the source term is that the location and distribution of the source is accurately represented in the model, thus ensuring that accurate dose maps are generated. Once the model has been completed it is analysed using a radiation transport modelling code such as Attila to produce the 3D dose maps (although other codes are available which could perform the same function). This paper describes an example of how this technique has been used to generate 3D dose maps for a customer and builds on earlier work with RadScan which provides quantitative in-cell assay.
Impact of Nuclide Vector Composition Contained in Conditionally Released Steel Reused in Motorway Tunnels Scenario on Calculated Individual Effective Doses
Authors: Hrncˇi´rˇ Toma´sˇ , Necˇas Vladimi´r
Published: August 2012

This paper presents ongoing results of the project presented at ICEM 10 [paper 40071] related to the topics of the reuse of conditionally released materials arising from decommissioning of nuclear installations. The subject of the presented paper is modelling of motorway tunnels, which reuse the conditionally released steel in form of reinforcing nets and bars for the concrete construction of tunnels. The general approach for the project was presented at ICEM 10. The activities of the project continued in evaluating the impact of the nuclide composition contained in the conditionally released steel on calculated individual effective doses from the external exposure (the internal exposure will be evaluated in next stages of the project). Mentioned radioactive steel arises during decommissioning and would be reused in motorway tunnels. Evaluated scenarios are related to critical groups of public (driver’s scenarios) and for professionals constructing the tunnels and performing the maintenance of tunnels (worker’s scenarios). The computer code VISIPLAN 4.0 3D ALARA planning tool was used for the calculation of individual effective doses for worker and for public groups. Various limits of the individual effective dose are used for public and for professionals. The aim of the ongoing modelling is to develop a set of data of the maximal radioactivity concentration for individual radionuclides contained in the conditionally released steel used in tunnel model constructions in order not to exceed the limits for the individual effective dose.
Reuse of Conditionally Released Steel: Proposals and Evaluation of Processes for Manufacturing of Steel Elements and Processes for Construction of Selected Scenarios
Authors: Beza´k Peter , Ondra Frantisˇek , Hajkova´ Eva , Danisˇka Vladimir , Necˇas Vladimi´r
Published: August 2012

The project include systematic scenarios analysis of conditionally released materials from the decommissioning of nuclear installations and the creation of new knowledge in this field, which will be used for implementing projects for reuse of these materials. New knowledge includes data about materials from the decommissioning (types of materials and radiological data on the basis of analysis of various scenarios). Scenarios contain information about conditionally released materials, data of the external exposure of personnel who will assemble those structures and who will be use the constructions up to the target scenario. Scenarios assume guarantee that the final products will be placed on the current position for a very long period from 50 to 100 years. The paper presents the review of activities for manufacturing of various steel construction elements made of conditionally released steels and activities for realisation of selected scenarios for reuse of construction elements. The ingots after melting of decommissioned radioactive steel materials are as the starting material for manufacturing of steel components. Ingots from the controlled area will be melted into induction furnace and the mixture of liquid steel will be alloyed for achieve of required chemical parameters. Typical steel products suitable for established scenarios are steel rebar of concrete, steel profiles of various forms, railway rails and rolled steel sheets. Target scenarios include an analysis of staff exposure during installation of steel constructions as well as exposure of individual from critical groups of population during their exploitation.
Integrating History and Measurement Into a Case for Site Release
Authors: Bartlett Angela , Davies Mike , Burgess Peter , Coppins Gavin
Published: August 2012

The United Kingdom nuclear research programme started in the 1940s. Research Sites Restoration Limited (RSRL) is responsible for the restoration of two sites which were at the forefront of this research, under a programme funded by the UK Nuclear Decommissioning Authority (NDA). These are the 100 hectare Harwell site in Oxfordshire and the 84 hectare Winfrith site on the south coast of England. The work performed on these sites covered a huge range of nuclides, combinations of nuclides, chemical and physical processes, far more complicated than a power station, for example. The sites have a complex history with records of hundreds of buildings, many kilometres of drainage systems, groundwater contamination issues and land areas which require remediation. Formal work towards site release began in the 1990s, but demolition and clearance for re-use started many years earlier. An efficient restoration programme requires appropriate quality data. It is vital to decide what you need to know and how well you need to know it. As part of this, a challenging number of factors need to be considered in its design. This paper discusses these factors using the examples of the approach used at the Harwell and Winfrith sites including: • historical knowledge and associated uncertainties; • relevant clearance criteria; • availability and limitations of surveying equipment; • effective targeted and validation sampling with appropriate analytical methods; • data capture and analysis techniques; • effective communication between RSRL and the relevant technical teams; • mapping technologies (Global Positioning Systems, Geographical Information Systems); • use of Babcock’s IMAGES land quality software tool; • integration of the above over long time scales. The RSRL programme of works at the Harwell and Winfrith Sites is producing large volumes of different types of information from decommissioning, site investigation and remediation projects. This will be required to be accessible and understandable to support the process of site release which will continue over many years. The paper illustrates the methods by which RSRL is using effective knowledge management to compile a verifiable record to support site release as the site restoration works progress.
Influence of the Conditional Release of the Materials With Very Low Level of Radioactivity on the Environment
Authors: Prvakova Slavka , Mrskova Adela , Pritrsky Jozef
Published: August 2012

Significant amount of solid materials (metals, non-metals, building structures) that could be contaminated or activated is produced during operation and especially decommissioning of nuclear power plants. Considerable fraction of these materials has level of radioactivity close to the radiological limits allowing its safe release into the environment thereby could be potentially recycled within the special constructions, as for example tunnels, roads or bridges. If the requirements of legislation on the radiological limits for handling such material and long-term safety of the constructions are met, contaminated material can be incorporated in the form of recycled concrete, remelted steel, etc. The paper presents implementation of the IAEA and EC recommendations into the form of detailed analytical approaches with the aim to develop integrated scenarios and to analyse long-term influence of the conditional release of the material with very low level of radioactivity on the environment. Further, an estimation of the key input parameters characteristic for the specific conditions of Slovak case in order to fulfil the radiological limits according to the Slovak legislation is included. Analysed integrated scenarios represent surface or underground civil construction with radionuclides released directly into the geosphere and transported by a groundwater flow to the biosphere. The migration of radionuclides is modelled in the near-surface conditions with the advection as a dominant transport mechanism. Computer code GoldSim is used to evaluate the long-term safety assessment of the conditionally released material on the environment.
The Complex Challenge of Repairing the Gantry Steelwork on the First Generation Magnox Storage Pond at Sellafield: Legacy Waste Storage, First Generation Magnox Storage Pond
Authors: Richardson Ian E.
Published: August 2012

This paper puts into context the challenges that were faced when repairing the Gantry Steelwork of the First Generation Magnox Storage Pond (FGMSP). The First Generation Magnox Fuel Storage Pond (FGMSP) provided fuel storage and decanning capability from the early 1960’s until 1986. A significant programme of work has been underway since the completion of operational activities to support the programmes strategic intent of retrieving and storing all legacy wastes, and remediating the structure of the plant to support decommissioning activities. A key enabler to the retrievals programme is the Pond Skip Handler Machine (SHM), removed from service in 2002 following the discovery of significant signs of corrosion and distress, an inevitable consequence of being located in a coastal, salt laden environment. The SHM provides sole capability to access and retrieve the inventory of over 1000 fuel skips. It is also fundamental to future operations and the deployment of desludging equipment to recover significant bulk sludge’s from the pond floor. Failure of the SHM steelwork gantry at worst case could potentially result in the Skip Handler Machine being derailed. This has some potential to damage to the pond structure and at worst case may result in local radiological and environmental consequences. This paper will examine the challenges faced by the team as they successfully defined, planned and executed remedial work to a specific aspect of the civil structure, the SHM gantry rail system, using a purpose built refurbishment platform; the Gantry Refurbishment System. The paper will examine how an “innovative” approach was adopted to resolve the related issues of: • Refurbishing an aged structure to meet extended future operational demands. • The application of pragmatic engineering solutions against current codes and standards including seismic performance. • Provision of safe access for the workforce to undertake the refurbishment work against significant radiological and conventional safety constraints. • The use of off site test facilities to prove work methods. • Engagement of the multiple workforces including supply chain. • Development of challenging safety cases and management control arrangements to undertake the work. The paper will detail the arrangements established to engage all stakeholder groups aligned to a common goal, programme and end position, together with the arrangements put in place for managing the activities of delivery teams, operators and emergency response teams engaged in task execution over the five year period. Finally, the paper will also share the learning from the projects completion, so that the improvement opportunities flowing from this approach can be shared across the industry for the future benefit of all.
Decommissioning of the Active Storage Facility for Residuals in Rheinsberg (Brandenburg/Germany) and Results of Groundwater Monitoring
Authors: Steinhoff Mathias
Published: August 2012

The Active Storage Facility for Residuals (Aktives Lager fu¨r Ru¨cksta¨nde – ALfR) — an open-air controlled area — was used for the storage of radioactive waste (liquid and solid waste) at the site of the NPP Rheinsberg in the former GDR. Liquid waste was cemented and dumped for near surface final disposal. During the period of operation of the ALfR some incidents happened. Residues of evaporated cooling water of unknown amount (ca. 1–10 m3), the drop-out from small leakages and the wash-off from contaminated open-air storage barrels contaminated the soil. Since 1997 the buildings of the ALfR have been decommissioned and the site will be remediated. From the beginning of decommissioning up to now the groundwater has been monitored. This paper shows different steps of decommissioning and explains the results of groundwater monitoring. Currently 33 sampling points are being used for sampling groundwater from different layers. The radiological parameters under examination are Cobalt-60, Cesium-137, Strontium-90 and Tritium. Non-radiological parameters like Bore, Potassium, Sodium, Nitrate, pH value and TOC (total organic carbon) are also being investigated. Since 1997 the activity concentration of Cobalt-60 and Tritium in the groundwater has been decreasing. The activity concentration of Strontium-90 had increased from 1997 to 2007. Since 2007 activity concentration of Strontium-90 has been decreasing. 350 m away from the ALfR, in the direction of groundwater flow, the Lake Stechlin is located in a nature protection area. The groundwater beyond the ALfR-site is not used as drinking-water and up to now there has been no detectable contamination of Lake Stechlin. Therefore there is no radiological impact on humans by contaminated groundwater. The hypothetic effective dose of Cobalt-60, Tritium and Strontium-90 resulting from postulated consumption of groundwater was determined using consumption rates according to the German Radiation Protection Ordinance. In case of using the groundwater as drinking-water the annual effective dose is 0,4 mSv/a at a postulated drinking-water well on the ALfR-site and 2,5 μSv/a at a postulated drinking water well in a distance of 142 m. Results of the observations: The observations of direction and velocity of the flow of the ground-water principally confirms the transportation model made at the beginning of decommissioning of ALfR. In vertical direction, the real spread of contamination in the groundwater is smaller than assumed. The dilution of Co-60 is higher than assumed. In contrast to the assumptions of the transportation model, Cesium-137 has a lower mobility.
The LLWR’s 2011 Environmental Safety Case
Authors: Cummings Richard , Baker Andrew J. , Sumerling Trevor , Shevelan John , Huntington Amy
Published: August 2012

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 1st 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.
An Assessment of the Radiological Impact of Coastal Erosion of the UK Low-Level Waste Repository
Authors: Sumerling Trevor , Fish Paul , Towler George , Penfold James , Shevelan John , Cummings Richard
Published: August 2012

The UK Low Level Waste Repository Ltd submitted an Environmental Safety Case for the disposal of low-level waste to our regulator, the Environment Agency, on the 1st of May 2011. This includes assessments of the long-term radiological safety of past and future disposals. A particular feature of the Low Level Waste Repository (LLWR) is that, because of its proximity to the coast, the site is vulnerable to coastal erosion. Our present understanding is that the site will be eroded on a timescale of a few hundred to a few thousand years, with consequent disruption of the repository, and dispersal of the wastes. We have undertaken a programme of scientific research and monitoring to characterise the evolution and function of the current coastal system that provides a basis for forecasting its future evolution. This has included modelling of contemporary hydrodynamics, geomorphological mapping, repeat LiDAR and aerial photographic surveys to detect patterns and rates of change, coastal inspections and reconstructions of post-glacial (i.e. last 15,000 years) sea levels and sediment budgets. Estimates of future sea-level rise have been derived from international sources and consideration given to the impact of such on the local coastline. Two alternative models of coastal recession have then been applied, one empirical and one physical-process based, taking account of the composition of Quaternary-age sediments between the coast and the site and uncertainties in future local sea level change. Comparison of the ranges of calculated times to site contact with sea-level rise indicate that the repository is most likely to be disrupted by undercutting of the engineered vaults and of the trenches. A novel and flexible radiological assessment model has been developed to analyse the impacts of the erosion of the repository and subsequent dispersal of wastes. The model represents the spatial layout of the site and distribution of radionuclides within the repository and is able to take account of a range of uncertainties. These include uncertainties related to the rate of erosion through the facility, amounts of co-erosion of geological and cap materials, alternative assumptions for residence of waste materials on the beach, alternative waste form associations, the wider dispersion of the eroded materials and marine sorption/desorption processes. Results indicate assessed annual doses and risks that are consistent with regulatory guidance levels.
Methodology of External Exposure Calculation for Reuse of Conditional Released Materials From Decommissioning
Authors: Ondra Frantisˇek , Vasˇko Marek , Necˇas Vladimi´r
Published: August 2012

The article presents methodology of external exposure calculation for reuse of conditional released materials from decommissioning using VISIPLAN 3D ALARA planning tool. Production of rails has been used as an example application of proposed methodology within the CONRELMAT project.
Transport and Deposition Properties of Model Slurries of One and Two Particle Species
Authors: Rice Hugh , Biggs Simon , Fairweather Michael , Young James
Published: August 2012

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.
Instrumented Measurements on Radioactive Waste Disposal Containers During Experimental Drop Testing
Authors: Quercetti Thomas , Musolff Andre , Mu¨ller Karsten
Published: August 2012

In context with disposal container safety assessment of containers for radioactive waste the German Federal Institute for Materials Research and Testing (BAM) performed numerous drop tests in the last years. The tests were accompanied by extensive and various measurement techniques especially by instrumented measurements with strain gages and accelerometers. The instrumentation of a specimen is an important tool to evaluate its mechanical behavior during impact. Test results as deceleration-time and strain-time functions constitute a main basis for the validation of assumptions in the safety analysis and for the evaluation of calculations based on finite-element methods. Strain gauges are useful to determine the time dependent magnitude of any deformation and the associated stresses. Accelerometers are widely used for the measuring of motion i.e. speed or the displacement of the rigid cask body, vibration and shock events. In addition high-speed video technique can be used to visualize and analyze the kinematical impact scenario by motion analysis. The paper describes some selected aspects on instrumented measurements and motion analysis in context with low level radioactive waste (LLW) container drop testing.
Photocatalytic and Photochemical Degradation of Liquid Waste Containing EDTA
Authors: Lepeytre C. , Lavaud C. , Serve G.
Published: August 2012

The decontamination factor of liquid waste containing 60Co is generally weak. This is due to the presence of complexant molecules. For instance, complexation of EDTA with 60Co decreases efficiency of radioactive waste treatment. The aim of this study was to degrade EDTA in H2O and CO2 and to concentrate free 60Co in order to increase decontamination factor. A first test of radioactive waste treatment by photocatalysis was allowed to increase decontamination factor (60Co) from 16 to 196 with a device requiring to be improved. The present work concerns the first step of the degradation process development with a more powerful device. These first experiments were leaded to follow the only EDTA oxidation. EDTA degradation was carried out by the following Advanced Oxidation Processes (AOP): UV/H2O2 (photochemistry); UV/TiO2 (photocatalysis); UV/TiO2/H2O2. A specific reactor was achieved for this study. The wavelength used was 254 nm (UVC). The photocatalytic degradation of EDTA was carried out with Degussa P-25 titanium dioxide (TiO2), which is a semiconductor photocatalyst. The degradation degree of EDTA and the intermediate products were monitored by TOC and ionic chromatography methods. The effects of various parameters such as pH and the quantity of H2O2 were studied. This allows us to conclude that basic pH slows down EDTA degradation. The study showed that UV/H2O2 process was the most effective treatment process under acid conditions. The rate of EDTA degradation was very high and reached 95% in 120 minutes. The presence of glyoxilic, oxalic, glycolic and formic acids was detected as degradation products. Among the intermediates produced by photochemistry, NO3− ions presence informed of the amine degradations. These results highlighted faster EDTA degradation by photochemistry than photocatalysis.
Characterisation of the Geology of the UK Low Level Waste Repository
Authors: Shevelan John , Smith Nicholas T.
Published: August 2012

The UK Low Level Waste Repository Ltd (LLWR) submitted an Environmental Safety Case (ESC) for the disposal of low-level waste (LLW) to the UK Environment Agency on the 1st of May 2011. As part of the ESC, the LLWR have to demonstrate that a programme of site investigation and site characterisation has been carried out to provide the requisite information for the ESC and support facility design and construction. This paper explains the development of the site investigation programme and how the understanding of the geology of the site has developed. The geological environment in the region of the LLWR consists of Quaternary age deposits overlying older bedrock. The site has been subjected to a series of site investigation programmes from 1939 to the present day. The development of 3-D geological models was necessary to integrate data from boreholes, trial pits, geophysical investigations and beach exposures and data gained from site operations. The understanding of the geology has developed with each new set of data. Early simple interpretations from a few boreholes have been superseded by a series of more complex interpretations each incorporating the increasingly detailed observations. Initial attempts to develop a lithostratigraphic representation of the geology proved difficult. It was also difficult provide a clear link between the geology and the hydrogeology using a lithostratigraphic approach as required for the development of hydrogeological models. A lithofacies approach to characterising the geology was adopted in 2007, which has allowed the grouping of geological units with similar hydraulic properties and the development of a regional 3-D geological model. The 3-D geological model has been used as the framework for the development of a hydrogeological model for the site. The development of the 3-D geological models has been iterative. It was observed that there are differences between models developed using solely mathematical interpolation and those controlled by geological interpretation. The different representations of the geological information have been used to consider the effects of uncertainty in the geological interpretation in the hydrogeological modelling.
Reuse of Conditional Released Materials From Decommissioning: A Review of Approaches and Scenarios With Long-Term Constructions
Authors: Danisˇka Vladimir , Pri´trsky´ Jozef , Ondra Frantisˇek , Zachar Matej , Necˇas Vladimi´r
Published: August 2012

Paper presents the overall scope and actual results of the project for evaluation of representative scenarios for reuse of conditionally released materials from decommissioning. Aim of the project is to evaluate the possibilities of reuse of conditionally released steels and concrete in technical constructions which guarantee the long-term preservation of design properties over periods of 50–100 years. Interaction of conditionally released materials with public is limited and predictable due to design and purpose of selected constructions and due to fact that in many scenarios these materials are embedded in non-radioactive materials such as bars in reinforced concrete. Worker’s scenarios for preparation, operation and maintenance of these constructions are analysed in detail including the manufacturing of elements for these constructions. Project aims to evaluate the scenarios of reuse of conditionally released materials in a complex way in order to develop the data for designers of scenarios and to evaluate the volumes of conditionally released materials based on facility (to be decommissioned) inventory data. The long-term constructions considered are bridges, tunnels, roads, railway constructions, industrial buildings, power industry equipment and others. Evaluation covers following areas: • Analysis of activities for manufacturing of reinforcement bars, rolled steel sheets and other steel elements and analysis of activities for construction of evaluated scenarios in order to evaluate the external exposure of professionals performing these activities; • Analysis of external exposure of professionals involved in operation and maintenance of the long-term constructions; analysis of external exposure of public groups which are exposed to evaluated constructions; • Analysis of internal exposure of public groups from the radionuclides released from the evaluated scenarios based on models for migration of radionuclides from the long-term constructions to critical groups of public; • Based on evaluation of external and internal exposure both for public and workers, maximal concentration of individual radionuclides is defined for construction elements of evaluated scenarios, manufactured from conditionally released materials; • Evaluation of volumes of steels and concrete in the frame of a decommissioning project which fulfil the defined radioactivity concentration limits. Visiplan 4.0 3D ALARA software is used for evaluation of external exposure of professionals and public, GOLDSIM software for evaluation of internal exposure of public and OMEGA code for evaluation of volumes of conditionally releasable materials. Several other parallel papers proposed for ICEM 11 are presenting selected details of the project.
Factors Affecting the Longevity of Interim ILW Stores, Monitoring Methods and Techniques Supporting Store Life Extension
Authors: Morris Jenny , Wickham Stephen , Hicks Tim , Richardson Phil , McKinney James , Tearle Mark , Morris Darrell , Brownridge Melanie
Published: August 2012

Life-limiting features of stores (i.e. components of a typical Intermediate Level Waste (ILW) store that are vulnerable to failure) can be separated into those related to store design, the continuing operation of the store and equipment and store management. Environmental factors, including temperature, humidity, atmospheric pollutants and radiation fields also affect the longevity of ILW storage buildings and the integrity of the waste packages within the store. The current condition of the store and the causes and rates of any observed deterioration need to be determined to estimate the current operating life of the store and to analyse the costs and benefits associated with life extension. Suitable monitoring techniques and methods for extending the operating life of stores, either by remediation or replacement of identified life-limiting features, have also been assessed. Techniques including visual inspection (directly or by use of CCTV or inspection cells), corrosion coupons, crane monitoring and a wide range of indirect measurement methods can be used to monitor the condition of life-limiting features and the store environment. Methods for extending store life include the application of paint or coatings to retard degradation, use of electro-chemical treatments and repair of the concrete structure or replacement of components.
Radiological Impact of Co-Location of the VLLW and LILW Repository at Mochovce Site
Authors: Hanusik V. , Kusovska Z. , Moravek J. , Balaz J. , Chren O.
Published: August 2012

JAVYS, the Slovak Nuclear Decommissioning Company, the owner and operator of National Radioactive Waste Repository (NRR) at Mochovce, is planning the enlargement of the existing NRR. The enlargement consists of the construction of new structures (double rows) like the ones existing for Low and Intermediate Level Waste (LILW) and of a new facility for the disposal of Very Low Level Waste (VLLW). A VLLW disposal site is being planned in Mochovce where a LILW disposal site is already located. As a part of ongoing licensing process for this change in utilization of the NRR we have updated the already existing assessment of potential radiological impact to members of the public after closure of the site, both from the migration of leachate in groundwater from the site and from possible inadvertent intrusion into the site, including future residential development on material excavated for the construction of a road. The radionuclides from the groundwater reach the biosphere through a stream flowing into the lake. It is conservatively assumed that individual in the critical group uses biosphere of the lake (for irrigations, fishing and recreation). In the case of unintentional intrusion into the repository, the exposed group consists of a small number of workers who excavate or examine repository materials. The assessment is based on preliminary conceptual design for VLLW module, estimation of future volumes for the different waste classes and their radiological inventories and considers some common aspects and potential interactions between both disposal systems, such as radiological criteria, critical group definition, institutional period, site environment, total activity inventory.