Interactions between the co-located intermediate-level waste/low-level waste and high-level waste/spent fuel components of a geological disposal facility

2012 ◽  
Vol 76 (8) ◽  
pp. 3475-3482 ◽  
Author(s):  
T. W. Hicks ◽  
S. Watson ◽  
S. Norris ◽  
G. Towler ◽  
D. Reedha ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Three Dimensional ◽  
White Paper ◽  
Intermediate Level ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Disposal Facility ◽  
Three Dimensional Models ◽  
High Level

AbstractThe 2008 UK government White Paper, published as part of the Managing Radioactive Waste Safety programme, identified benefits to disposing of all of the UK's higher activity wastes at the same site. That is, a single geological disposal facility (GDF) could be constructed that consists of a module for low- and intermediate-level waste, and a module for high-level waste and spent fuel.A safety case for a co-located GDF will have to consider the extent to which evolving thermo-hydro-mechanical-chemical and gas (THMCG) conditions in and around one module may affect conditions in the other module, including the extent to which barrier performance and radionuclide migration behaviour could be altered. Several research projects have been undertaken on behalf of Radioactive Waste Management Directorate aimed at understanding and evaluating the THMCG interactions that might occur during the disposal facility operational and post-closure phases.This paper describes research on THMCG interactions between disposal modules based on illustrative GDF designs for different host rock environments. Interactions were evaluated using simple analytical solutions and detailed three-dimensional models. The analyses demonstrated that interactions can be controlled by design constraints.

Implementation of a Geological Disposal Facility (GDF) in the UK by the NDA Radioactive Waste Management Directorate (RWMD): The Potential for Interaction Between the Co-Located ILW/LLW and HLW/SF Components of a GDF

2009 ◽  
Author(s):  
George Towler ◽  
Tim Hicks ◽  
Sarah Watson ◽  
Simon Norris
Keyword(s):  
Radioactive Waste ◽  
White Paper ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Disposal Facility ◽  
Wide Range ◽  
The Uk ◽  
Host Rocks ◽  
System Designs

In June 2008 the UK government published a ‘White Paper’ as part of the “Managing Radioactive Waste Safety” (MRWS) programme to provide a framework for managing higher activity radioactive wastes in the long-term through geological disposal. The White Paper identifies that there are benefits to disposing all of the UK’s higher activity wastes (Low and Intermediate Level Waste (LLW and ILW), High Level Waste (HLW), Spent Fuel (SF), Uranium (U) and Plutonium (Pu)) at the same site, and this is currently the preferred option. It also notes that research will be required to support the detailed design and safety assessment in relation to any potentially detrimental interactions between the different modules. Different disposal system designs and associated Engineered Barrier Systems (EBS) will be required for these different waste types, i.e. ILW/LLW and HLW/SF. If declared as waste U would be disposed as ILW and Pu as HLW/SF. The Geological Disposal Facility (GDF) would therefore comprise two co-located modules (respectively for ILW/LLW and HLW/SF). This paper presents an overview of a study undertaken to assess the implications of co-location by identifying the key Thermo-Hydro-Mechanical-Chemical (THMC) interactions that might occur during both the operational and post-closure phases, and their consequences for GDF design, performance and safety. The MRWS programme is currently seeking expressions of interest from communities to host a GDF. Therefore, the study was required to consider a wide range of potential GDF host rocks and consistent, conceptual disposal system designs. Two example disposal concepts (i.e. combinations of host rock, GDF design including wasteform and layout, etc.) were carried forward for detailed assessment and a third for qualitative analysis. Dimensional and 1D analyses were used to identify the key interactions, and 3D models were used to investigate selected interactions in more detail. The results of this study show that it is possible for ILW/LLW and HLW/SF modules to be co-located without compromising key safety functions of different barrier components, and this reflects international precedents, e.g. the Andra and Nagra repository designs. There are two key technical issues that need to be managed in designing the geometry of the co-located GDF: (i) the heat flux from the HLW/SF module interacting with the ILW/LLW module, and (ii) the potential for development of an alkaline plume from the ILW/LLW module interacting with the HLW/SF module; particularly within fractured host rocks.

An overview of radionuclide behaviour research for the UK geological disposal programme

2012 ◽  
Vol 76 (8) ◽  
pp. 3373-3380 ◽  
Author(s):  
S. Vines ◽  
R. Beard
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Waste Disposal ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Nuclear Decommissioning ◽  
Disposal Facility ◽  
The Uk ◽  
High Level

AbstractIn the UK, radioactive wastes currently planned for disposal in a geological disposal facility (GDF) are intermediate-level waste, some low-level waste and high-level waste. Disposal of other materials, including spent fuel, separated uranium and separated plutonium are also included in the planning of a GDF, if such materials are classified as wastes in the future. This paper gives an overview of the radionuclide behaviour research studies of the Nuclear Decommissioning Authority Radioactive Waste Management Directorate (NDA RWMD). The NDA RWMD's current understanding of the processes that control radionuclide behaviour in groundwater and how the engineered and natural barriers in a GDF would contain radionuclides is presented. Areas requiring further work are also identified.

The Implementing Geological Disposal Technology Platform: Key Challenges in Research and Development in Radioactive Waste Management

2014 ◽  
Author(s):  
Jacques Delay ◽  
Jiri Slovak ◽  
Raymond Kowe
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Added Value ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Technology Platform ◽  
Joint Activities ◽  
The Individual ◽  
High Level

The Implementing Geological Disposal of Radioactive Waste Technology Platform (IGD-TP) was launched in November 2009 to tackle the remaining research, development and demonstration (RD&D) challenges with a view to fostering the implementation of geological disposal programmes for high-level and long-lived waste in Europe. The IGD-TP’s Vision is that “by 2025, the first geological disposal facilities for spent fuel, high-level waste and other long-lived radioactive waste will be operating safely in Europe”. Aside from most of European waste management organisations, the IGD-TP now has 110 members covering most of the RD&D actors in the field of implementing geological disposal of radioactive waste in Europe. The IGD-TP Strategic Research Agenda (SRA), that defines shared RD&D priorities with an important cooperative added value, is used as a basis for the Euratom programme. It provides a vehicle to emphasise RD&D and networking activities that are important for establishing safety cases and fostering disposal implementation. As the IGD-TP brings together the national organisations which have a mandate to implement geological disposal and act as science providers, its SRA also ensures a balance between fundamental science, implementation-driven RD&D and technological demonstration. The SRA is in turn supported by a Deployment Plan (DP) for the Joint Activities to be carried out by the Technology Platform with its members and participants. The Joint Activities were derived from the individual SRA Topics and prioritized and assigned a timeline for their implementation. The deployment scheme of the activities is updated on a yearly basis.

Safety Assessment of Geological Disposal of High-Level Radioactive Waste in Boom Clay: Relation With the Radionuclide Inventory

2009 ◽  
Author(s):  
Pierre Van Iseghem ◽  
Jan Marivoet
Keyword(s):  
Radioactive Waste ◽  
Safety Assessment ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
High Level Radioactive Waste ◽  
Boom Clay ◽  
High Level ◽  
Clay Formation ◽  
Parameter Values

This paper discusses the impact of the parameter values used for the transport of radionuclides from high-level radioactive waste to the far-field on the long-term safety of a proposed geological disposal in the Boom Clay formation in Belgium. The methodology of the Safety Assessment is explained, and the results of the Safety Assessment for vitrified high-level waste and spent fuel are presented. The radionuclides having the strongest impact on the dose-to-man for both HLW glass and spent fuel are 79Se, 129I, 126Sn, 36Cl, and 99Tc. Some of them are volatile during the vitrification process, other radionuclides are activation products, and for many of them there is no accurate information on their inventory in the waste form. The hypotheses in the selection of the main parameter values are further discussed, together with the status of the R&D on one of the main dose contributing radionuclides (79Se).

scholarly journals Preliminary modelling of radionuclide migration in the argillaceous sediments of the Sumer Formation (Northwestern Bulgaria)

2020 ◽  
Vol 49 (3) ◽  
pp. 13-18
Author(s):  
Dimitar Antonov ◽  
Madlena Tsvetkova ◽  
Doncho Karastanev
Keyword(s):  
Radioactive Waste Disposal ◽  
High Level Waste ◽  
Spent Fuel ◽  
Radionuclide Migration ◽  
Geological Disposal ◽  
Disposal Facility ◽  
Deep Geological Disposal ◽  
High Level ◽  
Host Rocks ◽  
Selection Of

In Bulgaria, from the preliminary analyses performed for site selection of deep geological disposal of high-level waste (HLW) and spent fuel (SF), it was concluded that the most promising host rocks are the argillaceous sediments of the Sumer Formation (Lower Cretaceous), situated in the Western Fore-Balkan Mts. The present paper aims to compare the transport of three major radionuclides from a hypothetical radioactive waste disposal facility, which incorporates an engineering barrier of bentonite into the argillaceous (marl) medium. The simulations were performed by using HYDRUS-1D computer programme. The results are used for a preliminary estimation of argillaceous sediments as a host rock for geological disposal of HLW.

scholarly journals The Implementing Geological Disposal Technology Platform – addressing the needs of new Member States

2015 ◽  
Vol 79 (6) ◽  
pp. 1591-1597 ◽  
Author(s):  
R. Kowe ◽  
J. Delay ◽  
M. Hammarström ◽  
T. Beattie ◽  
M. Palmu
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
High Level Waste ◽  
Spent Fuel ◽  
Member States ◽  
Geological Disposal ◽  
Technology Platform ◽  
Leading Role ◽  
High Level ◽  
European Union Member States

AbstractThe Implementing Geological Disposal of Radioactive Waste Technology Platform (IGD-TP) was launched in November 2009 to facilitate international cooperation in common areas of research, development and demonstration (RD&D) with a view to advancing the implementation of geological disposal facilities for spent fuel, high-level and other long-lived waste in Europe.The IGD-TP's Vision is that “by 2025, the first geological disposal facilities for spent fuel, high-level waste and other long-lived radioactive waste will be operating safely in Europe”. Aside from most European waste management organisations, the IGD-TP currently has 124 members covering most of the RD&D actors in the field of implementing geological disposal of radioactive waste in Europe.Five years after its inception, the IGD-TP has been shown to play a leading role in coordinating joint actions for RD&D in radioactive waste geological disposal programmes. The work of the platform takes into account differences between the timing and challenges for the respective waste management programmes. Following implementation of Posiva's geological disposal facility in Finland it is expected that within the next 5 years the construction of the Swedish and French geological disposal facilities will commence. Within IGD-TP, the SecIGD2 project whose remit is “Coordination and Support Action under the 7th Framework programme” aims at supporting, at the European level, the networking and structuring of RD&D programmes and competences in countries with less advanced geological disposal programmes, including those in the new European Union Member States. Furthermore, the SecIGD2 supports the development and coordination of the necessary competences to meet the Vision 2025 as a part of the platform's Competence Maintenance, Education and Training (CMET) working group.

Design Options for the UK’s ILW Geological Disposal Facility

2009 ◽  
Author(s):  
Tim Hicks ◽  
Matt White ◽  
Tamara Baldwin ◽  
Neil Chapman ◽  
Fiona Neall ◽  
...  
Keyword(s):  
Public Consultation ◽  
High Level Waste ◽  
Geological Environment ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Disposal Facility ◽  
Wide Range ◽  
The Uk ◽  
High Level ◽  
Design Options

Over the last few years, a major national programme of public consultation has been under way in the UK resulting, in 2006, in the announcement by government of geological disposal as the most appropriate solution for the long-term management of the UK’s long-lived and higher-activity radioactive waste and the launch, in 2008, of an implementation programme. The approach being pursued is to solicit volunteer communities to host a geological disposal facility, which may contain not only intermediate-level waste (ILW) and some low-level waste (LLW), but also high-level waste (HLW), any spent fuel (SF) declared as waste, and potentially other materials that may be declared as waste. These wastes have different physical, chemical, thermal and radiological characteristics, and different concepts will be required to accommodate their disposal, potentially in a single facility. The volunteer approach means that the geological environment that might eventually emerge as the preferred location is not known at the outset. Indeed, the siting process may require evaluation of several different geological environments because the UK has rich geological variability for such a small landmass. Consequently, the Nuclear Decommissioning Authority (NDA), which is charged with designing, developing and implementing a geological disposal facility, has investigated facility designs that could be appropriate for a wide range of host rocks and geological environments. This paper presents the results of a project carried out on behalf of the NDA to collate and report information on concepts for the geological disposal of ILW/LLW; a separate project carried out a parallel evaluation of options for disposing of HLW and SF. Initially, the range of geological disposal facility design options available worldwide for the disposal of ILW/LLW was evaluated. Nine disposal concepts were identified and reviewed that would cater for any geological environment likely to arise in the UK. These concepts have different engineering and operational aspects. The appropriateness of each option for implementation in five different generic geological environments was assessed using expert judgement, with input from the NDA, consultants and the UK regulatory agencies. The paper presents a set of generic designs derived from the study and discusses the key issues that would need to be addressed should any of these designs be considered for implementation in specific geological environments in the UK. The findings of this work are intended to provide a resource to support comparisons of alternative disposal concepts and the identification of designs suitable for the disposal of UK ILW/LLW in different geological environments.

Assessment of Redox Conditions in the Near Field of Nuclear Waste Repositories: Application to the Swiss high-level and intermediate level waste disposal concept

2003 ◽  
Vol 807 ◽  
Author(s):  
Paul Wersin ◽  
Lawrence H. Johnson ◽  
Bernhard Schwyn
Keyword(s):  
Solid Phase ◽  
Near Field ◽  
Intermediate Level ◽  
Redox Conditions ◽  
High Level Waste ◽  
Redox Potentials ◽  
Spent Fuel ◽  
Reducing Conditions ◽  
High Level ◽  
Waste Repositories

ABSTRACTRedox conditions were assessed for a spent fuel and high-level waste (SF/HLW) and an intermediate-level waste (ILW) repository. For both cases our analysis indicates permanently reducing conditions after a relatively short oxic period. The canister-bentonite near field in the HLW case displays a high redox buffering capacity because of expected high activity of dissolved and surface-bound Fe(II). This is contrary to the cementitious near field in the ILW case where concentrations of dissolved reduced species are low and redox reactions occur primarily via solid phase transformation processes.For the bentonite-canister near field, redox potentials of about -100 to -300 mV (SHE) are estimated, which is supported by recent kinetic data on U, Tc and Se interaction with reduced iron systems. For the cementitious near field, redox potentials of about -200 to -800 mV are estimated, which reflects the large uncertainties related to this alkaline environment.

Impact of Advanced Fuel Cycles on Geological Disposal

2009 ◽  
Vol 1193 ◽  
Author(s):  
Jan Marivoet ◽  
Eef Weetjens
Keyword(s):  
Radioactive Waste ◽  
Cooling Time ◽  
High Level Waste ◽  
Waste Streams ◽  
Geological Disposal ◽  
Fuel Cycles ◽  
Thermal Output ◽  
Advanced Fuel ◽  
High Level ◽  
The Impact

AbstractIn recent years the increasing oil prices and the need for carbon-free energy to limit global warming have resulted in a revival of interests in nuclear energy. Advanced nuclear fuel cycles are being studied worldwide. They aim at making more efficient use of the available resources, reducing the risk of proliferation of nuclear weapons, and facilitating the management of the resulting radioactive waste. Recently, the Red-Impact project has investigated the impact of a number of representative advanced fuel cycles on radioactive waste management, and more specific on geological disposal. The thermal output of the high-level waste arising from advanced fuel cycles in which all the actinides are recycled is reduced with a factor 3 for a 50 years cooling time and with a factor 5 for a 100 years cooling time in comparison with the spent fuel arising from the once-through fuel cycle. This reduction of the thermal output allows for a significant reduction of the length of the disposal galleries and of the size of the repository. Separation of Cs and Sr drastically reduces further the thermal output of the high-level waste, but it requires a long-term management of those heat generating separated waste streams, which contain the very long-lived 135Cs. Recycling all the actinides strongly reduces the radiotoxicity in the waste, resulting in significantly lower doses to an intruder in the case of a human intrusion into the repository. However, the reduction of radiotoxicity has little impact on the main safety indicator of a geological repository, i.e. the effective dose in the case of the expected evolution scenario; for disposal in clay formations, this dose is essentially due to mobile fission and activation products. The deployment of advanced fuel cycles will necessitate the development of low activation materials for the new nuclear facilities and fuels and of specific waste matrices to condition the high-level and medium-level waste streams that will arise from the advanced reprocessing plants.

Dry Storage of Spent Research Reactor Fuel in Castor BR3® Casks at Belgoprocess in Belgium

2001 ◽  
Author(s):  
Marnix Braeckeveldt ◽  
Luc Ooms ◽  
Gustaaf Geenen
Keyword(s):  
Radioactive Waste ◽  
Nuclear Research ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Dry Storage ◽  
Nuclear Site ◽  
Test Reactor ◽  
Intermediate Storage ◽  
High Level

Abstract The BR3 reactor (10.5 MWe) at the Nuclear Research Center SCK•CEN was the first PWR plant installed in Europe and has been shut down in 1987. The BR3 reactor is from 1989 in a decommissioning stage and most of the spent fuel is presently still stored in the deactivation pool of the BR3 plant and has to be evacuated. The BR3 was used as a test-reactor for new fuel types and assemblies (Mixed Oxide (MOX) fuel, fuel rods containing burnable poison (Gd2O3) and other types of fuels). Some fuel rods, having undergone a destructive analysis, are stored in different laboratories at the SCK•CEN. In total, the BR3 spent fuel comprises the equivalent of almost 200 fuel assemblies corresponding to some 5000 fuel rods. Beside the spent BR3 fuel, a limited number of spent fuel rods, with equivalent characteristics as the BR3 fuel but irradiated in research reactors outside Belgium and stored in other buildings at the SCK•CEN nuclear site, were added to the inventory of spent fuel to be evacuated. Various options such as reprocessing and intermediate storage awaiting final disposal were evaluated against criteria as available techniques, safety, waste production and overall costs. Finally the option of an AFR (away-from-reactor) intermediate dry storage of the BR3 and other spent fuel in seven CASTOR BR3® casks was adopted. As the SCK•CEN declared this spent fuel as radioactive waste, NIRAS/ONDRAF, the Belgian radioactive waste management agency became directly involved and the decision was taken to construct a small building at the Belgoprocess nuclear site for storing the CASTOR BR3® casks. Loading at the SCK•CEN followed by transport to Belgoprocess and storage is scheduled to take place at the end of 2001. The CASTOR BR3® cask weighing some 25 tonnes, consists of a monolithic body and has two independent lids with metal seals guaranteeing the long term leak-tightness of the cask. The CASTOR BR3® cask is designed for transport and the intermediate storage of at least 50 years. Although a defect of the leaktightness of a CASTOR BR3® cask is very unlikely to occur, an intervention scenario had to be developed. As no pool is present at the Belgoprocess nuclear site to unload the fuel, an innovative procedure is developed that consists of transferring the basket, containing the spent fuel, into another CASTOR BR3® cask. This operation can be performed in the hot cell of the existing storage building for high level waste at the Belgoprocess site.

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