ANNEX Waste Acceptance Criteria for Disused Sealed Radioactive Sources in the IAEA Borehole Disposal Concept
1. Introduction
Sealed sources are widely used throughout the world in industry, research, agriculture and, not least, medicine. They are an indispensible part of modern life and make a very significant contribution to the health, safety and wellbeing of mankind. The number of different radionuclides used in sealed sources is relatively small - there are, perhaps no more than a few dozen in common use. In terms of activity and half-life, however, the range is enormous. Sources used as remote thermal generators or food sterilisers may reach the Peta-Becquerel levels of activity. Radionuclides used in sources can have half-lives of less than a year to thousands of years. Difficulties may arise when sealed sources fall out of use. When this happens it is rarely because the source is no longer sufficiently radioactive; rather, it occurs because the device in which the source is placed is no longer used. This may be because it has ceased to function properly or because it has become obsolete or, perhaps, because the enterprise has gone out of business. There are many examples where control over radioactive sources has been lost in such circumstances and, sometimes, deaths and serious economic losses are the result. This is a particular problem in countries that have no nuclear infrastructure. The IAEA Code of Conduct on the Safety and Security of Radioactive Sources [1] makes recommendations regarding the return of disused sealed sources (DSS) to the manufacturer but often this is not possible, especially where the sources pre-date the Code of Conduct. A compounding factor is that many of these DSS have relatively long half-lives. Further, it is usual for a sealed source to have a small mass so that, even when the level of activity is modest, the specific activity (e.g. Bq/kg) is usually too high for a disused source to be considered suitable for near-surface disposal. Especially in developing countries, therefore, it is not unusual for DSS to be held under poor storage conditions with little prospect of a permanent solution.
As part of the IAEA AFRA project, a number of African Member States requested the Agency to respond to this situation by developing a simple, economic yet safe means of disposal for DSS. The result is the IAEA BOSS (BOrehole disposal of Sealed radioactive Sources) systemBorehole Disposal Concept (BDC). It represents a technically complete disposal solution for the world’s DSS and is especially appropriate for Member States that do not have a nuclear infrastructure but where, nevertheless, long-term management of DSS is a concern. The BOSS systemBDC was developed for the IAEA by the Nuclear Energy Corporation of South Africa and is based on a high level of physical containment provided by placing the DSS inside a stainless steel capsule within a stainless steel container; putting these into a borehole at more than 30m depth and then backfilling with a grout based on sulphate-resistant Ordinary Portland Cement. It is estimated that, under ideal conditions (which are not difficult to achieve), absolute physical containment can be maintained for hundreds of thousands of years. This is sufficient to allow the vast majority of DSS to decay to exemption levels. The technology is simple and widely available: borehole construction, for example, employs equipment that is used for exploration of water resources; stainless steel is easily welded; and handling of DSS with activities below 1 Ci of Co-60 (or its equivalent) can be done in a lightly shielded facility using standard 100 mm lead bricks. For handling larger sources (up to 1000 Ci of Co-60 or its equivalent) a mobile hot cell has been designed and manufactured.
Safety assessments indicate that the system is capable of providing the required level of operational [2] and post-closure safety [3, 4]. Generic waste acceptance criteria have been derived for α, β, γ and neutron sources using either the lightly shielded conditioning unit or the hot cell. Despite the simplicity of the BOSS systemBDC, safety is not compromised in any way. What the system does, in effect, is to take advantage of the small physical size and single-radionuclide nature of DSS.
It is emphasised that the BOSS systemBDC, as it is currently envisaged, has been specifically developed for the disposal of DSS and the secondary wastes that might occur during the associated pre-disposal activities. It is not intended for the disposal of e.g. radioactive wastes arising from the operation of Nuclear Power Plants (NPPs), nor for the disposal of spent fuel from research reactors or other types of radioactive waste.
Finally, it is important to note that the WAC described here are preliminary for three reasons:
(i) they are generic (i.e. not specific to a particular site);
ii) they exist separately from a specified inventory; and
(iii) they have been formulated in the absence of some of the pre-disposal procedures (especially those relating to welding of the containers and use of the mobile hot cell).
It follows that, before the BOSS systemBDC can be deployed, further work will be needed to provide the missing information.
2. Waste acceptance criteria for the borehole disposal concept
2.1. Components of the system
For the BOSS systemBDC, the required components are standard and may be briefly listed as follows:
· stainless steel container with concrete cement insert that the stainless steel capsule (holding the sources) fits into;
· borehole with containers equally spaced along it (spacing about 1 m) and with a no container placed at a depth of less than 30 m;
· borehole casing;
· concrete cement backfill completely filling the void between casing and borehole wall;
· cement concrete backfill completely filling the void between containers and casing;
· cement concrete plug above the disposal zone;
· deflection plate (to deflect any drill that may be inadvertently or deliberately attempting to drill a vertical borehole along the original trajectory);
· surrounding geosphere;
· lightly shielded (100 mm lead brick) conditioning unit in which low activity (<1 Ci Co-60 equivalent) DSS are welded into capsules and containers;
· mobile hot cell in which high activity (<1000 Ci Co-60 equivalent) DSS are welded into capsules and containers; and
· transfer flask for moving high activity DSS to the borehole.
In terms of safety functions, the last three items in the list are entirely devoted to operational safety and, in particular, to providing the needed level of shielding. Clearly, the hot cell and the transfer flask are needed when high activity sources are being handled (Category 1, 2 and possibly 3 see [5]). For lower activity sources (some Category 3 and Category 4 and 5), light shielding and well-developed procedures are adequate. The other items are primarily provided for post-closure safety which, as explained, relies to a large degree on absolute physical containment.
2.2. Identification of key parameters
2.2.1. General
As with any WAC, a range of general requirements will apply. These include the need for source characterisation (including a statement of whether the DSS was leaking when encapsulated), container and capsule specification and certification and labelling. Where a source is delivered for conditioning within the machine or part of the machine in which it was originally used, drawings should be provided so that the source can be safety removed.
2.2.2. Administrative
Documentation accompanying the source should be original and should tally be consistent with the physical information e.g. serial numbers, device description., Where the original documentation has been lost - in the case of an orphan source perhaps - it will be necessary to recreate it using, for example, the IAEA catalogue of sealed sources [6]. It will also be necessary to characterise the source of course [7]. Having created a paper record, disposal containers will be engraved with the radiation trefoil, the radionuclide(s), the activity and the date. A record of the predisposal activities should form part of the documentation.
2.2.3. Radionuclide inventory per container
The radionuclide content of an individual container may be limited by:
· tThe maximally permissible dose to the operator during pre-disposal handling (dose limit or constraint); and,
· tThe maximally permissible dose to the public in the post-closure period (dose limit or targetconstraint) and heat output.
These are examined in greater detail in the Quantification of Acceptable Limits section.
It will benefit operational safety if the conditioning unit is provided with shielding that is suitable for the sources to be handled (e.g. lead for gamma sources, polythene for neutron sources). It may be advisable, therefore, to keep gamma and neutron sources separate by not mixing them within the same capsule.
2.2.4. Radionuclide inventory for the repository
It is arguable whether limits placed on the total inventory of a repository (in this case a borehole) can be said to form part of the WAC. We include them here regardless.
While it is clear that a borehole disposal will need to be designed to suit the chosen site, the main consideration here is the location and length of the disposal zone (the disposal zone is the part of the borehole that contains the disposed wastes) which should be chosen to make best use of the site properties. Apart from this, however, we may consider the borehole design to be relatively fixed. We know, fFor example, that grade the use of Type 316 L stainless steel will be used for the container and capsule and that asulphate-resistant OPC for the concrete cement backfill will likely be used. Conceivably, bentonite or sulphate resistant cement could be used in place of OPC if the geochemical conditions warranted it.
The generic safety assessment (GSA) [3, 4] aims to present a post-closure safety assessment for the BOSS conceptBDC that is not site specific (hence generic). To do this, it envisages a matrix of hypothetical geological and climatic environments in which a disposal borehole disposal could be situated. In the most unfavourable environmental conditions (aerobic, humid, high chloride) containers are calculated to fail after by 1874,000 years. Under favourable conditions (anaerobic, low chloride) lifetimes arecan be more than ten times this. For those radionuclides that are sufficiently long-lived to still constitute a hazard at the time of container failure, safety assessment is performed as for other radioactive waste disposals. Assessment of the groundwater pathway begins by modelling the release of radionuclides from the near field, their migration through the geosphere into an aquifer and the subsequent extraction of contaminated water by means of a borehole for use by humans. These uses include drinking, irrigation of crops and, watering of cattle, milk consumption etc. In general, the permissible radionuclide content of the repository as a whole will largely depend on:
o the half-life of the radionuclides concerned;
o the geochemical properties of the site, noting that these largely determine the corrosion rate of the containers and capsules; and
o the hydrogeology and geochemistry of the surrounding geosphere because these determine the amounts and concentrations of radionuclides reaching the biosphere and, hence, the calculated doses values.
An example of some typical results from the GSA are is shown in Table 1 [34]. The data, which are here limited to radionuclides with long half lives, show that containment is best achieved under saturated, low flow and chemically reducing conditions.
2.3. Quantification of acceptable limits
2.3.1. Radionuclide inventory per container
As explained, the activity limit for a single waste disposal container is determined by one of the following:
tThe maximally permissible dose to the operator during pre-disposal handling (dose limit or constraint);,
tThe maximally permissible dose to the public in the post-closure period (dose limit or targetconstraint); and
heat output.
These are examined in turn.
Dose to the operator during pre-disposal or disposal activities
External doses to the operator arise from handling gamma sources, neutron sources and, if they are energetic enough to produce bremsstrahlung radiation, beta sources. Any assessment will need to take account of:
o the strength of the DSS and the type of radiation that it emits;
o the distance of the operator from the source;
o the operator’s exposure time;
o the amount of shielding; and
o the dose target.
In the case of the BOSS systemBDC, considerations of operator dose will lead to two sets of acceptance criteria: one for the lightly shield facility; and another for the hot cell.
To establish a limit for the activity of a specific radionuclide (or combination of radionuclides) that may be placed in a container, it is first necessary to know what operations are to be performed and the equipment that will be available. Operator dose is determined by the strength of the source, the amount of shielding, the distance of the operator from the source and the time of exposure. These parameters are not always obvious. When estimating limits for the size of DSS that could be handled in the lightly shielded facility, for example, it was observed that most of the operator dose occurred during the relatively short time periods (a few seconds only) when DSS were being transferred unshielded between the various work stations [8]. Such information is only known from trials and practical experience [9].
The last item in the list, the dose target, should be stated in the radiation protection plan but, almost certainly, will lie in the range 2 to 10mSv. Knowledge of the various parameters will allow the operator doses to be calculated and reasonably minimised by, for example, despatching larger DSS to the hot cell and/or spreading the dose over a larger number of operators. Depending on the inventory, external dose rates for gamma, bremsstrahlung and neutron irradiation may need to be considered.
Dose to the public in the post-closure period
The GSA [3, 4] derives limits for the activity of radionuclides on the assumption of a single borehole containing 50 packages. Limits for each package are then calculated simply by dividing the total activity by 50. This is probably adequate for a generic case. With the benefit of site characterisation data and an actual inventory, however, further calculations would allow the number of boreholes needed and the radionuclide content of each package to be precisely defined. Almost certainly the values would be more relaxed than those estimated (as here) on a generic basis. Table 1 gives the maximum permissible activity of radionuclides with a half-life of more than 30 years that could be placed in a disposal package for various disposal environments for the Design Scenario’s liquid releases. In addition, the GSA for high activity sources [4], which takes into account temperature and radiolytic effects, estimates a bounding value of 5E+14 Bq per package for Cs-137 for the most restrictive disposal system (i.e., disposal in a saturated, high flow, porous geosphere). Per package limits for solid releases for the Design Scenario are 4E+10 Bq, 1E+10 Bq and 9E+11 Bq for Pu-238, Pu-239 and Am-241, respectively [3]. The per package activity limit for all other radionuclides are in excess of 2E+16 Bq [3, 4].