Migration Rates of Brine Inclusions in Single Crystals of Nacl

1981 ◽  
Vol 6 ◽  
Author(s):  
I-Ming Chou
Keyword(s):  
Radioactive Waste ◽  
Temperature Gradients ◽  
Spent Fuel ◽  
Ion Diffusion ◽  
Cold Side ◽  
Migration Rates ◽  
Permanent Storage ◽  
Salt Brine ◽  
Salt Deposits ◽  
High Level

Rock-salt deposits have been considered as a possible medium for the permanent storage of high-level radioactive wastes and spent fuel. Brine inclusions present in natural salt can migrate toward the waste if the temperature and the temperature gradients in the vicinity of the radioactive waste are large enough. This migration is due to the dissolution of salt at the hot side of the salt-brine interface, ion diffusion through the brine droplet, and the precipitation of salt at the cold side of the salt brine interface.

European Concepts for Shared Storage and Disposal Facilities for Radioactive Wastes?

2007 ◽  
Author(s):  
Ewoud Verhoef ◽  
Hans Code´e ◽  
Vladan Sˇtefula ◽  
Charles McCombie
Keyword(s):  
Radioactive Waste ◽  
Implementation Strategies ◽  
Practical Implementation ◽  
Spent Fuel ◽  
Pilot Studies ◽  
Eu Member States ◽  
Waste Storage ◽  
Geological Conditions ◽  
European Development ◽  
High Level

Geological disposal is an essential component of the long-term management of spent fuel and high-level radioactive waste. Implementation of a suitable deep repository may, however, be difficult or impossible in some (especially small) countries because of challenging geological conditions or restricted siting options, or because of the high costs involved. For these countries, shared regional or international storage and disposal facilities are a necessity. The European Parliament and the EC have both expressed support for concepts that could lead to regional shared facilities being implemented in the EU. The EC, therefore, funded two projects that form the first two steps of a staged process towards the implementation of shared regional or international storage and disposal facilities. In the period 2003 to 2005, the EC funded SAPIERR I, a project devoted to pilot studies on the feasibility of shared regional storage facilities and geological repositories, for use by European countries. The studies showed that shared regional repositories are feasible, but also that, if they are to be implemented, even some decades ahead, efforts must already be increased now. The first step would be to establish a structured framework for the work on regional repositories. This is the goal of SAPIERR II (2006–2008): to develop possible practical implementation strategies and organisational structures. These will enable a formalised, structured European Development Organisation (EDO) to be established in 2008 or afterwards for working on shared EU radioactive waste storage and disposal activities. The EDO can work in parallel with national waste programmes. Participating EU Member States will be able to use the structures developed as, when and if needed for the furtherance of their individual national policies.

A Review of Corrosion Considerations of Container Materials Relevant to Underground Disposal of High-Level Radioactive Waste in Belgium

2006 ◽  
Vol 932 ◽  
Author(s):  
Bruno Kursten ◽  
Frank Druyts
Keyword(s):  
Radioactive Waste ◽  
Large Scale ◽  
General Corrosion ◽  
Spent Fuel ◽  
High Level Radioactive Waste ◽  
Boom Clay ◽  
Nuclear Site ◽  
In Situ Experiments ◽  
Underground Disposal ◽  
High Level

ABSTRACTThe underground formation that is currently being considered in Belgium for the permanent disposal of high-level radioactive waste and spent fuel is a 30-million-year-old argillaceous sediment (Boom Clay layer). This layer is located in the northeast of Belgium and extending under the Mol-Dessel nuclear site at a depth between 180 and 280 meter.Within the concept for geological disposal (multibarrier system), the metallic container is the primary engineered barrier. Its main goal is to contain the radioactive waste and to prevent the groundwater from coming into contact with the wasteform by acting as a tight barrier. The corrosion resistance of container materials is an important aspect in ensuring the tightness of the metallic container and therefore plays an important role in the safe disposal of HLW. The metallic container has to provide a high integrity, i.e. no through-the-wall corrosion should occur, at least for the duration of the thermal phase (500 years for vitrified HLW and 2000 years for spent fuel).An extensive corrosion evaluation programme, sponsored by the national authorities and the European Commission, was started in Belgium in the mid 1980's. The main objective was to evaluate the long-term corrosion performance of a broad range of candidate container materials. In addition, the influence of several parameters, such as temperature, oxygen content, groundwater composition (chloride, sulphate and thiosulphate), γ-radiation, … were investigated. The experimental approach consisted of in situ experiments (performed in the underground research facility, HADES), electrochemical experiments, immersion experiments and large scale demonstration tests (OPHELIE, PRACLAY). Degradation modes considered included general corrosion, localised corrosion (pitting) and stress corrosion cracking.This paper gives an overview of the more relevant experimental results, gathered over the past 25 years, of the Belgian programme in the field of container corrosion.

A Computational Model for Estimating the Cost of Spent Fuel and High-Level Radioactive Waste Transport

2003 ◽  
Author(s):  
E. R. Johnson ◽  
R. E. Best
Keyword(s):  
Radioactive Waste ◽  
Spent Nuclear Fuel ◽  
Cost Model ◽  
Transportation Cost ◽  
Cost Calculation ◽  
Spent Fuel ◽  
Transport Costs ◽  
High Level Radioactive Waste ◽  
High Level ◽  
The Cost

JAI has developed a simple computer program for use in determining a preliminary estimate of costs for transporting spent nuclear fuel or high-level radioactive waste by legal weight truck or by rail. The JAI Corporation Spent Fuel and High-Level Radioactive Waste Transportation Cost Model © is a Microsoft Excel 2000-based collection of spreadsheets. Both the truck and rail sub-models consist of three spreadsheets, or modules — as follows: • The “Input” spreadsheet accepts the user’s inputs (the user’s configuration of the transportation scenario to be modeled); • The “Cost Calculations” spreadsheet lists cost components and associated calculations; • The “Results” spreadsheet summarized the calculated transportation costs. The program does not calculate costs between two specific points, but rather over a specific distance. The individual inputs required can be entered by the user — or the user can accept the default values built into the program. The input to the program is divided into the following elements: 1. Scenario configuration; 2. Financial assumptions; 3. Capital-related costs; 4. Operating costs; 5. Freight-related costs; 6. Security-related costs. The rail portion of the program also permits the calculation of the cost of heavy haul and barge transport. The cost calculation spreadsheet contains all the algorithms used for calculating each element of cost and summing them — and the results spreadsheet shows the separate cost of capital, operations, freight, security and miscellaneous costs, plus the total cost for the shipment(s). The program offers an easy way for obtaining preliminary estimates of the cost of transporting spent fuel or high-level radioactive waste, and a way to quickly estimate the sensitivity of transport costs to changes in conditions or shipping scenarios.

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.

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.

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