Iron boxes for ILW transport and storage

Intermediate Level Waste has radioactivity levels exceeding the upper boundaries for Low Level Waste (LLW) but does not require heating to be taken into account in the design of storage or disposal facilities. ILW is generated from a number of activities such as decommissioning, spent fuel reprocessing, research facilities, reactor operation and/or historical waste storage practices.

The UK strategy for intermediate level waste is to retrieve, condition and package ILW, and keep the packages in storage facilities (for up to 150 years) until they can be emplaced in the UK Geological Disposal Facility (GDF), or managed over the long term in near-surface waste facilities in Scotland. Within the GDF environment, the waste package must be capable of limiting the release of contents in specified accident conditions that include exposure to a hydrocarbon fire and being dropped onto a hard unyielding surface in a worst impact attitude.

Various options have evolved over the last 30 years for packaging some types of ILW. The first generation of these packages is reinforced concrete boxes (6 cubic metre boxes) with immobilised contents; the waste package (container and wasteform) provide both shielding and containment under normal conditions of transport, and under accident conditions in the GDF. With concrete waste packages, and for some contents, there are technical difficulties to be overcome. For more mobile radionuclides, especially for caesium-137, there is a risk that the radionuclides could migrate through the concrete shielding to the surface of the container causing the surface contamination limits to exceed those specified for transport. Furthermore, there is a risk, if the environment is not controlled, of deterioration of the waste package from chloride-induced corrosion of the rebar; this could compromise the package performance over an extended storage period.

The second generation was designed for a wider range of contents. These were stainless steel boxes in two lengths, 2m and 4m, having concrete shielding lining the inside walls of the stainless steel container; again performance requirements were reliant on both container and waste form. To date, only prototypes of these packages have been made. The use of a stainless steel fabricated 'skin' would ensure that any migration of radionuclides would not lead to an increase in contamination levels on the surface of the package, however as the stainless steel affords little radiation shielding, dose rates may still rise.

In 2006 Magnox, working with the UK's Nuclear Decommissioning Agency, examined methodologies for accelerating its programme for decommissioning its fleet of Magnox reactors which had reached the end of life. Magnox examined a number of novel solutions for dealing with waste streams including overseas solutions. This concluded with Magnox introducing the concept of Robust Shielded Containers (RSC) for the long-term storage and eventual disposal of ILW. These RSCs were Ductile Cast Iron Containers (DCICs) which had been developed in Germany as ILW storage, transport and disposal containers by GNS (the package designer and design authority), Siempelkamp (the manufacturer) and BAM (the governmental technical authority). In these RSCs, the waste can be stored without requiring a shielded building and the waste does not need to be encapsulated (although it may require conditioning). The containers consisted of two types: a cuboidal DCIC (an IP-2 transport package; the subject of this paper) and a cylindrical DCIC (an IP-2 in its own right and Type B when fitted with impact and thermal limiters on the top and base). The use of RSCs supplied by GNS and other commercial companies to Magnox is expected to achieve substantial cost savings and programme acceleration.

The RSCs have several advantages compared to previous designs. The waste container itself meets the performance requirements for storage, transport and disposability. As the waste form is only required to provide a limited contribution to the package performance requirements, the need to encapsulate waste as an integral part of waste packaging operations is reduced. The load path for lifting and tie-down is an integral part of the container body of the DCICs; the body is made from a single complete casting and lifting and tie down features are machined into this casting. The corrosion of DCICs is predictable and allowances are built into some DCICs (for example, Croft Safstores) to allow for general corrosion losses. The container integrity is assessed by various NDT methods (for example magnetic particle or ultrasonics) and the same techniques can be used to examine DCICs during and after prolonged storage.

Advantages of DCICs

For a fixed external volume of waste package, as the shielding thickness is increased (for either concrete or ductile cast iron) the cavity is reduced. As DCI is a more efficient shield material than concrete, for the same shielding effect, the cavity of a DCI shielded container will be considerably larger than the equivalent concrete shielded container. The equivalent shielding of 150mm of concrete is 50mm of iron, based on comparison of relative intensities using respective attenuation coefficients (from radionuclides likely to dominate shielding requirements in decommissioning wastes, for example caesium-137 and cobalt-60 over the energy range for gamma rays of interest).

In considering the economics of packaging waste, the following factors should be taken into account:

• The total number of boxes required for the volume of waste under consideration, bearing in mind that for the same volume more concrete shielded boxes will be required compared to DCICs
• The resources, hence cost, required for processing, packaging and storing additional waste packages, bearing in mind that more concrete shielded boxes are likely to be needed compared to DCICs
• The cost of transporting additional waste packages
• The plant and equipment, hence cost, of a grout plant to infill the box, plus the lid-casting plant and operations. Although the same plant might be able to perform both, two separate operations may be required.
• The type of waste material that can be placed in concrete packages considering the technical issues that may need to be addressed to meet transport and disposability issues following prolonged storage; this may limit certain waste materials and/or require design changes to the waste package (this may then require revalidation for transport and disposal).
• The cavity sizes of different storage boxes are presented in Figure 1 as a function of shield wall thickness (which is the cast thickness of the container); sizes of these are given in Table 1. For example, the number of different sorts of containers required to package a 1000m3 waste volume is considered, and is shown in Figure 2 (excluding any adjustments for packing fractions or voidage.

Figure 2 shows two key trends:

• Significantly more waste packages will be needed using concrete shielding than with ductile cast iron shielding for equivalent sizes of boxes
• Larger-capacity containers present the most efficient packaging option in terms of the number of containers required. The larger-capacity containers offer the opportunity to reduce the amount of waste processing by allowing larger items to be packaged.

Croft has developed a range of RSCs in DCI: these are called Safstores (see Table 1). These meet UK requirements for both transport and disposability. In addition Croft also develops bespoke solutions to meet specific customer needs. A typical Croft IP-2 Safstore (design patent pending) has a double lid arrangement: a shield lid of the same thickness as the container body, and an outer transport lid with replaceable verifiable seal system. The first Safstor containers were fabricated in mid-2013 and have been shipped to customers.

About the authors

Mark Janicki, project manager, Croft Associates Ltd, F4 Culham Science Centre, Abingdon, Oxon OX14 3DB, UK
A version of this paper was presented at PATRAM 2013, the 17th International Symposium on the Packaging and Transportation of Radioactive Materials, August 17-23, San Francisco, California.