When examining storage options for RAW, it is important to consider the whole storage system rather than concentrating on just the store building itself (CoRWM, 2009). A number of interacting components and operations combine and contribute to create the necessary robust, safe and secure storage arrangements. These factors must be considered in an integrated manner. There are two main concepts in the storage of RAW. If the pack­aged waste forms are basic, then a high quality often shielded store will be needed. On the other hand, if the waste form is high quality and shielded, then the store can be of poorer quality or the waste containers can simply be left in the open. A generic shielded store is shown in Plate I (between pages 448 and 449) and an example of a high quality store has recently been constructed at Hunterston in Scotland (Fig. 1.15(a)) which has 2 m thick reinforced concrete walls and roof and careful control of atmosphere. Figure 1.15(b) also shows the stillages containing 4 ILW drums that will be stacked on top of each other in the store (and eventually in the GDF).

The waste form or product, its container, the building structure, the ven­tilation system, the handling equipment, the monitoring and inspection regime and the maintenance and refurbishment regime all have roles to play in ensuring safety and security of the store. As illustrated in Fig. 1.16, the waste storage system involves a number of levels. The wasteform (1) is the primary protective barrier, the waste container (2) is the secondary barrier. Control of store environment (3) is important in maintaining the

integrity of the waste form and waste container, while the store structure (4) is the final layer of weather/atmosphere protection for waste package and an important element of physical security of waste.

Packages inevitably evolve during storage, and those changes affecting the safety function need to be understood and controlled to satisfy the regulators of the safety of the store and waste. Different storage concepts and designs require different performances from these various components and operations and therefore place different degrees of reliance on them. Quite different combinations of them can provide equally safe and secure storage. For example, most existing modern stores in the UK have massive concrete structures holding unshielded containers, but the alternative ‘mini­store’ concepts rely on heavily shielded containers within lightly built stores. This latter concept is used in some other EU countries. In a storage system, not every component need last for the whole design life. It is possible at the design stage to plan to replace or refurbish various components and build in at the outset specific features to enable this. More straightforward items to consider are building fabrics, external ventilation systems and power supplies. The more complex refurbishments or replacements to con­sider are cranes, active area surveillance equipment and major building structures.

In the late 1970s and early 1980s, the need for alternative storage in the US began to grow when the storage ponds at many nuclear reactors began to fill up with stored spent fuel. As there was not a national storage facility in operation, utilities began looking at options for storing spent fuel. Dry cask storage was one of the most practical options for temporary storage. The first dry storage installation in the US was licensed by the Nuclear Regulatory Commission (NRC) in 1986 at the Surry NPP in Virginia. Spent fuel is currently stored in dry cask systems at a growing number of power plant sites. The NRC estimates that the SF ponds at many US NPP will be full by 2015, so requiring the use of temporary storage. The 2008 NRC guidelines call for fuels to have spent at least five years in a storage pool before being moved to dry casks. Due to the demise of the Yucca Mountain project, more US SF and waste is being stored in sealed metal casks filled with inert gas. Examples of high quality containers include CASTOR, which is an acronym for CAsk for Storage and Transport of Radioactive material (Fig. 1.17).

In the UK, options being examined for SF include multi-purpose contain­ers (MPC) suitable for storage, transport and disposal of a range of SF types (Fig. 1.18). As well as high quality packages for SF, they have also been developed for less active wastes. So-called yellow boxes (Fig. 1.19) have been used extensively in Europe and used to store spent resin waste from existing storage tanks at the Dungeness plant in England. The containers are transportable and offer self-shielded protection, weighing around 18 tonnes when empty. The waste is expected to be stored in them for at least a decade.

As more waste is generated and being stored, but in many countries without an end-point of geological disposal in sight, an issue is whether to store all waste at sites or to consolidate wastes at centralised national or regional stores. The BRC, for example, recommended this option be exam­ined in the US and it is also being considered in the UK.

The waste package as a final product of RAW processing, prepared and ready for long-term storage, or disposal, consists of two components: the waste form and the waste container. In some cases there are also additional barriers or shielding used to increase the safety features of the waste package. The waste package should be prepared in a form corresponding to the requirements for handling, transport, storage and disposal.

A wasteform is defined as ‘radioactive waste after treatment and condi­tioning, usually in solid form, prior to its packaging into the waste con­tainer’. A waste container is defined as ‘the vessel into which the waste form is placed for future handling, storage and disposal’. The waste container fulfils the role of a protective barrier and shielding tool. The waste container should guarantee the tightness for the entire period of storage and/or dis­posal of the waste.

To ensure the long-term safety of waste disposal, WAC should be devel­oped based on a safety assessment of the considered disposal options and should be approved by the relevant authorities. Waste acceptance require­ments (criteria) are by definition quantitative or qualitative criteria for processed RAW to be accepted by an operator of a repository for disposal, or by an operator of a storage facility for storage. WAC are specified by the relevant authorities, or proposed by an operator and approved by the rel­evant authorities. Waste acceptance requirements might include, for example, restrictions on the activity concentration, or the total activity of particular radionuclides (or types of radionuclide) in the waste, or require­ments concerning the waste form or waste package [10]. In the past, the term waste acceptance criteria was only applied and used in the context of waste disposal. Later on, the approach to specify WAC has been extended also to some other steps of the waste life cycle — in particular, for transport and storage. In general, WAC can be specified for any foreseen waste man­agement operation and handling. WAC can prescribe and cover various waste package features and properties, such as: [2]

• requirements for waste container (e. g., design features, mechanical sta­bility, thermal resistance and also some additional features — shielding, corrosion protection, etc.);

• limitations for activity — permissible activities of individual radionu­clides, total activity, activity of selected critical radionuclides;

• radiological safety parameters — surface dose rates, surface contamination.

There are usually several other requirements, developed and specified based on the safety assessment of risks of the planned operations (trans­port, storage, disposal, etc.) with prepared waste packages. A more detailed description of this subject can be found in Chapter 3.

WAC are site-specific, but not container-specific. They are developed based on a safety assessment of the design and implementation of the waste disposal or waste storage facility and eventually waste transport tools. WAC are general criteria, usually not specified for particular waste containers and/or waste packages. Therefore they are not simply applicable in every­day technological practice.

To overcome the above limitations of WAC, the general waste acceptance requirements are usually transformed into detailed waste package specifica­tions (WPS). WPS should be developed and individually implemented for each type of RAW package and should reflect specific characteristics of the waste package. WPS are therefore waste package (and also waste con­tainer) specific and they are normally substantially more detailed than WAC. They shall be a consistent part of the QA/QC system applied by the waste package producer. Application of WPS allows simple control and verification of waste packages for both the waste producer as well as the waste disposal facility operator. Compliance of waste packages with WPS is considered a guarantee of compliance of said waste package with the WAC for a particular waste lifetime step. More details and guideline for development of WPS can be found in Ref. [11].

Waste acceptance criteria for disposal can normally be developed based on the safety assessment of an available, already constructed, or intended waste disposal facility. In any case, a clear idea of the waste disposal option should be available. However, many countries are in the situation where processing of RAW is unavoidable and the decision regarding a disposal facility is still deferred. In such cases, there are two principal options on how to proceed with waste processing to avoid future complications with acceptance of waste packages at the disposal site:

• Develop and apply generic WAC, based on international experience, approaches, and analogy with similar nuclear programmes. In this manner, a sufficiently conservative approach shall be taken and it shall be demonstrated that a national waste management policy and vision of future disposal options had been considered. These criteria can then be used for development of waste package specifications for available waste management technologies and waste packages already in use.

• Develop and apply only waste package specifications for particular waste streams and waste processing technologies, based on a detailed analysis of potential disposal options. A sufficiently conservative approach and margins in critical parameters shall be applied to avoid future problems with acceptance of waste packages for disposal.

4.1.1 Gaseous and airborne waste treatment

Operations involving the handling of radioactive material may generate airborne radioactive contamination. The basic difference between airborne effluents and RAW in condensed (i. e., liquid or solid) phase is that airborne material has no definite volume and its dispersion in the environment is very fast. Gaseous and airborne wastes are discharged to the environment through ventilation and air-cleaning systems, which are a vital part of the general design of a nuclear facility [ 11-14] . The combination of a well — designed ventilation system with thorough cleaning of exhaust air is the main method of preventing radioactive contamination of the air in working areas and in the surrounding atmosphere. Ventilation and air-cleaning systems should provide efficient treatment of gaseous streams under normal operations, maintenance and accident conditions. High efficiency particu­late air (HEPA) filters are most commonly used for removal of radioactive particulates and aerosols from gaseous streams [12, 14]. Sorption beds charged with activated charcoal are common for removal of volatiles (e. g. iodine) and as delay beds for noble gases. Wet scrubbers are used for the removal of gaseous chemicals, particulates and aerosols from process off­gases. Additional components of the air-cleaning system include pre-filters, and temperature and humidity control systems, as well as monitoring equip­ment such as gauges that show pressure differentials. The treatment of gaseous streams results in secondary waste either solid (spent filters or sorption beds) or liquid (scrubbing solutions). The physical and chemical properties of the selected air-cleaning media should therefore be compat­ible with the treatment and conditioning processes for the solid or liquid waste streams in which they will be treated.

For consistency, the definitions given in Chapter 4 of this book, which are from the IAEA [ 12, 13] , are adopted here except for the definitions of encapsulation vs. embedding where examples have been used to make the distinctions clearer.

6.1.1 Conditioning

Conditioning includes those operations that produce a waste package suit­able for handling, transport, storage, and/or disposal. This may include the following:

• conversion of the waste to a solid waste form

• enclosure of the waste or waste form in containers

• providing an overpack if necessary.

The waste form is the waste in its physical and chemical form after treat­ment and/or immobilization prior to packaging. Thus the waste form is a component of the waste package.

This section presents the degradation behaviour of SNF in mild and near­neutral environments under (i) oxidizing or reducing aqueous disposal conditions, and (ii) in dry storage environments. During the aqueous dis­solution of SNF, highly soluble fission products such as Tc-99 or I-129 are released congruently with (i. e., in proportion to) the SNF matrix (UO2 ) dissolution. On the other hand, actinides such Pu-239 or Np-237 are released at a concentration below or equal to their solubility limits (or colloid con­centration), which are in turn determined by the SNF matrix dissolution rate, groundwater flow rate and solubility limit. Colloids are suspended solid particles of less than 1 micrometer in size that can contain actinides. An oxidizing aqueous environment promotes electrochemical dissolution of the SNF matrix in soluble species with the aid of oxidants such as dis­solved oxygen and hydrogen peroxide (Shoesmith, 2000) . In a reducing environment, the UO) matrix will dissolve chemically in soluble species (Sunder and Shoesmith, 1991). Generally, the electrochemical dissolution rate is faster than the chemical dissolution rate. In the presence of radiolysis effects, the SNF matrix may dissolve in either an electrochemical or a chemical process, depending on the magnitude of the radiolysis (Ahn et al, 2011a). In conjunction with container failure and sorption and/or flow behaviour of backfill, the SNF matrix dissolution serves as the source term of radionuclide release in the PA. In a dry storage environment, mechanical degradation of the SNF matrix could occur by air oxidation/humid air hydration or impact fragmentation upon the canister failure under normal conditions (e. g., SCC failure) or external hazard conditions (e. g., aircraft or seismic impact). In the canister, if incomplete drying of SNF assemblies occurs, the residual water may increase RH sufficiently to oxidize (by oxygen from the radiolysis of water molecules) or hydrate the SNF matrix. With severe external hazards, high temperatures or impact stress may frag­ment the SNF matrix by oxidation or mechanical disintegration. The respir­able SNF particles (i. e., suspended aerosol, less than 10 pm [3.9 microinch] in size) produced by the fragmentation serve as the primary source term for radionuclide release in air.

Any technological installation, as well as the disposal facility of RAW, is a source of potential danger because in the course of operation, they are forced to interact with the environment and, therefore, to provide some impact on it. Adverse effects associated with contamination of the sur­rounding area can be assessed via the radiological condition of the territory using a five-score system: (1) normal; (2) satisfactory; (3) accident; (4) emer­gency; (5) disaster (Busygin et al., 2009). The ratio r = A/RL can be used for quantitative assessment. Parameter A is an actual level of contamination, and parameter RL is the reference level, typical for this area. The value of r is continuous and varies widely, which complicates the classification and interpretation of the effects of radioactive contamination. Therefore, we propose a system assessment area, based on rankings of r in accordance with Table 9.1 .

Contamination of the environment and the site can clearly result from accidental events in radiation-hazardous facilities. Regulatory and legisla­tive documents allow a posteriori estimation and classification of emergency events, based on measurements of actual contamination levels and compar­ing these results with a certain threshold. However, in practice it is neces­sary to calculate and predict the consequence of events prior to their occurrence, i. e. to give an a priori assessment of the events. Since the events themselves, as well as their effects, are effectively random, then the evalu­ation must be made in terms of random variables, i. e. must have a proba­bilistic nature. An international scale is used to link the seven levels of technological accidents at NPP and their consequences on the environment (INES, 2008) as given in the first two columns of Table 9.2 . Based on this scale, we propose an additional relationship between the levels of incidents with their probabilities for all radiation-dangerous objects, which do not belong to the nuclear fuel cycle facilities. These relationships are given in columns 3 and 4 of Table 9.2 (Puzanov et al., 2004, p. 220).

In some regulations (GAN, 2000) for objects which do not involve the nuclear fuel cycle, a three-score grading system for the class of incidents and

their consequences is recommended, as well as liaison on the levels of con­tamination. Using Table 9.3 we can associate a class of incident with the probability of their realization and we can specify the requirements for safe operation of facilities. As a case study, we consider the emergency situation at the NPP ‘Fukushima-1’. According to many experts, the situation is con­sistent with a IV-V level of complexity. Initially the reactor coolant system failed under exposure to the earthquake measuring 9 on the Richter scale. According to Table 9.3, we can conclude that, due to this level of earthquake, the failure probability of the cooling system must not exceed 10-3-10-2. It is important to emphasize that the initiating event, i. e. earthquake itself, is not included in the script, since it is not an element of the event tree but is the external condition under which the event tree is realized.

11.1.3 Environment contamination from the accident and its current state

The extent of surface contamination

The 1986 accident at the Chernobyl nuclear power plant, resulted in a sub­stantial release of radionuclides to the atmosphere and caused extensive contamination of the environment. According to the International Atomic Energy Agency (IAEA, 2006), a small part of the nuclear fuel (up to 3.5%) and a substantial fraction of volatile radionuclides were released from the damaged unit 4. The total activity amounted to approximately 12.5 x 1018Bq, and included 6.5 x 1018Bq of noble gases (IAEA, 2001).

A considerable territory of the former Soviet Union, particularly in Belarus, Russia, and Ukraine, as well of Western Europe, primarily the Scandinavian countries and the Alpine region, was severely contaminated. High levels of radioactive contamination in areas outside the Chernobyl Exclusion Zone arose for the following reasons: release of contaminated masses into the atmosphere to a height of 2,000 m and their intense move­ment at these altitudes; rainfall; and the presence of complex landscapes that dictated changes in directions and altitudes of the contaminated air masses movement.

The overall area of Western Europe countries where levels of 137Cs con­tamination exceeded 20 kBq m-2 (almost 10 times higher that global back­ground levels) due to the Chernobyl disaster amounted to approximately 280,000 km2 . Almost 75% of Ukraine’s territory suffered from radioactive contamination by 1 37Cs, which exceeded the pre-accident levels by more than double. The radionuclide decay, which has occurred in the 25 years since the Chernobyl accident, substantially corrected a pattern of radionu­clide distribution over Ukraine ’s territory. Over this period, the area of localities where 137Cs contamination levels exceeded 10 kBq m-2 has reduced to almost half what it was immediately after the accident. The area of sites where 90Sr contamination exceeded 4 kBq m-2 is now less than one third, i. e. practically 90% of Ukraine’s territory is characterized by the pre-accident levels of 90Sr contamination.

However, the level and extent of Ukraine ’ s territory contamination by Pu isotopes have not changed. 241Am activity is gradually increasing due to MPu decay; and the area of its distribution where levels exceed 0.2 kBq m-2 shall be 30% wider than the area of plutonium isotope fallout having the same density. The area of Ukraine contaminated by ’Sr, 241 Am, and Pu isotopes is substantially smaller than that contaminated by 137Cs.

Severely contaminated (over 1.5 MBq m-2 of 137Cs) localities (almost 300 km2) within the boundaries of ChEZ will remain uninhabitable for hundreds of years. These water-producing areas shall remain a long-term source of surface water and groundwater contamination due to surface washout and vertical migration.

The Swedish effort to become self-sufficient in plutonium and uranium production left installations behind that needed to be decommissioned and

remediated. In 1988, the Swedish Parliament made a law regulating the financing and the responsibilities for cleaning up after the activities by AB Atomenergi, later Studsvik AB. Following that law, AB SVAFO was founded. SVAFO is currently in charge of all decommissioning work with the exception of sites owned and used by commercial nuclear power plants. These sites include the Active Central Laboratory (ACL), Ranstad uranium mine and the Agesta nuclear reactor.

ACL was originally intended for research and development of reprocess­ing and production of MOX fuel, although over the years it also hosted other nuclear activities. The laboratory was opened in 1963 and closed in 1997 (Johnsson et al., 2004). SVAFO acquired ACL and the ventilation and filtering building (AFC) in 1998. SVAFO decided to go for complete decommissioning with the ultimate goal of demolishing the building. The laboratory was contaminated mainly with Co-60, Sr-90, Cs-137, H-3 and transuranium elements. The work started in 1998 and was completed in 2005 when SVAFO sent an application for ‘free-release’ of the building to the Radiation Protection Agency (SSI, now Swedish Radiation Safety Author­ity, SSM). The buildings were finally demolished in 2006 (Hedvall et al., 2006; Johnsson et al., 2004).

The Ranstad uranium mining and processing facility was built between 1960 and 1965. It was test operated between 1964 and 1969. The geology is alum shale with uranium content of about 300 g per tonne. During the operations, 215 tonnes of uranium was obtained from 1.5 million tonnes of alum shale. With falling uranium prices during the 1960s, the mining was not profitable and after Sweden signed the non-proliferation treaty in 1968, there was no longer a need for a domestic production of uranium. The mine was an open pit mine 2 km long, 100 m wide and 10-15 m deep. The uranium had been extracted using sulphuric acid with a gain of up to about 60-70%. Consequently, there was a large amount of uranium left in the mill tailings, which amounted to 1,000,000 m3 covering an area of 230,000 m2 (Stridlund and Aquilonius, 1999a ).

When the mining permit expired in 1984, planning for the remediation started and was carried out during the period 1990 to 1992. The mill tailing deposit had natural stable slopes and had previously been covered by a thin layer of moraine. This layer was now covered with a sealing layer of clay — moraine mixture. A layer of crushed limestone above the sealing layer created a drainage layer. On top of this is a protective layer of 1.5 m moraine. The overall purpose was to prevent oxygen and water from reach­ing the mill tailings and thereby stop the leaching of metal into the water system in the environment.

When the mill tailings were placed in the area, two lakes were formed. The first lake was designed to collect the leachate from the depositions. The water was then transferred to a purification plant where it was treated with lime to precipitate leached metals, which were deposited in a sedimentation pond before discharging the water to the second lake. That lake is now called Blackesjon. Following the restoration, the water quality was moni­tored in a number of locations to verify the function of the remediation system. Lake Blackesjon is now connected to the existing natural water system since the set environmental goals have been reached (WSP Environmental, 2005).

The pumping of the open pit mine ceased in 1990 and it has now been transformed into a lake, Tranebarssjon. The lake bottom is backfilled lime­stone and alum shale, covered by a thinner layer of backfilled moraine (Stridlund and Aquilonius, 1999b). The shoreline has been smoothened and there is a natural growth of vegetation in and around the lake. The lake and the adjacent wetland has become a sanctuary for a large number of bird species.

The Nuclear Energy Agency (NEA) of the Organisation for Economic Co­operation and Development (OECD) is an international organisation of 31 Members[4] with the mission to assist its member countries in maintaining and further developing, through international co-operation, the scientific, technological and legal bases required for the safe, environmentally friendly and economical use of nuclear energy for peaceful purposes [20]. Radio­active waste management is one of the areas of NEA’s work which includes: nuclear safety and regulation, nuclear energy development, radiological protection and public health, nuclear law and liability, nuclear science, maintenance of data banks and information and communication. With respect to RAW management, the NEA focuses on providing assistance to member countries in developing safe, sustainable and societally acceptable strategies for management of all types of radioactive materials, with par­ticular emphasis on the management of long-lived waste and spent fuel and on decommissioning of nuclear facilities.

Disposal is emplacement of waste in an appropriate facility without the intention to retrieve, although some countries use the term disposal to include discharges of effluents to the environment [39].

Exempt waste is the waste that meets the criteria for clearance, exemp­tion or exclusion from regulatory control for radiation protection purposes as described in Ref. [40]. Exempt waste is acceptable for disposal in landfill sites used for domestic and industrial waste.

Simple trenches have been used for many decades for the disposal of short-lived low and intermediate level waste. They are generally considered appropriate only for those wastes including disused sources that will decay sufficiently within an anticipated period of institutional control (generally between 100 and 300 years) to represent no risk to the public, as determined by safety assessments. The design and function of such repositories are described in Refs [41-44] . Large-scale (typically thousands of cubic metre capacity) near-surface engineered vault repositories have similar contain­ment objectives and are used for similar types of RAW as simple trenches. Their engineering is intended to allow ease of waste emplacement and increased efficiency in the management and closure of the repository. The issue of post-institutional control intrusion can still be a dominant factor in waste acceptability [7, 11]. For the near-surface disposal option, a perform­ance assessment is also required to determine either that the activity of the RAW can be contained until it has decayed or, if some migration is antici­pated, that consequent doses are acceptable.

Near-surface shafts and/or boreholes can be considered as alternative or complementary to near-surface vaults. These disposal options have the advantages of being economical and also minimizing the probability of human intrusion. If necessary an engineered barrier system (EBS) can be added to the design and construction of these facilities to provide additional protection against radionuclide migration and human intrusion. More heavily engineered near-surface facilities have been designed with the specific intention of reducing the likelihood of intrusion by emplacement of a massive concrete plug or cover over a large shallow shaft or borehole. For example, a reinforced concrete slab at least 1 m thick is considered to be a deterrent to inadvertent intrusion. These intrusion-resistant designs [45] will be helpful if institutional controls break down before the typically envisaged 300-year period. However, they do not offer a sufficient guaran­tee against intrusion to be considered for disposal of higher activity or longer lived waste.

Radioactive wastes that are not acceptable for disposal in near-surface disposal facilities, because they will not decay sufficiently within the period of institutional control, may be suitable for disposal at greater depth in disposal units characterized by one of several configurations [7]. At present, with the exception of deep tunnels and mines, it is uncommon to find con­struction work (e. g., deep foundation engineering) carried out at depths greater than about 30 m [44] , so disposals at depths greater than this are only vulnerable to intrusion by deep drilling for water or mineral explora­tion — a much lower probability. As a result, the intrusion exposure risks posed by higher activity waste disposed of at intermediate depths are small. Shafts or boreholes to depths of several tens of metres or more are rela­tively simple to construct and can offer an attractive disposal option for small volumes of waste such as radioactive sources [44]. Evaluation of such options needs to consider the stability of the hydrogeological system over the time period of concern for containment, which may be several hundreds or thousands of years depending on the types of radioactive sources to be disposed of.

Very low permeability host rocks, with little or no advection of ground­water, can also provide adequate containment without the need for addi­tional EBSs. Some clay and claystone formations at intermediate depths can provide such an environment, and evidence of lack of flow can be obtained from pore water environmental isotope analyses and evaluation of any fracturing that may be present in the rock. The isolation capability of this option depends on the ability to provide good shaft or borehole backfilling and sealing. The use of natural materials that reconstitute the original properties of the penetrated rock formations is recommended for all or some part of the sealing system. This may involve removal of some lining or casing to allow sealing against the host formations. If the disposal borehole/shaft is subject to significant water inflow or the geotechnical characteristics of the geological materials do not allow the excavation to be sufficiently stable, EBSs need to be emplaced to provide a level of contain­ment commensurate with the hazardous life of the waste.

There are some disposal facilities for RAW in large rock cavities at depths of several tens of metres, generally in hard crystalline rocks such as granite (e. g., in Sweden and Finland). They are designed to contain short-lived low

and intermediate level waste. The containment provided by such repositor­ies often comprises massive concrete vaults or silos, with additional EBSs such as clay backfills and buffers. This type of containment should be ade­quate for the disposal of many, if not all, types of RAW. For emplacement of high activity waste in a mined, intermediate depth repository, it is necessary to consider packaging and activity concentrations that suit the thermal char­acteristics of the host rock and EBSs of the repository. In addition, disused mines and/or caverns can be considered for intermediate depth disposal. Such facilities have not been widely used for the disposal of RAW. The objective of using deeper boreholes, at depths typical of geological reposi­tories, would be to achieve greater isolation for limited volumes of RAW, including disused radioactive sources, in an environment that is character­ized by lower flow, more stable chemistry and longer potential return paths to the biosphere, compared with the other options. In a very low permeabil­ity environment (e. g., some clay and claystone formations), there may be no effective water movement at depths of a few hundreds of metres. In such conditions, provided an adequate borehole seal can be constructed, contain­ment of radionuclides is provided by the geological barrier and there is no requirement for supplementary EBSs beyond those needed to emplace the radioactive sources into the borehole and to maintain borehole stability during emplacement operations (casing and cementing). The option is par­ticularly suited to the highest activity and long half-life radioactive sources, for which long containment periods are required (e. g., -10-20 half-lives or more). For example, strong 226Ra sources could require isolation for -20,000­30,000 years. The depth and design of disposal also significantly reduces the likelihood of inadvertent intrusion, resulting in exposures to high concentra­tions of radionuclides before sources have decayed.

Mined repositories, comprising caverns or tunnels with varying types of EBSs, are being developed in many countries that have nuclear power industry wastes to manage [11, 46, 47]. They are designed to contain long-lived low and intermediate level waste, HLW and SFW. The contain­ment provided by all such repositories is expected to be adequate for the disposal of all types of RAW provided that legal and regulatory require­ments on repository inventory permit (some countries have strict con­straints on the types of waste that can be placed in specific repositories which are purely legal and unconnected with safety and performance). In addition, disused deep mines and/or caverns could be considered for geo­logical disposal [46, 47] .