2.1. BACKGROUND
The assessment context defines the scope and content of the safety assessment. Specifically, it specifies the assessment’s:
• Purpose and scope (Section 2.1);
• Target audience (Section 2.2);
• Regulatory framework (Section 2.3);
• End-points (Section 2.4);
• Philosophy (Section 2.5); and
• Timeframes (Section 2.6).
2.2 PURPOSE AND SCOPE
The GSA has the following main purposes:
1. To demonstrate and build confidence in the use of narrow diameter boreholes as a safe disposal concept for disused sealed radioactive sources of less than 110 mm in length and 15 mm in diameter.
2. To produce a GSA for an envelope of disposal system conditions and assumptions against which a specific disposal system can be compared by identifying:
• Inventories suitable for disposal using the borehole disposal concept;
• Suitable levels of engineering;
• Suitable site characteristics;
• The need for and duration of the institutional control period required to provide adequate safety; and
• The half-life around which there is no practical limit for disposal from a post-closure perspective.
3. To identify the key parameters that need to be characterized for a specific site.
The GSA’s scope is the assessment of the post-closure (i.e. once the waste has been emplaced and the borehole backfilled and closed) radiological impacts on humans arising from the disposal of disused sealed radioactive sources at least 30 m below the ground surface in a narrow diameter borehole.
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2.3. TARGET AUDIENCE
This publication is a technical report and as such is written primarily for a technical audience whose prime interest is in the regulation and implementation of safe radioactive waste disposal2. The publication is considered to be of particular interest to those countries that have disused sealed radioactive sources and no suitable disposal options at present.
Two main technical audiences can be identified.
• ‘Developers’ including those individuals and organizations that have a direct interest in the disposal of disused sealed radioactive sources to borehole facilities. This group could include not only any organization directly involved in pursuing a disposal facility development programme, but also other nuclear industry organizations and radioactive waste producers.
• ‘Regulators’ including those organizations who would have a direct responsibility to decide whether to grant a ‘licence to operate’ a borehole disposal facility. In addition, there could be a wider range of organizations, such as local authority and other governmental organizations, which would need to be consulted if a borehole facility were to be developed for the disposal of disused sources.
Included in both these groups are the scientists who would provide technical support to the developers and regulators.
It is recognized that there is a range of other audiences that could be interested in the borehole disposal of disused sources (for example the media, politicians, and the public). However, given its technical focus, this publication is not specifically aimed at these audiences. It is recognized that additional publications will have to be developed that is tailored to the specific interests and needs of these other audiences.
2.4. REGULATORY FRAMEWORK
The assessment is not related to any specific site, organization or country, and so it is considered inappropriate to use a particular country’s regulatory framework. Furthermore, few countries have national requirements for the safe disposal of disused sealed radioactive sources that can be used as regulatory framework for the assessment, and where frameworks do exist, for example in South Africa [13], guidance provided related to issues influencing the post-closure safety of a borehole disposal facility is relatively limited. Given this, it is considered appropriate to use the recommendations of the IAEA Specific Safety Guide [8]. This Specific Safety Guide provides post-closure protection objectives and criteria which in turn are based on the recommendations of the IAEA and the ICRP [14], [15], [16].
Consistent with Ref. [8], the GSA adopts an individual effective dose constraint of 0.3 mSv/y for adult3 members of the public for all potential future exposures other than those arising from human intrusion.
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2 The report assumes that the reader is familiar with the technical terms used in safety assessment. Key technical terms are defined in Ref. [1].
3 Doses to children and infants could also be calculated, especially if there was a need to demonstrate consideration of a wide range of calculation end points. However, various post-closure assessment studies, such as Refs [17], [18], have demonstrated that the differences between adult, child and infant doses are usually less than a factor of two. Therefore, for the purposes of the GSA, consideration will be limited to adult doses as an indicator of impacts.
2.5. END POINTS
In most post-closure safety assessments, some measure of impact on humans or the environment is the calculation end point. The waste activity concentrations and total activity levels (i.e. the facility inventory) are usually the starting points of the assessment. In contrast, in this publication, the calculation end points are the reference activity levels for disposal to a borehole, which can be expressed as total activity values and per waste package activity values for each radionuclide, and the measure of impact (the annual individual effective dose), can be seen as the starting point of this calculation. However, in practice, the calculation of reference activity levels first requires a unit inventory for a borehole (1 TBq of each radionuclide per waste package in a borehole) to be assumed for which the dose is calculated. Assuming a linear relationship between the inventory and the dose4, total and per waste package activity levels that meet the appropriate radiological protection criteria can then be derived for each radionuclide disposed in the borehole. Further details concerning the calculation of reference activity levels are provided in Section 5.3.
Radiological impacts on non-human biota are not considered in this publication since it is assumed that if individual humans are shown to be adequately protected, then non-human biota will also be protected, at least at the species level [19]. The basis of this assumption is currently being investigated by various international organizations such as the International Commission on Radiological Protection (ICRP), the IAEA and the European Commission (see for example Refs [16], [20], [21]).
However, in the absence of any, as yet, clear consensus and guidance on the assessment of radiological impacts on non-human biota, the recommendations of ICRP Publication 60 [19] are adopted.
Non-radiological impacts, which might arise from the content of chemically or biologically toxic materials in the waste (for example beryllium in some Am-241 sources) or engineered barrier materials, are considered to be beyond the scope of the GSA given its emphasis on radiological impacts.
2.6. PHILOSOPHY
Different approaches can be applied to the assessment of the end points discussed in Section 2.5.
Whilst the nature of the end points may have been clearly defined, the nature of the approach used to calculate the end points also needs to be made clear. From this perspective, the assessment philosophy is an expression of the approach that is applied to the assessment.
Consistent with best international practice, the ISAM Safety Assessment Approach (Fig. 1) is used to undertake the assessment with the aim of ensuring that the assessment is undertaken and documented in a consistent, logical and transparent manner.
2.6.1. Nature of assumptions adopted
In undertaking an assessment, various assumptions have to be adopted. Assumptions are often categorized as ‘realistic’5 or ‘conservative’6, although, care needs to be taken when using such terms.
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4 The limitations of this assumption are discussed in Section 5.3.
5 Realism can be defined as “the representation of an element of the system (scenario, model or data), made in light of the current state of system knowledge and associated uncertainties, such that the safety assessment incorporates all that is known about the element under consideration and leads to an estimate of the expected performance of the system attributable to that element”.
6 Conservatism can be defined as “the conscious decision, made in light of the current state of system knowledge and associated uncertainties, to represent an element of the system (scenario, model or data) such that it provides an under-estimate of system performance attributable to that element and thereby an over-under-estimate of the associated radiological impact (i.e. dose or risk)”.
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A mixture between a realistic and a conservative approach is applied to the GSA. The key issue is to document and justify the nature of each assumption in the assessment. Typically, realistic assumptions are used where information is available and the associated uncertainty is relatively well known, whilst conservative assumptions are used where the information is highly uncertain.
2.6.2. Data availability
The assessment is not related to any specific location and, consequently, is carried out using well-justified values for the defined disposal systems derived from a range of national and international literature. All data sources are documented. The near field conditions are based on parameter values obtained as part of the development of the borehole disposal concept under the IAEA’s AFRA project [9], supplemented by generic data available from literature where necessary.
Since the assessment is generic, no site characterization studies were undertaken to supplement existing available information.
2.6.3. Treatment of uncertainties
The treatment of uncertainty is a key component of any assessment to establish the safety of a radioactive waste disposal facility. Uncertainties require consideration in a variety of ways and assimilation into the structure of the assessment as appropriate. They arise from three main sources [22]:
• Uncertainty in the evolution of the disposal system over the timescales of interest (scenario uncertainty);
• Uncertainty in the conceptual, mathematical and computer models used to simulate the behaviour and evolution of the disposal system (e.g. Owing to the inability of models to represent the system completely, and to approximations used in solving the model equations);
and
• Uncertainty in the data and parameters used as inputs in the modelling.
In addition, Ref. [12] suggests that a further type of uncertainty, subjective uncertainty (uncertainty due to reliance on expert judgement), is also linked with the above sources of uncertainty.
The uncertainty in the future evolution of the site is treated using a transparent and comprehensive scenario development and justification methodology (Section 4). Data and parameter uncertainty that exist are treated using a deterministic sensitivity analysis, whilst model uncertainties are treated using alternative conceptualizations and mathematical representations of the system (Section 6). Subjective uncertainties can be managed by using a systematic and transparent assessment approach, which allows subjective judgements to be document, justified and quantified (as far as possible).
2.7. TIMEFRAMES
Table 2 summarizes the timeframes for the various activities associated with the construction, operation, closure and subsequent release of the borehole from institutional control. It is assumed that following construction of the borehole, waste is disposed for a maximum period of one year since the volume of waste packages to be disposed is small (less than 0.2 m3) and, from an operational (and post-closure safety) perspective, it is best for this to be disposed over a relatively short period of time.
It is assumed that following the disposal of the disused sources, the site is closed immediately and the institutional control period starts. During this period, surveillance of the site might be undertaken for the purpose of public assurance (active institutional control). Local/national government records and planning authority restrictions may be maintained to prevent unauthorized use of the land and inadvertent human intrusion (passive institutional control). For the illustrative purposes of the GSA (and consistent with the lower end of the range considered in Ref. [23]), the duration of the institutional control period is assumed to last 30 years. During the institutional control period, it is
assumed that members of the public do not have access to the land in the immediate vicinity of the borehole and that inadvertent human intrusion into the facility does not occur.
TABLE 2. TIMEFRAMES FOR THE VARIOUS ACTIVITIES ASSOCIATED WITH THE CONSTRUCTION, OPERATION, CLOSURE AND SUBSEQUENT RELEASE OF THE BOREHOLE FROM INSTITUTIONAL CONTROL
Activity Timeframe
Borehole construction and waste emplacement About one year
(at a maximum)
Site closure Immediately following the waste disposal operation
Institutional control period (e.g. surveillance, local/national government records, planning authority restrictions, site marked on official maps)
Illustrative period of 30 years
No control (neither active nor passive) – all
records/knowledge conservatively assumed to be lost From 30 years onwards
In terms of the cut off time for calculations, the generic regulatory framework adopted for the assessment does not impose any explicit limit on the timescale for assessment. Therefore, calculations presented in the GSA are undertaken out to a time when it can be demonstrated that the peak value of the primary safety indicator (dose) has been passed for the radionuclide and disposal system of interest. It is important to recognize that uncertainties associated with these estimates will increase as the timescales become longer.
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