Containment and Containerization

Containment and Containerization

Lecture Notes

I.Introduction

When the activity in waste is relatively low and half-lives are less than about 30years, near surface disposal will often be suitable. Some waste, however, is too radioactive and too long-lived for near surface disposal. Such waste requires higher levels of isolation by containerization and containment than can be provided by near surface facilities.

A site should be identified that provides favourable conditions for containment and isolation of the waste from the biosphere and for preservation of the engineered barriers (e.g., with low groundwater flow and a benign geochemical environment).

Waste of higher radiotoxicity, which may include some disused radioactive sources, can be surrounded by an encapsulation matrix and placed in durable containers. The purpose of the encapsulation matrix is to contain the radionuclides in the waste through a combination of physical and chemical functions that are effective for hundreds or even thousands of years.

Other engineered barriers, such as a borehole backfill, may allow this containment period to be extended even further, but complete containment of all radionuclides for all time cannot be expected. Containment of radionuclides is also provided by the natural barriers by means of geochemical and physicochemical retention processes that lead to retardation of the transport of radionuclides in the geosphere. Evidence from natural analogues indicates that these processes can be effective over very long timescales.

II.Scope

This lecture focuses on containment and containerization methods to ensure effective isolation for hundreds or even thousands of years. A well-designed and well-located borehole disposal facility should provide reasonable assurance that radiological impacts in the post-closure period will be low both in absolute terms and in comparison with any other waste management options that are currently available at reasonable cost.

During the early development of the borehole disposal facility, a number of site, design, and operational options will be available. Choices should be made with a view to striking an optimum balance of operability, containment, and isolation within a reasonable financial cost. The multiple safety function approach should be utilized so that the safety of the facility does not depend unduly on a single barrier or a single chemical or physical property. Safety assessment should be used to examine the various design options: first, to see whether compliance with the regulatory constraints is achievable;and second, to help deliver the constrained optimization.

III.Concept

Containment

The borehole disposal facility should be designed to take account of the characteristics offered by the site, to optimize protection and to keep doses within the dose and/or risk constraints. The borehole disposal facility should then be constructed, operated, and closed according to the assessed design so that the assumed safety characteristics of both the engineered and the natural barriers are realized.

The requirements concerning containment established in Ref. [1] call for the engineered barriers, which include the waste form and packaging, to be designed and the natural barriers to be selected to provide containment of the radionuclides in the waste, especially during the initial period when the level of activity is most intense and radioactive decay can significantly reduce the hazard posed by the waste.

This will allow the majority of shorter lived radionuclides to decay in situ. At the same time, release of gaseous radionuclides and a small fraction of some other highly mobile species may be inevitable from waste package of some types, but generally these radionuclides present relatively minor radiological hazards. In any event, the safety assessment should demonstrate that doses and risks arising from such releases fall within the regulatory constraints.

A distinguishing feature of borehole disposal is that it is not limited to the depth ranges considered for near surface disposal (metres to tens of metres) or geological disposal (hundreds of metres). On the contrary, it may be relatively straightforward and cost effective to select an appropriate geological horizon (and therefore a suitable hydrogeological condition) for the disposal, with due consideration given to the containment of radionuclides and their isolation from humans. It is envisaged that the appropriate depth would lie in an intermediate depth range (e.g., 30 m to a few hundred metres), between near surface and deep geological disposal. Figure 1 shows some typical components of a borehole disposal system.

FIG. 1. Scheme of some possible components of a borehole disposal system

Isolation

While containment refers primarily to the radionuclides in the waste, isolation is more concerned with the waste itself and the need to keep this potentially dangerous material away from humans, human influence, resources used by humans, and the biosphere for as long as it remains a significant hazard. The isolation period of ‘several thousand years’ mentioned in the requirements concerning containment established in Ref. [1] would not apply if the radionuclides in the waste were short lived. In choosing sites, consideration should be given to erosion, tectonic uplift, and landslip that might cause the waste to be brought close to the surface over the assessment period. One of the aims of isolation is to prevent human intrusion, which could affect the subsequent isolation of the waste and containment of the radionuclides within it. It is clear that isolation is also important in promoting security. While human intrusion is inherently unpredictable, some actions can be taken at the design stage to lessen both its probability and its consequences. If possible, for instance, borehole disposal facilities should be located away from known underground mineral and water resources. In general, disposal at greater depth should improve security and should help to reduce both the probability and the consequences of human intrusion.

In the absence of institutional control, a depth of 30 m should be considered the minimum necessary to achieve waste isolation. This should therefore be the minimum depth required for waste that might constitute a security risk. However, for waste that would otherwise be eligible for near surface disposal and for short lived radionuclides, where the waste may no longer constitute a hazard after, perhaps, one hundred years, disposal at a shallower depth together with institutional control could be an option. Engineered anti-intrusion barriers that are mechanically strong and heavy may also be useful in enhancing isolation. For a small scale borehole disposal facility, the resources needed for institutional control could be reduced by locating the boreholes at a site with an existing security infrastructure;for example, at an existing nuclear facility.

For waste placed deeper than 30 m, isolation is primarily provided by the geosphere and the main factors to be considered in determining a depth that will provide an appropriate level of isolation are the rate of surface erosion, the timescale of the assessment, and the depth of any permafrost. Of course, isolation is not the only issue to be considered when determining borehole depth: the influence of the host geological environment on containment should also be considered.

To provide confidence in long-term safety, a waste disposal system should employ a number of complementary engineered and natural barriers. Often, these barriers will be effective over different timescales and will provide a number of safety functions. Depending on the hazards associated with the waste, the barriers may vary in number and complexity.

They may include, for instance:

A waste container made of a corrosion resistant material that gives a container lifetime of about a thousand years.
A cement based backfill placed between the container and the borehole casing to create high pH conditions that limit solubility and promote sorption and so provide chemical containment for thousands of years.
A location where the rate of groundwater movement and the degree of radionuclide sorption onto the surrounding rocks together ensure that the radionuclides would take many thousands of years to migrate to the biosphere.

Although the safety of a borehole disposal facility will ultimately be judged by global measures of the total system performance, these barriers should not be unduly dependent on each other. So, for instance, in the example just outlined, the container lifetime may be extended by the high pH conditions provided by water leaching from the cement backfill, and the longevity of the backfill will be assisted by low groundwater flow. However, a groundwater flow that is higher than expected should not result in rapid corrosion of the container and the release of its contents. Similarly, failure of the cement to provide the expected high pH conditions should not lead to failure of the container or a more rapid migration of radionuclides through the surrounding rocks. These unwanted possibilities could be prevented by using a container material that shows adequate corrosion resistance over a range of pH conditions and sufficient cement backfill to provide long-lived high pH conditions even if the groundwater flow were at the top end of the range of possibilities. In this way, a worse than expected performance from one of the barriers would not lead to the failure of the entire system.

Multiple barriers and multiple safety functions should be used to enhance both safety and confidence in safety by ensuring that the overall performance of the disposal system is not unduly dependent on a single barrier or function. This should provide reasonable assurance that, if a barrier does not perform as expected, then a sufficient margin of safety remains (see the requirements for multiple safety functions established in Ref. [1]).

The various components of the engineered and natural systems also need to be complementary. Examples of non-complementary components are:

The use of ordinary Portland cement when the surrounding groundwater or geology has high levels of sulphate (common in some types of clay).
The use of swelling clays in highly saline environments or in groundwater with high levels of potassium.

Choice of engineered barriers

The engineered barriers can provide a significant degree of containment for the radionuclides in the waste. The use of corrosion resistant materials should allow the engineered barriers to be sufficiently long lived to make a useful contribution to safety. Thus, the safety case should be able to take some credit for a period of containment within the package itself. The engineered barriers may include (see, for example, Fig. 1):

The original casing (for disused sealed sources).
Welded metal (e.g., plain carbon or stainless steel) capsules for some small volume waste (e.g., radium sources).
A metal (e.g., plain carbon or stainless steel) waste container.
An encapsulation matrix (e.g., cement grout, bentonite, or lead) within which radioactive waste (e.g., radium sources) may be embedded, creating the waste form within the container.
Borehole backfill (e.g., cement grout) surrounding the waste packages.
Metal or plastic borehole casing to support borehole walls during drilling or emplacement operations; following waste emplacement, it may be beneficial to remove the casing above the disposal zone.
Casing seal to fill any voids between the casing and the borehole.
Borehole seal — a clay or cement plug several metres long placed in the borehole above the disposal zone (which can sometimes be complemented by a plug at the bottom of the disposal zone).

The effectiveness of, and confidence in the effectiveness of, the engineered barriers will be greatest when they employ a range of chemical and physical properties to contain the radionuclides. So, for instance, whereas the role of the waste container is primarily one of physical containment, a cement based encapsulation matrix can provide some chemical containment by reducing radionuclide solubility and providing surfaces that radionuclides can sorb onto. An important consideration in the choice of engineered barriers is their compatibility with the surrounding geochemical environment and their durability and integrity over the period of time for which the waste remains hazardous.

FIG. 2. Illustrative section through a borehole used for disposal

Conditioning

What is conditioning?

In the BDC system, conditioning is the placing of one or more DSRS within astainless steel capsule that is subsequently seal welded.

Conditioning provides:

Conversion of the DSRS to a special form radioactive material (in the sense of the IAEAtransport regulations).
Proper documentation for the DSRS.
Physical protection from damage and radionuclide release.
A standard sized package that is more easily handled.

All of this greatly facilitates, and improves the safety of, subsequent transportation, storage, anddisposal.

For Category 3 sources and below (e.g., Ra-226 brachytherapy sources), characterization andconditioning may be performed in the lightly shielded conditioning unit. The operation of theconditioning facility, including procedures for welding, leak testing, and radiologicalprotection, is described below and in Section 6 of Ref. [2]. More powerful gamma emitters(up to 40 TBq Co-60 equivalent) will require the use of the BDC hot cell.

FIG. 3. Conditioning unit located at Pelindaba, South Africa

General requirements for conditioning

DSRS accepted for conditioning must meet the relevant waste acceptance criteria. The following guidelines should be observed when setting up a supervised area where sourcesare manipulated:

Continuous control of contamination must be carried out. Storage containers should be subjected to contamination control before, during, and after transfer of this content to another the conditioning capsules.
The transfer zones should be covered with double sheets of polyethylene wherever contamination may be expected to take place. This should include covering of the lead bricks (Fig. 4) used for shielding. Confirmatory air sampling will be performed in this area when leaking sources are handled.
The operational area should be made as small as practical, so that all operations may be carried out within a limited area.
Suitable ventilation should be maintained, extracting to the outside air via HEPA filters. Ventilation should be controlled in such a way so as to avoid spread of loose contamination. This normally is the transfer area where leaking sources could be transferred into a stainless steel capsule.
The operation should be optimized to limit the number of manipulations of the sources for various operations.
Special attention must be given to the receiving area where the incoming transport packages are handled.

FIG. 4. Lead brick shielding standing in front of the leak testing unit.

The areas should be subjected to the following:

All containers within the supervised area should be checked for contamination before the campaign and at regular intervals during it.
Radiation levels within the supervised area should be established prior to the intended work (e.g., encapsulation operation). The radiation level in the transfer zone should be checked before and after the transfer of the contents of a transport or storage container to a capsule.

FIG. 5. Capsule welding in progress (note the cylindrical lead shield).

Permissible contents of a capsule

The amount of radioactivity that is permissible to place inside a capsule is mainly determined by the operational safety assessment and is defined by the waste acceptance conditions.

Operations in the Conditioning Unit

The supervised area for the conditioning operation should consist of the following five zones (Fig. 6):

Receiving zone.
Transfer zone.
Welding zone.
Leak testing zone (Fig. 6).
Container filling/storage zone.

The receiving zone is the area where DSRS in their transport or storage containers(which may include the original device) are received. At this point, the team should know the total inventory of the received package, as much as possible about its content (e.g., number of sources, their activities, and geometry), external contamination, if any, and any other useful information (e.g., ease or difficulty of opening of the shield, approximate weight of the shield). During the removal of the DSRS, the workers will wear protective clothing and dust masks in case the source is leaking. The masks will also contain activated carbon to reduce exposure to Rn-222 that may emanate from Ra-226 sources. The extraction hood will further limit the potential inhalation dose to the workers.

Removal of aDSRS from a device will usually take place behind the lead shield. In some cases, however, this may not be possible, and here the radiation protection officer (RPO) will advise on the need for temporary shielding. In any event, dismantling will stop when the component that contains the source is small enough to fit into the capsule.

FIG. 6. Conditioning and containerization facility situated adjacent to the BDC facility

The transfer zone is the area where the capsule is transferred to the disposal container. Prior to thetransfer, a calibrated radiation monitor allows the level of radioactivity to be checked. Shielding needs to be arranged so that the monitor readings are not affected by radiation from other SRS that may be present in the conditioning unit. Workers are protected using a lead brick wall (see Figs 4-6).