DISPOSAL OF RADIOACTIVE WASTE

DISPOSAL OF RADIOACTIVE WASTE

1.         INTRODUCTION

1.1    General

Radioactive waste arises from the generation of electricity in nuclear power plants, from nuclear fuel cycle operations and from activities in which radioactive material is used. It also arises from activities and processes in which radioactive material of natural origin become concentrated in waste material and safety needs to be considered in its management. Radioactive waste can be generated in a wide range of activities varying o activities in hospitals to nuclear power plants to mines and minerals processing facilities. There is a variety of alternatives for treatment and conditioning of the wastes prior to disposal. Likewise, there are a number of alternatives for safe disposal of these wastes, ranging from geological disposal to near surface disposal and direct discharge to the environment and disposal in space. To simplify their management, a number of schemes have evolved for classifying radioactive waste according to the physical, chemical and radiological properties of significance to those facilities managing this waste. These schemes have led to a variety of terminologies, differing from country to country and even between facilities in the same country.

The properties of radioactive waste are likewise variable, not only in terms of radioactive content and activity concentration but also in terms of physical and chemical properties. Its rate of generation is also variable. Common characteristics of all radioactive waste is its potential to present a hazard to people and the environment, and it must therefore be managed so as to reduce any associated risks to acceptable levels. The potential hazard can range from large to trivial; a variation reflected in management and disposal options necessary for various types of waste.

The safety principles are applied in all activities for radioactive waste management are set out as the requirements for radiation protection are set out in the International basic Safety standards for protection against Ionizing Radiation and for the safety of radiation sources. The preferred strategy for the management of all radioactive waste is to contain it (i.e. to confine the radionuclides to within the waste matrix, the packing and the disposal facility) and to isolate it from the accessible biosphere.

Radioactive waste may arise initially in various gaseous, liquid and solid forms. In waste management activities, the waste is generally processed to produce stable and solid forms, and reduced in volume and immobilized, as far as practicable, to facilitate their storage, transport and disposal. Various gaseous or liquid effluents that may result from this processing may be discharged to the environment provided that they meet the conditions for authorized discharge.

1.2     Concepts relating to disposal (and storage) of radioactive waste

The term ‘disposal’ refers to the emplacement of radioactive waste into a facility or a location with no intention of retrieving the waste. Disposal options are designed to contain the waste by means of passive engineered and natural features and isolate it from the accessible biosphere to the extent necessitated by the associated hazard. The term disposal implies that retrieval is not intended; it does not mean that retrieval is not possible.

By contrast, the term ‘storage’ refers to the retention of radioactive waste in a facility or a location with the intention of retrieving the waste. Both options, disposal and storage, are designed to contain waste and to isolate it from the accessible biosphere to the extent necessary. The important difference is that storage is a temporary measure following which some future action is planned. This may include further conditioning or packaging of the waste, and ultimately its disposal.

A number of design options for disposal facilities have been developed and various types of disposal facilities have been constructed in many States and are in operation. These
design options have different degrees of containment and isolation capability appropriate to the radioactive waste that they will receive. The specific aims of disposal are:

·         To contain the waste

·         To isolate the waste from accessible biosphere and to substantially reduce the likelihood of all possible consequences of inadvertent human intrusion into the waste.

·         To inhibit, reduce and delay the migration of radionuclides at any time from the waste of accessible biosphere.

·         To ensure that the amounts of radionuclides reaching the accessible biosphere due to any migration from the disposal facility are such that possible radiological consequences are acceptably low at all times.

Within any State or region a number of disposal facilities of different designs may be required in order to accommodate radioactive to of various types. The following disposal options have been adopted in one or more States, corresponding to recognized classes of radioactive waste. The classification of radioactive waste is discussed below:

·         Specific landfill disposal: disposal in facility similar to a conventional landfill facility for industrial refuse but it may incorporate measures to cover the waste. Such a facility may be designed as a disposal facility for very low level radioactive waste (VLLW) with low concentrations or quantities of radioactive content. Typical waste disposed of in a facility of this type may include soil and rubble arising from decommissioning activities.

·         Near surface disposal: disposal in a facility consisting of engineered trenches or vaults constructed on the ground surface or up to a few tens of meters below ground level. Such a facility may be designated as a disposal facility for low level radioactive waste (LLW).

·          Disposal of intermediate level waste: Depending on its characteristics, intermediate level radioactive waste (ILW) can be disposed of in facilities of different types. Disposal could be by emplacement in a facility constructed in caverns, vaults or silos at least a few tens of meters below ground level and up to a few hundred meters below ground level. It could include purpose built facilities and facilities developed in or from existing mines. It could also include facilities developed by drift mining into mountainsides or hillsides, in which case the overlying cover could be more than 100 meters deep.

·         Geological disposal: disposal in a facility constructed in tunnels, vaults or silos in a particular geological formation (e.g. in terms of its long term stability and its hydro geological properties) at least a few hundred meters below ground level. Such a facility could be designed to accept high level radioactive waste (HLW), including spent fuel if it is to be treated as waste. However, with appropriate design a geological disposal facility could receive radioactive waste of all types.

·         Borehole disposal: disposal in a facility consisting of an array of boreholes, or a single borehole, which may be between a few tens of meters up to a few hundreds of meters deep. Such a borehole disposal facility is designed for the disposal of only relatively small volumes of waste, in particular disused sealed radioactive sources. A design option of very deep boreholes, several kilometers deep, has been examined for the disposal of solid high level waste and spent fuel, but this option has not been adopted for a disposal facility in any State.

·         Disposal of mining and minerals processing waste: disposal usually on or near the ground surface, but the manner in which and the large volumes in which the waste arises, its physico-chemical form and its content of long lived radionuclides of natural origin distinguish it from other radioactive waste. The waste is generally stabilized in situ and covered with various layers of rock and soil.

1.3     Development of disposal facilities

The development (i.e. the site selection and evaluation, design and construction) of most types of disposal facility is likely to take place over extended periods of time. The period over which disposal facilities will be operated prior to closure will in most cases also extend over decades. Different activities will be conducted this period of development, such as site evaluation, design and construction, with decisions being made to proceed to the next set of activities or the next step in the development of the facility.

Such a step by step approach to the development enables: the ordered accumulation and assessment of the necessary scientific and technical data; the evaluation of possible sites; the development of disposal concepts; iterative studies for design development and safety assessment with progressively improving data; technical and regulatory reviews; public consultation; and politic decisions.

The step by step approach to development, together with the consideration of a range of options for the design and operational management of a disposal facility, is expected to provide flexibility for responding to new technical information and advances in waste management and material technologies. It also enables social, economic and political aspects of the disposal facility to be addressed, to ensure that all reasonable measures have been taken to further prevent, inhibit or delay releases to the environment.

It is convenient to identify three periods associated with the development, operation and closure of a disposal facility: the pre-operational period, the operational period and the post-closure period. Various activities will take place in these periods and some may be undertaken to varying degrees throughout part or all of the lifetime of the facility.

·         The pre-operational period includes concept definition, site evaluation (selection, verification and confirmation), safety assessment, and design studies. It also includes the development of those aspects of the safety case for safety in operation and after closure that are required in order to set the conditions of authorization, to obtain the authorization and to proceed with the construction of the disposal facility and the initial operational activities. The monitoring and testing programs that are needed to inform operational management decisions are put in place.

·         The operational period begins when waste is first received at the facility. From this time, radiation exposures may occur as a result of waste management activities, and these are subject to control in accordance with the requirements for protection and safety. Monitoring, surveillance and testing programs continue to inform operational management decisions, and to provide the basis for decisions concerning the closure of the facility or parts of it. Safety assessments for the period of operation and after closure and the safety case are updated as necessary to reflect actual experience and increasing knowledge. In the operational period, construction activities may take place at the same time as waste emplacement in and closure of other parts of the facility. This period may include activities for waste retrieval — if considered necessary — prior to closure, activities following the completion of waste emplacement, and the final closure and sealing of the facility.

·         The post-closure period begins at the time when all the engineered containment and isolation features have been put in place, operational buildings and supporting services have been decommissioned, and the facility is in its final configuration. After its closure, the safety of the disposal facility is provided for by means of passive features inherent in the characteristics of the site and the facility and characteristics of the waste packages, together with certain institutional controls, particularly for near surface facilities. Such institutional controls are put in place to prevent intrusion into facilities and to confirm that the disposal system is performing as expected by means of monitoring and surveillance. Monitoring may also be carried out to provide public assurance. The license will be terminated after the period of active institutional control when all the necessary technical, legal and financial requirements have been fulfilled.

2.     APPROACHES TO RADIOACTIVE WASTE CLASSIFICATION

Classification systems for radioactive waste may be derived from different points of view, such as safety related aspects, process engineering demands or regulatory issues. Classification of radioactive waste may be helpful at any stage between the arising of the raw waste and its conditioning, interim storage, transportation and disposal.

Therefore, classification of radioactive waste will serve many purposes. It will help:

·        at the conceptual level

—     in devising waste management strategies;

—     in planning and designing waste management facilities;

—     in designating radioactive waste to a particular conditioning technique or disposal facility;

·        at the operational level

—     by defining operational activities and in organizing the work with waste;

—     by giving a broad indication of the potential hazards involved with the various types of radioactive waste;

—     by facilitating record keeping;

·        for communication

—    by providing words or acronyms universally understood which improve communication among experts from different countries, and between experts, generators and managers of radioactive waste, regulators and the public.

2.1     Qualitative classification

There already exist 'natural' classification systems, e.g. grouping the radioactive wastes in terms of their origin. A great many activities involving the use of radionuclides and nuclear power generation result in generation of radioactive waste. Such activities include all steps in the nuclear fuel cycle (i.e. the activities associated with the generation of nuclear power) as well as other non-fuel-cycle activities. Radioactive waste may also be generated outside the nuclear activities by the (mostly large scale) processing of raw materials containing naturally occurring radionuclides which in some cases may be considered as being radioactive. Examples include phosphate ore processing and oil or gas exploration. The radionuclide content of radioactive waste from fuel cycle activities greatly exceeds the radionuclide content of materials from non-fuel cycle activities.

Another 'natural' classification system is the differentiation of radioactive wastes according to the physical state, i.e. solid, liquid, and gaseous. This system stems from the process engineering needs for the treatment of the different radioactive waste streams and is often refined corresponding to individual radioactive waste treatment systems. A classification system of this type is mostly specific to individual facilities and follows their technical needs and possibilities. It may, however, incorporate safety considerations such as the radiation protection necessary for radioactive waste classes with higher radioactivity content.

A widely used qualitative classification system separates radioactive waste into three classes: low level waste (LLW), intermediate level waste (ILW) and high level waste (HLW). A further distinction is made between short lived and long lived Exempt waste.

TABLE NO. 1:

     IMPORTANT PROPERTIES OF RADIOACTIVEWASTE USED AS             CRITERIA FORCLASSIFICATION

·        Origin

·        Criticality

·        Radiological properties:

o   Half-life

o   Heat generation

o   Intensity of penetrating radiation

o   Activity and concentration of radionuclides

o   Surface contamination

o   Dose factors of relevant radionuclides

·        Other physical properties:

o   physical state (solid, liquid or gaseous)

o   size and weight

o   compactability

o   dispersibility

o   volatility

o   solubility, miscibility

·        Chemical properties:

o   potential chemical hazard

o   corrosion resistance/corrosiveness

o   organic content

o   combustibility

o   reactivity

o   gas generation

o   sorption of radionuclides

·        Biological properties:

o   potential biological hazards

2.1.2  Military and Civilian Wastes

The first nuclear reactors were those built during World War II to produce plutonium for weapons. In order to extract the plutonium, it is necessary to reprocess the spent fuel, first converting it to liquid form. The residue remaining after the plutonium and uranium (and sometimes other elements) are extracted constitutes the reprocessed wastes. Reprocessing increases the volume of the residue and puts it in a form that can more readily escape into the environment. This residue was originally stored as liquids in large underground, single-walled tanks.

Given the pressures of wartime development, there was no well-engineered, long-term plan for the permanent disposal or storage of these wastes. Weapons production continued and increased after World War II, with large programs at the DOE’s facilities at Hanford (Washington) and Savannah River (South Carolina), but disposal plans were still not developed in a timely fashion. As a result, there were mishaps, including large leaks from some tanks at Hanford during the 1970s, and concerns arose about possible further leaks and conceivable chemical explosions.

These wastes constitute the bulk of the military wastes. Nuclear reactors used on naval vessels constitute a second, smaller source of the military wastes, but the radioactivity levels and physical volumes involved are considerably less than those from weapons production

Civilian or commercial wastes are those produced by reactors built for commercial electricity generation. The amount of radioactivity produced in this manner to date is much greater than that for the military wastes, because more reactors have been involved, operating over longer total periods. The volume is less, however, because the wastes have remained as solid fuel rods. Overall, they are easier to handle than the reprocessed military wastes, being in compact solid form. The focus here will be on commercial wastes, because their successful disposal is crucial to the future of nuclear power.

2.1.3  High level waste

(i)                 The highly radioactive liquid, containing mainly fission products, as well as some actinides, which is separated during chemical reprocessing of irradiated fuel (aqueous waste from the first solvent extraction cycle and those waste streams combined with it).

(ii)               Any other waste with radioactivity levels intense enough to generate significant quantities of heat by the radioactive decay process,

(iii)              Spent reactor fuel, if it is declared a waste.

2.1.4  Intermediate level waste (medium level waste)

Waste which, because of its radionuclide content requires shielding but needs little or no provision for heat dissipation during its handling and transportation.

2.1.5  Low level waste

Waste which, because of its low radionuclide content, does not require shielding during normal handling and transportation. Such a condition is shown in figure 1.

Figure 1: Storage of low level waste (LLW) at ANSTO. LLW contains low levels of radioactivity, and therefore shielding is not required to protect workers during storage or transportation.

2.2     WASTE CLASSES

The revised classification system is presented in Fig. 2. The principal waste classes include exempt waste, low and intermediate level waste, which may be subdivided into short lived and long lived waste, and high level waste.

2.2.1  Short lived waste

Short-lived waste refers to radioactive waste which will decay to an activity level which is considered to be acceptably low from a radiological viewpoint, within a time period during which administrative controls can be expected to last. (Such waste can be determined by radiological performance assessment of the storage or disposal system chosen.)

2.2.2  Long lived waste

Long lived waste is radioactive waste that will not decay to an acceptable activity level during the time which administrative controls can be expected to last. Alpha bearing waste is radioactive waste containing one or more alpha emitting radionuclides, usually actinides, in quantities above acceptable limits established by the national regulatory body.

Considering Fig. 2 vertically, radioactivity levels range from negligible to very high concentrations of radionuclides. As the level rises, there is an increased need to isolate the waste from the biosphere; suitable disposal options may range from simple and conventional methods to geological isolation. In addition, there is an increased need to consider shielding from radiation, and the generation of heat from radioactive decay.

Considering the figure horizontally, decay periods range from short (seconds)to very long time spans (millions of years) and similarly radioactive wastes range from those containing minor quantities of long lived radionuclides to those containing significant quantities thereof. As appropriate, radioactive waste may be (1) stored for decay and then exempted, (2) disposed of in near surface facilities, or (3) isolated from the biosphere in deep geological formations. This situation is reflected as two subclasses of radioactive waste distinguish short lived and long lived low and intermediate level waste.

Figure 2. Revised waste classification system

Such a distinction between short lived and long lived low and intermediate level waste can be of substantial benefit because the radiological hazards associated with short lived radionuclides can be significantly reduced over a few hundred years by radioactive decay. Different time periods for the isolation of short lived and long lived low and intermediate level waste will be necessary. Activity limitations for a given disposal facility will in particular depend on the radiological, chemical, physical and biological properties of individual radionuclides. It can by no means be implied that long lived radionuclides are inherently more hazardous than short lived radionuclides.

2.2.3  Exempt waste (EW)

Exempt waste (EW) contains so little radioactive material that it cannot be considered 'radioactive' and might be exempted from nuclear regulatory control. That is to say, although still radioactive from a physical point of view, this waste may be safely disposed of, applying conventional techniques and systems, without specifically considering its radioactive properties.

Below, a more detailed discussion is presented for each of the revised waste classes. Boundary levels between classes are presented as orders of magnitude and typical characteristics of waste classes are summarized in Table 2.

Application of a classification system for the management of radioactive waste implies an adequate separation of wastes generated. A decision chart for the segregation of radioactive and exempt waste is presented in Fig. 2.

Figure 3: decision chart for segregation of radioactive and exempt waste.

2.3     Low and intermediate level waste (LILW)

Low level waste has been defined in the past to mean radioactive waste that does not require shielding during normal handling and transportation. Radioactive waste which required shielding but needed little or no provision for heat dissipation was classified as intermediate level waste. A contact dose rate of  2 mSv/h was generally used to distinguish between the two classes.

This distinction appears of secondary importance in the present context. Classification should be related to individual radionuclides, taking the various exposures and exposure pathways into account, such as inhalation (e.g. in the case of an incident) and ingestion (e.g. in the case of long term releases in the post operational period of a repository). Thus, low and intermediate level waste may be subdivided into short lived and long lived waste. Additional considerations which must be taken into account in managing low and intermediate level waste are presented subsequently under 'Additional Considerations'.

2.3.1  (a) Short lived waste (LILW-SL)

Short lived low and intermediate level waste (LILW-SL) contains low concentrations of long lived radionuclides. The possible hazard represented by the waste can often be significantly reduced by administratively controlling waste as part of storage or after disposal. Although the waste may contain high concentrations of short lived radionuclides, significant radioactive decay occurs during the period of institutional control. Concentrations of long lived radionuclides that will not decay significantly during the period of institutional control are controlled to low levels consistent with the radio toxicity of the radionuclides and requirements set forth by national authorities.

Because LILW-SL may be generated in a wide range of concentrations, and may contain a wide range of radionuclides, there may be a range of acceptable disposal methods. The waste form or packaging may also be important for management of this waste. Depending upon safety analyses and national practices, these methods may range from simple surface landfills, to engineered surface facilities, and to disposal at varying depths, typically a few tens of meters or in deep geological formations if a co-disposal of short and long lived waste is anticipated. National practices may impose varying levels of isolation depending upon the hazard represented by different classes of radioactive waste.

2.3.2  (b) Long lived waste (LILW-LL)

 Long lived low and intermediate level waste (LILW-LL) contains long lived radionuclides in quantities that need a high degree of isolation from the biosphere. This is typically provided by disposal in geological formations at a depth of several hundred meters. The boundary between short lived and long lived waste cannot be specified in a universal manner with respect to concentration levels for radioactive waste disposal, because allowable levels will depend on the actual radioactive waste management option and the properties of individual radionuclides. However, in current practice with near surface disposal in various countries, activity concentration is limited to 4000 Bq/g of long lived alpha emitters in individual radioactive waste packages, thus characterizing long lived waste which is planned to be disposed of in geological formations. This level has been determined based on analyses for which members of the public are assumed to access inadvertently a near surface repository after an active institutional control period, and perform typical construction activities (e.g. constructing a house or a road).

Applying this classification boundary, consideration should also be given to accumulation and distribution of long lived radionuclides within a near surface repository and to possible long term exposure pathways. Therefore, restrictions on activity concentrations for long lived radionuclides in individual waste packages maybe complemented by restrictions on average activity levels or by simple operational techniques such as selective emplacement of higher activity waste packages within a disposal facility. An average limit of about 400 Bq/g for long lived alpha emitters in waste packages has been adopted by some countries for near surface disposal facilities.

In applying the classification system, attention should also be given to inventories of long lived radionuclides in a repository that emit beta or gamma radiation. For radionuclides such as 129I or "Tc, allowable quantities or average concentrations within a repository depend strongly on site specific conditions. For this reason, national authorities may establish limits for long lived beta and gamma emitting radionuclides based on analyses of specific disposal facilities.

2.4     High level waste (HLW)

The high level waste (HLW) class largely retains the definition of the existing classification system. This waste contains large concentrations both of short and long-lived radionuclides, so that a high degree of isolation from the biosphere, usually via geological disposal, is needed to ensure disposal safety. It generates significant quantities of heat from radioactive decay, and normally continues to generate heat for several centuries.

 An exact boundary level is difficult to quantify without precise planning data for individual facilities. Specific activities for these waste forms are dependent on many parameters, such as the type of radionuclide, the decay period and the conditioning techniques. Typical activity levels are in the range of to TBq/m3, corresponding to a heat generation rate of about 2 to 20 kW/m3for decay periods of up to about ten years after discharge of spent fuel from a reactor. From this range, the lower value of about 2 kW/m3 is considered reasonable to distinguish HLW from other radioactive waste classes, based on the levels of decay heat emitted by HLW such as those from processing spent fuels.

2.4.1  ADDITIONAL CONSIDERATIONS

A number of additional important factors should be considered when addressing specific types or properties of radioactive waste.

2.4.2  Waste containing long lived natural radionuclides

Many countries must address the disposal of very large quantities of waste containing long lived natural radionuclides. Such waste typically contains natural radionuclide like uranium, thorium, and radium and is frequently generated from uranium/thorium mining and milling or similar activities. It may also include waste from decommissioning of facilities, where other isotopes may also be present. The characteristics of these wastes are sufficiently different from other wastes that they may require an individual regulatory approach.

Although these wastes do contain long lived radionuclides, their concentrations are generally sufficiently low that either they can be exempted or disposal options similar to those for short lived waste may be considered, depending on safety analyses.

2.4.3  Heat generation

Although heat generation is a characteristic of high level radioactive waste, other radioactive wastes may also generate heat, although at lower levels. Heat generations dependent upon the type and content of radionuclides (half-life, decay energy, etc.). Furthermore, the heat removal situation is highly important (thermal conductivity, storage geometry, ventilation, etc.). Therefore, heat generation cannot be defined by a single value. The relevance of heat generation can vary by several orders of magnitude depending on the influencing parameters and the temperature limitations. Management of decay heat should be considered in a repository if the thermal power of waste packages reaches several W/m3. Especially in the case of long lived waste, more restrictive values may apply.

2.5     Liquid and gaseous waste

The treatment of liquid waste (which may contain particulate solids) and gaseous waste (which may contain aerosols) aims at separating the radionuclides from the liquid or gaseous phase and concentrating them in a solid waste form. The separation is pursued until the residual concentration or total amount of radionuclides in the liquid or gaseous phase is below limits set by the regulatory body for the discharge of liquid or gaseous waste from a nuclear facility as an effluent. Treatment may include a storage period for radioactive decay.

Liquid and gaseous radioactive waste exceeding discharge limits set by national authorities should be conditioned for storage, transport and disposal. Only following sound safety analysis should radioactive waste in liquid or gaseous form be transported off the site or disposed of in terrestrial repositories in their original forms. Storage for decay at the facility of their origin may be considered as part of the conditioning process.

The classification of liquid and gaseous radioactive waste may be based on the different types of treatment that can be used, and on potential radiological, chemical and biological hazards. When solidified or conditioned for disposal these wastes fall under one of the solid radioactive waste classes.

3.         Hazard Measures for Nuclear Wastes

3.1     Exposures from Direct Contact

The spent fuel assemblies are placed, with remote handling, into cooling pools. They eventually are to be transferred from the cooling pools into protective canisters which are placed in heavy casks, again with remote handling. The canisters and casks provide substantial shielding—essentially as much as one wants if a price is paid in weight. Therefore, radiation exposure from close contact with spent fuel assemblies is not a critical safety issue, assuming proper handling during the operations that precede their placement in the casks.

The high radiation levels from unshielded spent fuel provide important protection against theft by people who do not have the elaborate equipment and facilities required to handle the fuel assemblies. However, in considering the radiation hazards from nuclear wastes, the usual focus is not on exposures of terrorists or others from direct contact. It is on the possible exposure of the general public many centuries hence, through the escape into the biosphere of radionuclides from waste repositories.

3.2     Hazards from Wastes in a Repository

The ultimate measure of the hazards created by nuclear wastes is the dose or spectrum of doses received by people who may ingest or inhale radionuclides that escape from the repository. Calculating this dose requires as a first step the evaluation of mechanisms by which the waste containers might be damaged, permitting the escape of radionuclides. Then, for each radionuclide, it is necessary to consider its amount in the repository as a function of time, the rate at which it would escape from damaged containers, its subsequent movement from the repository site to the biosphere, its pathways for entering the human body, and the dose resulting from its ingestion or inhalation.

4.         DOCUMENTATION OF RADIOACTIVE WASTE

Waste documentation has to contain all information which is required for repository planning and for providing a sound technical and administrative basis for waste acceptance at future repositories. It consists of

·         A waste type documentation which comprehensively covers the general features of the set of waste packages summarized under the heading of a "type" within a waste package type specification, taking into account the results of any prototype testing and issues from previous waste characterization programmes.

·         Individual waste package documentation dedicated to real package-specific data, including a waste package datasheet with the results of the quality control programme, a storage logbook and, if applicable, package-specific results from the waste characterization programmes.

The waste package type specification is a technical document in which the waste producer describes and specifies the following aspects for a set of similar waste packages:

·         Manufacturing conditions (conditioning procedures and plant)

·         Structure and properties of the package and its components (nominal values, band-widths, guarantees)

·         Quality control programme

·         Datasheet specimen for single packages

5.      Storage and Disposal of Nuclear Wastes

5.1     Material management and waste conditioning

There are (limited) possibilities of avoiding or at least minimizing radioactive waste. During planning or construction of a research facility, one should try to keep the amount of potentially irradiated mate­rials as low as possible and to use materials with a low tendency for build-up of critical safety relevant nuclides (e.g. steels with low Ni and/or Co impurities). The option of a free release of materials from controlled zones. In addition the recycling (e.g. copper wires) or reuse (e.g. use of activated concrete as shielding material in other facilities) within nuclear installations reduces the volume of radioactive or conventional waste to be disposed of.

The resulting radioactive waste has to be conditioned. For waste from nuclear research facilities (mostly large components), cementation in large containers is probably the most convenient strategy. Materials can be mixed as long as the waste acceptance criteria are met. (Large) components with voids can be used to incorporate small parts and then be cemented as a unit into a disposal container. Waste volume reduction by

• Compression
• Melting
• Incineration

They should be considered as a possible pre-treatment step.

5.2     Basic Steps and Activities in Radioactive Waste Management

Waste Generation occurs during the operational period and during the decommissioning of nuclear facilities. It can be in the form of solid, liquid or gaseous waste.

Pretreatment is the initial step that occurs just after generation. It consists of, for example, collection, segregation, chemical adjustment and decontamination and may include a period of interim storage. This step provides the best opportunity to segregate waste streams, based on similar methods of future management, and to isolate those wastes that are nonradioactive, or those materials which can be recycled.

Treatment involves changing the characteristics of the waste. Basic treatment concepts are volume reduction, radionuclide removal and change of composition. Typical treatment operations include: incineration or compaction of dry solid waste or organic liquid wastes (volume reduction); evaporation, filtration or ion exchange of liquid waste (radionuclide removal); and precipitation or flocculation of chemical species (change of composition).

Conditioning involves those operations that transform radioactive waste into a form suitable for handling, transportation, storage and disposal. These operations may include immobilization of radioactive waste, placing waste into containers and providing additional packaging. Common immobilization methods include solidification of LLW and ILW liquid radioactive waste, for example in cement, and vitrification of HLW in a glass matrix. Immobilized waste may be placed in steel drums or other engineered containers to create a waste package.

Storage of radioactive waste may take place between and within the basic radioactive waste management steps. Storage may be used to facilitate the next step or to act as a buffer between and within steps. Periods of storage may extend to many years until the waste is removed from the storage facility for further processing and disposal as applicable. Storage facilities may be co-located with a nuclear power plant or a licensed disposal facility, or may be separate entities. The intention of storage is to isolate the radioactive waste, provide environmental protection and facilitate control.

Retrieval involves the recovery of waste packages from storage either for inspection purposes, for subsequent disposal or further storage in new facilities. Storage facilities may be designed such that the original emplacement equipment may be operated in reverse in order to retrieve waste packages. Others may require the installation of retrieval equipment at the appropriate time.

Disposal consists of the authorized emplacement of packages of radioactive waste in a disposal facility, without the primary intention of retrieval and without a need for any further actions to ensure future safety. Disposal may also comprise of discharging radioactive waste (for example, liquid and gaseous effluent into the environment).

5.3     There are a number of activities that are carried out during the management of all types of waste. These are:

Minimization of waste is fundamental good practice in radioactive waste management. It should be considered during the design of facilities, and applied during all of the basic steps. Effective methods of minimizing the accumulation of radioactive waste include the clearance of waste that is below regulatory control, and the reuse or recycling of radioactive material.

Characterization of radioactive waste involves determining the physical, chemical and radiological properties. It may be carried out in association with several of the basic steps. It may be required for record keeping, acceptance of waste moving between steps and also to determine the best method of managing waste.

Segregation of radioactive waste involves accumulating together those wastes that have similar physical, chemical and radiological properties and that will be subject to similar methods or options for future management. It also avoids mixing together radioactive wastes that have different properties and different methods of future management. It is most effectively carried out during the early steps of radioactive waste management.

Transportation of radioactive waste may take place between and within the basic steps. The term transport generally refers to moving radioactive waste between nuclear sites, whereas transfer refers to moving radioactive waste within a nuclear site.

Figure 4.  The Basic Steps of Radioactive Waste Management

5.4                      Radioactive Waste Minimization

Radioactive waste is a product of many operations within the nuclear industry. Avoiding the creation of radioactive waste in the first instance and, secondly, minimizing the rate at which waste, which must be created, is produced is one of the foremost principles of good radioactive waste management.

The generation of radioactive waste shall be kept to the minimum practicable, in terms of both its activity and volume, by appropriate design measures and operating and decommissioning practices. This includes the selection and control of materials, recycle and reuse of materials, and the implementation of appropriate operating procedures. Emphasis should be placed on the segregation of different types of waste and materials to reduce the volume of radioactive waste and facilitate its management.

5.4.1  Waste minimization Practice

In general, measures to reduce radioactive waste production at source are more effective than measures taken after the waste has been created. Waste minimization is fundamental good practice, reduces hazards on site, reduces the potential impact on the environment, and in many cases is cost effective. Waste minimization includes the following practices (in some cases the practices reduce the accumulation of waste rather than it’s creation):

·         Avoidance of the production of secondary wastes;

·         Segregation of waste streams (by waste category, physical and

Chemical properties);

·         Preventing spread of contamination;

·         Recycling and reusing material;

·         Waste clearance;

·         Decontamination;

·         Volume reduction;

·         Disposal.

Expects the safety cases for all nuclear facilities to include a demonstration that the rate of production and accumulation of waste has been reduced so far as is reasonably practicable. This should include an optimization study of the activity in liquid and gaseous routine discharges, solid waste arising, occupational exposure and environmental impact.

5.4.2  Consideration during Design

Waste minimization should be considered at the design stage of a new plant, and when modifications are made to existing plant. The implications for waste generation should be taken into account in:

·         process selection;

·         plant layout;

·         choice of components and materials; and

·         decontaminable construction materials.

Similarly, good operating practices should be defined at the outset to limit the generation of secondary wastes (for example, use of reusable protective clothing and suitable packaging materials).

6.                Stages in Waste Handling

The main stages in nuclear waste handling are as follows:

·         Storage of spent fuel in cooling pools at the reactors.

·         Dry storage of spent fuel at reactor sites.

·         Reprocessing of spent fuel.

·         Interim storage of reprocessed waste or spent fuel at centralized facilities.

·         Permanent disposal of spent fuel, reprocessed waste, or residues of transmutation, by placement in repositories or by other means.

·         Transportation of spent fuel or reprocessed waste as it moves through the stages above.

6.1     Storage of Spent Fuel at Reactor Sites

Limitations of Cooling Pools

Originally, it was expected that the spent fuel from nuclear reactors would be held in cooling pools at the reactor sites for a brief time and that reprocessing would be carried out after about 150 days. Instead the fuel has remained at the reactor sites, and some pools have held fuel for over 20 years. The capacity of these pools is limited, and although no reactor has had its operation stopped by a shortage of cooling pool space, individual reactors have faced a severe squeeze. The capacity for storage of spent fuel assemblies at the reactor site can be increased to a modest extent by modifying the geometric arrangement of the assemblies in the cooling pool. A much larger expansion can be achieved by using dry storage.

6.2     Dry Storage of Spent Fuel at Reactor Sites (in situ)

Dry storage systems, the spent fuel rods are transferred to special casks when the total activity and the heat output are reduced enough for air cooling to suffice. This solves the problem of limited cooling pool capacity and is an option that can be implemented pending decisions on the establishment of centralized facilities. It also defers the contentious issue of transportation of nuclear wastes. The dry storage casks are cooled by natural convective airflow, without pumps. They must be licensed by the Nuclear Regulatory Commission, which by 2003 had approved 15 different cask designs

A diagram of the cask used at the Palisades facility is shown in Figure 5.

The fuel assemblies are transferred underwater in the reactor cooling pool to a steel canister (called a basket). The canister is then brought to a decontamination area, where it is pumped dry, filled with helium, and sealed with redundant welded lids. The sealed canister is placed in the storage cask, which consists of an outer cylinder with a 29-in.-thick concrete wall and a steel inner liner that is at least 1.5 in. thick. As shown in figure 5 and figure 6.

 The canister is cooled by natural air convection in a gap between the canister and the cask liner. In its first years, on-site storage often attracted significant legal and political challenges, although the first facility, at the Surry reactor, was installed in 1986 without any significant opposition. Legal efforts were made to stop on-site dry storage at the Palisades reactor in 1993, but these failed in the Michigan courts. However, in Minnesota, the state courts ruled that proposed dry cask storage at the Prairie Island nuclear plant required legislative approval, which was eventually granted in May 1994 under a plan that allowed a gradual installation of storage casks, but under conditions that appear to point toward a long-term phasing out of nuclear power in Minnesota

Figure 5:   Dry storage cask system used at the Palisades nuclear power plant. Top: canister     (basket) for holding fuel assemblies; bottom: storage cask for containing sealed canister.

Figure 6:  Dry cask storage on a concrete pad with cutaway schematic of a cask

An extensive survey of the interim storage issue, in a joint report by groups from Harvard University and Tokyo University, endorses dry cask storage in the following terms:

Dry storage technologies, especially dry casks, have been increasingly widely used in recent years. The combination of simplicity, modularity, and low operational costs and risks offered by dry cask storage systems make them highly attractive for many storage applications. This report does not put forth dry cask storage as the only acceptable option nor does it take a crisp position on the choice between on-site and centralized storage.

6.3     Reprocessing

Reprocessing is the technique of dissolving spent fuel rods and then removing the plutonium and uranium elements. Nuclear plants use the extracted plutonium and uranium as fuel and can continue reprocessing the spent fuel rods until the concentrations of uranium and plutonium are too low to allow reprocessing to be cost effective. The reprocessing method’s wastes are liquid and dangerous and difficult for disposal. Reprocessing plants utilize vitrification to turn the liquid wastes into solid waste. Vitrification is the process of dissolving the liquid high-level waste in molten glass and allowing the mixture to harden into a solid

Reprocessing allows the recycling of spent fuel rods decreasing waste by 75%. The waste also contains no plutonium which decreases the waste’s required storage time because plutonium has a half-life of 24,000 years which is longer than the other materials’ half-lives. Reprocessed waste decays to a background radiation level within 2,000 years, while unreprocessed waste takes 100,000 years.

6.4     Drawbacks and Environmental Damage

There are many drawbacks to both reprocessing and on-site storage which make them unable to facilitate the permanent storage needed for nuclear waste.

6.4.1 On-site Storage Drawbacks

In the United States there are 44,000 tons of spent uranium fuel rods. Nuclear plants store all the high-level waste on-site. There are many risks to storing the spent fuel on-site in cooling pools and steel and concrete casks. The amount of spent fuel also increases the risks of on-site storage because the cooling pools are full and the storage of new waste is in dry casks.  The Nuclear Regulatory Commission determined that spent fuel rods stored in dry casks are safe for 100 years. Therefore, the cooling pools and the dry casks are not permanent solutions because dry casks can safely store waste for 100 years and cooling pools do not have any more room for spent fuel storage. Adding to the problem is that the United States is increasing the total spent fuel by about 2,200 tons per year. Therefore, the government must find a solution which will last for 100,000 years and safely store the waste preventing environmental contamination. Also, on-site nuclear waste would increase radioactive waste contamination in the event of a nuclear plant meltdown. A meltdown could destroy the containment shelter allowing for more radioactive waste to contaminate the surrounding area.

6.4.2 Reprocessing Drawbacks

Reprocessing is also not a good solution because it has many drawbacks. The waste produced during the reprocessing method is liquid and more difficult to handle than the solid waste. Reprocessing plants must change the waste into a solid and then find a suitable storage facility for the waste.  Nuclear plants also must transport the nuclear waste to reprocessing plants. For example, Japan transports their nuclear waste across the ocean to Great Britain for reprocessing. Transporting the waste increases the risk of environmental contamination. Another drawback is that reprocessing isolates the plutonium from the waste material. The weapons grade plutonium causes security concerns because a terrorist group could use the plutonium to build a nuclear weapon. The security concerns of reprocessing nuclear waste led the United States to discontinue reprocessing in the 1970s. France and Britain reprocess their high-level waste. Reprocessing reduces, but does not eliminate, the requirement for the nuclear industry to store spent fuel securely to prevent the waste from contaminating the environment

7.                      Nuclear Waste Transportation

7.1              Waste Transportation Plans

The prospect of large amounts of spent fuel being transported from the reactor sites to a centralized location—whether an interim facility or a permanent repository—has led to a debate over the dangers that might result. The shipments, from many parts of the country, would have to pass through numerous governmental jurisdictions. Unless a public consensus develops that the dangers posed by these shipments are small, political and legal challenges could complicate the implementation of any transportation program.

The safety of the shipments depends on the characteristics of the wastes and their containers. The wastes are in solid form, mostly pellets of spent fuel contained in assemblies of fuel rods. The assemblies are to be placed into transportation casks at the reactor sites. The casks are massive structures that are designed to provide adequate shielding to keep the external radiation levels low and to be rugged enough to withstand potential transportation accidents. Accidents are expected to be rare, but if they occur, the cask, the fuel rod cladding, and the solid form of the fuel rod are the defense against the release of radionuclides.

7.2            Transportation System and Cask Design

The shipment of the transportation casks is to be made by a combination of truck, train, and barge. the transportation plans in terms of two possible scenarios, characterized as mostly rail and mostly truck. In the mostly rail case, some use is made of trucks for reactor sites where rail transport is not readily available. In this scenario.

The truck casks are smaller and a truck-only system would require more shipments than a mostly rail system. The DOE estimates that the transfer of 63,000 tonnes of commercial spent fuel and 7000 tonnes of defenses wastes (the amount authorized for Yucca Mountain) will require about 53,000 truck shipments (in the mostly truck case) These shipments would be spread over 24 years (optimistically, from 2010 to 2033) at an average rate of 2200 shipments per year. This means an average of six shipments per day would converge on the repository site from different parts of the country.

The physical arrangement for shipping is sketched in Figures 7 and 8. which present “artists concepts” of the configuration for truck and rail shipments. For truck shipments, a single cask would be placed on a trailer that is elongated to reduce the radiation level in the driver’s cab. In train transport, a single cask would be placed on a flatcar. The casks are designed to provide radiation shielding to protect people in the vicinity of the casks, including both the truck drivers and the general public. They must also protect the fuel assemblies against damage in case of an accident. A number of alternative designs have been under consideration, and until final decisions are made, their expected features are described by “generic” designs. Table 3 gives dimensions for generic truck and rail casks.

The fuel is surrounded by three concentric metallic protective layers: a stainless steel liner, a lead or depleted uranium layer for gamma-ray shielding, and a stainless-steel outer shell. In addition, a neutron shield surrounds the shell, and it, in turn, is protected by a relatively thin metallic outer layer. Uranium is chosen as a possible gamma-ray shield because of its high density (19 g/cm3) and high atomic number. The casks have “impact limiters” at the front for protection in case of collisions.

Fig. 7. Artist’s conception of transportation cask and carrier for truck transport;

total length = 18 m (56 ft).

Fig. 8. Artist’s conception of transportation casks and carrier for train transport;

total length = 21 m (66 ft).

Table 3. Dimensions of generic transportation casks (in inches).

Figure 9. A truck transporting three casks of transuranic waste to the Waste Isolation Pilot Plant (WIPP) in New Mexico, USA

8.         Deep Geologic Disposal

8.1     Multiple Barriers in Geologic Disposal

The handling of nuclear wastes is in the first instance the responsibility of the country in which the wastes are produced. Although there have been suggestions for the establishment of international repositories most countries are proceeding the basis of using a site within its own borders. In all countries that are engaged in active planning, the favored solution has been to place the wastes in deep geologic repositories. Protection against the escape of radionuclides into the biosphere is then provided by a number of barriers, with the overall set of barriers commonly divided into the engineered system and the natural system.

The waste package, auxiliary components such as a shield or backfill, and the configuration of the repository together constitute the engineered system. The waste package consists of the solid waste (the spent fuel assemblies or the resolidified products of reprocessing) and the surrounding protective containers. These are usually concentric cylinders made of materials that are chosen because of their ability to resist corrosion and prevent water from reaching the waste material. In some designs, additional protection against water intrusion is provided by a protective shield above the canister, and entry or escape of water may be hindered by backfill surrounding the waste package. The engineered system cannot be designed independently of the natural environment, because factors such as the water flow rate, the water chemistry, and the heat conductivity of the medium strongly influence the choice of waste package design and the repository configuration.

The natural system is the surrounding rock through which water would move to the repository, and from the repository to the biosphere. It includes the rock out of which the repository is excavated. A good repository site is one for which the location and type of rock  

(a) Prevent or limit the flow of water into the repository,

(b) Provide geochemical conditions favorable for a low rate of corrosion of the waste package and low solubility of radionuclides in the event of entry of water,

(c) Slow the outward migration of water to the biosphere,

(d) Retard the motion of major radionuclides so that they move more slowly than the water, and

(e) Are at low risk of future disruption by earthquake, volcano, erosion, or other natural phenomena.

 Together, these attributes provide a series of natural barriers.

Repositories may be in rock in a saturated zone, lying below the groundwater table, or in an unsaturated zone, lying above the water table. Except in arid climates, the water table usually lies too close to the ground surface to permit having a geologic repository in the unsaturated zone, and in almost all countries the planned repositories are in the saturated zone.

In the saturated zone, the gaps and pores in the rock are filled with water, although the site may still be suitable if the movement of water through the rock mass is at a slow rate. In the unsaturated zone, the pores hold less water but seepage of rain water introduces some moisture into the rock, and the environment of a repository in the unsaturated zone is unlikely to be completely dry.

8.2     Alternative Host Rocks for a Geologic Repository

A large number of different types of rocks have been considered for waste repositories. There is no single overall “best” choice, as evidenced by the different choices made by different countries. Among the physical factors that go into the consideration of a particular rock formation are the extent to which water entry would be inhibited, its retardation of the flow of any escaped radionuclides, and its behavior when heated by the repository wastes. Rocks that have been considered as candidates for repositories include the following

·         Bedded salt. Bedded rock salt was the initial candidate of choice. The existence of a salt bed was taken as evidence that there had been no water intrusion for many thousands of years. Further, salt has high thermal conductivity, which would limit the temperature rise of the wastes. Salt melts at relatively low temperatures, and the waste would eventually be surrounded by a tight resolidified mass of salt. On the negative side, salt brine is highly corrosive and may attack the canister.

·         Salt domes. Under some circumstances, the pressure on a thick bed of salt will cause some of the salt (which, in general, has a lower density than the surrounding rock) to break through the overlying material and rise upward to form a salt dome. One advantage of salt domes over bedded salt is a generally lower water content. The Gorleben site, a waste disposal site under consideration in Germany, is a salt dome.

·         Granite. Granite and similar rocks (granitoids) are very abundant. They are stable and generally homogenous, with low permeability to water movement. However, they are susceptible to fractures, which could provide paths for relatively rapid water flow. Granite is the choice in Sweden and Canada.

·         Basalt. Basalt is an alternative rock formation, although a National Research Council review has suggested that “a major reason for considering basalt for repositories is its abundance in federal land near Hanford, Washington and the Idaho National Engineering Laboratory (INEL) and not its overall favorable characteristics”.

·         Tuff. Tuff is the residue of material blown out of exploding volcanoes. At high temperatures, some of the material fuses to form “welded tuff,” a material of low permeability. Tuff, both welded and unwelded, is the rock type at Yucca Mountain. Tuff can be highly fractured, and a study of the fracture structure is an important component of the Yucca Mountain waste repository site characterization.

The suitability of a particular site depends not only on the type of rock but also on location-specific aspects, including the history of past human disturbance of the region the thickness of the available rock layers, and the absence of valuable mineral resources.

Figure 10: Conceptual Design of Yucca Mountain Disposal Plan

Table no. 4:

8.3     The Waste Package

8.3.1 Relation of the Waste Package to Its Environment

The waste package—and for reprocessed wastes, the waste form—is selected to limit corrosion of the canister and leaching of the waste. The rate of corrosion or leaching of any particular material depends on the chemical composition of the water attacking it, and, therefore, on the type of rock through which the water has traveled. Therefore, the choice of waste packaging materials must be tailored to the chosen site.

Fig. 11:  Cross-sectional illustration of proposed PWR waste package, together with drip shield (above) and support structure (below).

Figure 12:  Waste Package or waste container

8.3.2 Components of the Waste Package

The waste package consists of the spent fuel or reprocessed waste plus the surrounding protective container, which is typically in the form of one or more concentric cylinders. The overall requirement is that the container provide protection against physical damage (e.g., from falling rock in the drift) and resist corrosion. Commonly, there is an inner cylinder, or canister, with metal panels that are arranged in a honeycomb structure to hold the individual fuel assemblies. The panels also provide mechanical support, facilitate heat transfer to the container wall, and absorb neutrons to prevent the development of criticality. The same functions could be performed by filling the canister with a powder or sand. An outer cylinder, sometimes called the over pack, was originally envisaged as the heavier and physically stronger barrier.

Figure 13: Loading radioactive waste packages of LLW and ILW into disposal containers.

8.3.3  Placement of the Waste Packages

A typical underground repository design is a large cavern, honeycombed with tunnels, called “drifts.” In the early thinking about Yucca Mountain, for example, each tunnel was to have a series of vertical boreholes for emplacing the waste packages; a cylinder containing spent fuel assemblies or solid reprocessed waste would be placed in each borehole and the top portion of the borehole refilled and sealed. More recently, the Yucca Mountain planning has been based on larger canisters, placed horizontally on the floor of the tunnel, sitting on rail tracks. This simplifies their handling, provides flexibility in moving them, and makes retrievability easier if desired in the future. These two configurations are depicted schematically in Figure 13 and 14.

Figure 14. Placement of the Waste Packages in fuel spent fuel

In some designs, there is no extensive further protection against contact with water, other than the natural dryness of the environment plus the effects of repository heating. Alternatively, the cavity around the package can be filled with a backfill to impede water movement. A common choice is bentonite, a material made largely of clays. Bentonite swells when water enters it, impeding the flow of water toward or away from the waste containers. Further, it adsorbs many radionuclides, reducing their migration to a rate even slower than that of the water itself. However, bentonite may not be effective if subjected to high temperatures, and thus it may be more suitable for the relatively cool environment of the planned Swedish repository than for Yucca Mountain if plans for a hotter environment are adopted. Materials known as buffers can also be added to the backfill to condition the chemical composition of any water moving through the backfill.

Fig. 15:  Illustrative sketch of alternative containers and emplacement geometries for deep geologic disposal of spent fuel or reprocessed wastes. Left: thin-walled container in vertical borehole; right: thick-walled container in horizontal drift. The backfill in this illustration is coarse rock.

8.4     Variants of Geologic Disposal

In addition to excavated caverns, which everywhere remain the adopted approach for planning purposes, several other sorts of geologic site have been considered, although none is an active contender at the moment:

8.4.1   Deep-borehole disposal.

 The wastes would be placed in holes at depths of several kilometers in crystalline rock, such as granite. This option has not been given a great deal of attention. For example drilling the numerous boreholes probably entails higher costs than those for excavated caverns and the option faces difficulties in sealing the boreholes and maintaining retrievability. Suitable rock deposits are so common that the study even mentioned the possibility of locating boreholes at or near reactor sites. This recommendation by a prestigious group may foreshadow greater future consideration of deep-borehole disposal.

8.4.2   Rock melt

 High-level wastes in liquid form would be put into an underground cavity where, at high concentrations and confined volumes, they would melt the surrounding rock. Subsequent cooling and solidification, perhaps after about 1000 years, would trap the wastes in a well-sealed environment. Of course, the wastes would not be as well trapped at early times, while the material is liquid. Further, it would be difficult if not impossible to retrieve the wastes, even at times shortly after disposal.

8.4.3           Sub seabed Disposal

Deep-seabed or sub seabed disposal (SSD) provides a technically interesting alternative to geologic disposal. At present, it is not a viable option, because SSD is banned by an international agreement commonly known as the London Dumping Convention. Nonetheless, it warrants consideration, because the Convention could be modified if a consensus were to develop that SSD is practical and poses no significant environmental threats.

8.4.3.1          Main Features of Sub seabed Disposal

The deep seabed, at places where the ocean is several thousand meters deep, has been formed from the deposition of sediments over millions of years. The seabed is in the form of a water-saturated clay layer, on the order of 50 m thick. Its physical properties as a site for high-level waste burial were described in favorable terms in a 1994 National Academy of Sciences study on the disposition of plutonium from dismantled nuclear weapons:

The deep ocean floor in vast mid-ocean areas is remarkably geologically stable; smooth, homogenous mud has been slowly building up there for millions of years. The concept envisioned for HLW[high-level wastes] was to embed it in containers perhaps 30 meters deep in this abyssal mud, several kilometers beneath the ocean surface. . .the mud itself would be the primary barrier to release of the material into the ocean, because the time required for diffusion of radionuclides through this mud would be very long.

About 30% of the ocean floor (∼ 108 ) is composed of sediments of this sort, and one goal of SSD studies is to identify specific regions that have been geologically stable for long time periods, in the range of  years or more. An individual repository might be    in area. The canisters containing nuclear wastes could be emplaced in this sediment either by free fall through the ocean, by forcible injection, or in predrilled holes. The canisters might be individually emplaced, with one canister per penetration of the ocean floor or stacked, with several canisters apiece in holes drilled deep into the seabed. The holes would be sealed either mechanically or by natural processes in the sediment. Once the canisters are in place and buried, they are in a uniform environment over wide regions, so that the durability of the canisters and the rate of migration of radionuclides through the clay will not vary greatly from place to place.

8.4.4                     High-Level Nuclear Waste in Space

The current cost to launch an object into orbit around the earth is about $20,000 per kilogram. Beamed energy technology (BEP) based on laser-powered propulsion of objects into space may considerably lower the cost. Figure 5 is a model that shows the very small size of the BEP launch container. If BEP is successful, it could send waste into high orbit for about $200 per kilogram. However, BEP is at least 15 to 25 years from being a real alternative because the highest flight using BEP technology is currently less than a few hundred meters. Moreover, a conservative estimate of the cost of developing BEP technology is $10 billion. Therefore, the adoption of BEP technology is unlikely. Nevertheless, BEP would solve the problem of nuclear waste storage and disposal because BEP could send nuclear waste out of our atmosphere into orbit.

8.4.4.1        Nuclear Waste in Space

BEP represents a possible future solution to the nuclear waste storage problem, but it will be a long time before anyone can say whether it is effective. It promises to be a clean technology—the only trash that is left in space is the small capsule containing the nuclear waste, and there is no potential for explosions in the atmosphere. BEP would require tremendous resources and a lot of time to develop, but if the technology can do what scientists predict, it represents the easiest and cheapest of the solutions to the nuclear waste problem. Nevertheless, BEP ought to be dismissed from consideration for now because it is so great a leap in technology. It is not possible to say with certainty that it would ever be possible to send our waste into space in this way. As a result, BEP is not a factor in the ethical problem humanity faces, because that problem is occurring right now and cannot wait such a long time for an unproven technology.

9.         Safety Relevant Waste Properties And Acceptance Criteria

When discussing the safety relevant waste properties, it is convenient to distinguish between the operational phase and the post-closure phase. The operational phase of a geological repository is, in principle, comparable to the operation of a storage facility. The most important safety relevant processes and the safety-determining factors for the operational phase of a repository are summarized in Table 5.

Groundwater flow is the most important pathway for any potential radionuclide releases in the post-closure phase. For the analyses of this release scenario it is convenient to divide the system into three components: the near field, the geosphere and the biosphere. The near field (consisting of the disposal areas with the wastes and the engineered barriers and a few meters of directly surrounding host rock) ensures that only very small release rates will occur because of the very low water flow rates and retention by geochemical phenomena. Transport through the near field and the geosphere does take such a long time that the majority of radionuclides decays before they can reach the biosphere. The long transport times are again mainly due to the small water flow rates and the geochemical retention processes.

Based on these considerations for both the operational and post-closure phase quantitative and qualitative criteria were developed for the preliminary waste acceptance criteria. Key criteria are summarized in Table 6.

Table 5:          Safety relevant processes and factors for the operational phase of a repository

	Process	IMPORTANT FACTORS
		Waste package	Repository
Normal operation	Direct radiation	Nuclide inventory Shielding by package Shielding by waste matrix Number of packages	Shielding, operational procedures, remote handling, etc.
	Release of volatile nuclides	Nuclide inventory (average) Gas tightness of package Release of volatile nuclides from waste matrix Number of packages	Ventilation Backfilling of disposal areas
Incidents & accidents	Release of airborne nuclides due to mechanical impact	Nuclide inventory (maximum) Dispersability of waste matrix Mechanical protection by package	Operational procedures, handling equipment, transport container, disposal container. Remote handling Ventilation
	Release of airborne nuclides due to thermal impact	Nuclide inventory (maximum) Thermal stability of waste matrix Thermal protection by package	Operational procedures, handling equipment, transport container, disposal container Remote handling Ventilation

Table 6:  Key criteria in preliminary waste acceptance criteria

10.           Passive Safety in the Storage of Radioactive Materials and Radioactive Waste

Radioactive materials and radioactive waste should be stored according to the principles of passive safety. The more hazardous the waste (for example, HLW) and the more mobile its form, the greater the safety benefit from passively safe storage and the sooner this should be achieved. Only when this is not reasonably practicable, should potentially mobile wastes be accumulated in a raw state for significant periods.

Passive safety requires the radioactive wastes and materials to be immobilized in a form that is physically and chemically stable and stored in a manner which minimizes the need for control and safety systems, maintenance, monitoring and human intervention. The wastes and materials should be stored in discrete packages which are resistant to degradation and hazards and which can be inspected and retrieved for final disposal. There is need to ensure that waste packages remain in a safe condition for a period of at least 150 years. The passive safe storage of radioactive materials and radioactive waste has potential long term safety benefits which clearly help to achieve this requirement.

10.1   The Achievement of Passive Safety

Passive safe storage of radioactive materials and radioactive waste is most appropriately achieved by providing multiple physical barriers to the release of radioactivity to the environment. The physical barriers include the form of the waste or material itself, the material used for encapsulation, the waste container and the storage building or structure, each of which should be designed to provide effective containment and prevent leakage.

In its strictest sense, passive safety requires that safety is assured without dependence on active systems, maintenance, monitoring or human intervention. However, with respect to the long term storage of radioactive waste, it may be necessary or advantageous for active systems to be in place. In such cases, the systems should be designed for minimum maintenance and, in the event of failure, immediate repair/replacement should not be necessary in order to ensure continuing safety of the storage facility and its contents.

10.2   Form of the Radioactive Material and/or Radioactive Waste

The primary consideration is to ensure that the radioactive material or radioactive waste is immobile and is contained in order to minimize the potential for dispersal. The radioactive material or waste should therefore be in a form which is physically and chemically stable and should also be resistant to any significant deterioration over the storage period.

Certain raw radioactive wastes may be in a form for which the radioactivity is already immobile and therefore meet the requirements for passive safety without the need for processing. Such cases will require to be demonstrated, but examples could include robust metallic components.

In many cases, the raw radioactive material or radioactive waste will require conditioning to place it into a passively safe form to immobilize the radioactivity. Typical waste forms that fall into this category are gases, liquids, wet solids, slurries, sludges, powders, particulate material, bulk material and radioactive materials including spent fuel. The conditioning processes that are typically used for immobilization of liquids and solids are encapsulation in cement or vitrification.

Other raw radioactive wastes may be in a form for which some intermediate processing may be required prior to conversion into a passive safe form. For example, highly reactive or corrosive substances should be neutralised or made less reactive by chemical processes. In the few cases where a raw radioactive waste is not suitable for processing, then these wastes should be identified and an acceptable alternative strategy for their future management developed.

The particular properties that can be expected to be met by a passively safe form of radioactive waste or radioactive material are:

·         Stored potential energy should, as far as possible, have been removed from the system. This can arise from, for example, the effects of gravity, chemical energy, water pressure, internal pressure and Wigner energy (graphite).

·         The form of the material or waste should have low chemical reactivity, for example, low solubility, low flammability, not be explosive and not need inerting.

·         Where the waste or material is known to generate gases, the packaging should include provision for venting.

·         The form of the material or waste should be resistant to degradation over the period of storage. Potential mechanisms include corrosion, action of water and microbiological action.

·         The form of the material or waste should not require cooling other than by natural circulation.

10.3   The Waste Container and Encapsulation Material

In order to contribute to passive safety the container should have attributes similar to those already identified for the form of the material or waste. It should be resistant to degradation over the period of storage and should be resistant to the range of foreseeable internal or external hazards to an extent that neither the containment function nor the ability of the container to be handled safely are significantly impaired.

In general, the waste package (i.e. the waste form, the encapsulation material and the container) should be designed to be suitable for long term storage, transport and potential final disposal of the waste. This will minimize the amount of reworking prior to disposal.

Waste packages should be uniquely identifiable via appropriate labeling. The method of labeling should be designed to ensure identification over the expected period and conditions of passive safe storage. Before being placed in storage, waste packages should have been monitored and cleared for the presence of surface contamination which could otherwise initiate or accelerate corrosion of the package. Suitable arrangements should be available for dealing with any surface contamination that is found.

10.4   Storage Building or Structure

The storage building or structure is the final physical barrier to the release of radioactivity to the environment. It is noted however, that in aiming to achieve passive safety the most significant barriers are first and foremost the waste form itself, and secondly the waste container. In some cases, the role of the storage building or structure may be limited to providing environmental protection, radiation shielding and presenting a secure boundary against unauthorized intrusion or interference and entry of wildlife.

It should demonstrate that the design of the storage building or structure is fit for purpose, taking account of the expected time required for passive safe storage and the hazards posed by the stored wastes i.e. the design should be proportionate to the defined purpose of the building and to the risks.In some cases, a building may be designed for a shorter life with the intention of periodic refurbishment. In these cases, justification should be provided that the waste can be stored safely while the refurbishment is carried out.

The building should be designed to be resistant to the range of foreseeable internal and external hazards. The storage building will need to provide sufficient protection to the stored wastes so as to optimize the life of the packages and to facilitate safe transfer to the final disposal facility (or to a further storage facility) at the appropriate time. This may necessitate control and monitoring of the environment of the storage building (temperature, relative humidity and constituents of the atmosphere) and also of the surface temperature of the waste packages in order to minimize corrosion rates. This may be particularly important on near coastal sites where chloride levels in the atmosphere are relatively high. Such environmental control cannot be achieved by purely passive means and it may be necessary to adopt a forced ventilation system with control of relative humidity and a filtered inlet to remove atmospheric contaminants such as salts.

Monitoring systems and alarms will need to be provided to detect off-normal conditions such as off-normal temperature and relative humidity in the atmosphere of the facility, buildup of flammable gases, water ingress, fires and unauthorized intrusion. A radiation monitoring system would provide the ability to detect radioactivity in liquid or gaseous forms in the event of damaged/deteriorated packages. Groundwater should also be routinely monitored. Wherever possible, the panels and electronics associated with the monitoring system should be situated in a safe area of the building or externally.

The design of the building should facilitate the retrieval of all waste packages either for inspection, possible remedial treatment, further storage elsewhere or for disposal at the end of the period of passive safe storage or at an earlier time should radioactive waste management strategies change. Waste handling equipment may not be continuously available, but should be capable of being returned to service when needed and should be maintainable within a safe area either inside or external to the building. Depending on the radiation levels associated with the waste packages, remote or manual handling techniques will be necessary.

One of the foreseeable mechanisms for the mobilization of radioactivity in waste is the ingress and action of water in a store. Potential sources of water ingress are groundwater, rainwater, flooding and condensation. An effective means of reducing potential water ingress is to situate the storage building above ground level. If a building is below ground level then it is best situated above the local water table. In general, it can be expected that the design of a storage building will include features to monitor for water ingress and the means to remove the water. These features could involve a sloping floor, collection sump with level alarm and safe facilities to pump out water and monitor for radioactivity prior to authorized disposal.

The need for human involvement to ensure safety should be minimized. Ideally, no continuous human presence or supervision should be required. Human involvement should be limited to confirmatory surveillance, inspections and responding to incidents on a reasonable timescale.

10.5   Hazards

The multiple physical barriers to the release of radioactivity from the waste should provide resistance to dispersal as a result of a range of foreseeable external and internal hazards. Any hazards, which could cause deterioration of the waste form, container or building over the storage period, should be taken into account including, for example, corrosion, water or microbiological action. For external hazards, such as weather and flooding, account should be taken of long term trends such as rising sea level or climatic change. The waste itself may be the source of hazards, in that it may have the potential for criticality or radio lytic gas generation.

10.6   Records

The interim storage of radioactive waste in a passive safe form may last for a period of more than 150 years before the disposal facility is closed. Comprehensive records need to be assembled as part of the storage arrangements. They need to be securely retained and to be accessible when required.

10.7   Radiation Shielding

Adequate shielding of operators and the public against the radiation hazard from the radioactivity in the waste should be provided by a combination of the waste form, the waste container and the storage building or structure.

10.8   Radioactive Discharges

If the long term storage of radioactive waste will involve the discharge of radioactivity to the environment, for example, gaseous discharges may occur via ventilation systems and liquid discharges may occur from systems designed to maintain dry conditions in the store. Provision should be made for mitigating the release of radioactivity from the facility in the event of off-normal conditions, for example, by filtration or isolation.

Table 7:      General Principles for Passive Safety

Principle

The radioactivity should be immobile

The waste form and its container should be physically and chemically stable Energy should be removed from the waste form

A multi barrier approach should be adopted in ensuring containment

The waste form and its container should be resistant to degradation

Storage environment should optimize waste package life

The need for active safety systems to ensure safety should be minimized

The need for monitoring and maintenance to ensure safety should be minimized

The need for human intervention to ensure safety should be minimized

The storage building should be resistant to foreseeable hazards

Access should be provided for response to incidents

There should be no need for prompt remedial action

The waste packages should be inspect able

The waste packages should be retrievable for inspection or reworking

The lifetime of the storage building should be appropriate for storage period prior to disposal

The storage facility should enable retrieval of wastes for final disposal (or restoring)

The waste package should be acceptable for final disposal

11.           Inspection of Accumulated and Stored Radioactive Materials and Radioactive Waste

Radioactive materials and radioactive waste are accumulated during the operating and decommissioning phases in the lifecycle of a nuclear facility. For those radioactive wastes for which there is no current disposal route, licensees will need to plan for long periods of storage. for radioactive waste and material being placed in storage now, an overall period of containment of at least 150 years should be assumed

The fundamental objective of inspecting accumulated and stored waste is to confirm that the waste packages and facilities are, and will remain, in an acceptable condition for continuing safe storage, retrieval, conditioning and final disposal.

In defining their inspection regimes, licensees should develop acceptance criteria against which the condition of the waste is to be assessed. They should justify the method of inspection (which could involve visual, nondestructive or destructive techniques) and the frequency of inspections. Where the inspection regimes are based on predicted rates of degradation, inspections should also be undertaken at appropriate time intervals to confirm that the waste is not deteriorating to an unexpected degree.

11.1             Inspection of Accumulations of Raw Radioactive Waste

At nuclear facilities some radioactive wastes are accumulated in their raw form, with the intention of retrieving them for treatment and packaging, either after the plant has shutdown, or when sufficient waste has been accumulated to make a campaign cost-effective. Typical examples of such waste forms include spent ion-exchange resins and sludges which are accumulated in tanks, awaiting retrieval and encapsulation in cement in steel drums.

There are a number of processes that can change the physical and chemical form of the raw radioactive waste during accumulation. For example, the agglomeration and consolidation of ion-exchange resins and sludges stored under water in tanks can adversely affect the ability to retrieve them by hydraulic means. Similarly, the corrosion of metallic solid radioactive wastes can adversely affect their retrieval by mechanical means. An important aim of inspections is to confirm that the rate of degradation of the waste will not impact on the ability to retrieve and process the waste in the future as planned. This will normally be achieved through direct sampling and analysis of the accumulated waste to verify its condition.

11.2            Inspection during Passive Safe Storage of Radioactive Waste

Radioactive wastes for which there is no current disposal route should be processed for long term passive safe storage. Although one of the aims of passive safety is to minimize the need for surveillance and inspection to ensure safety, it is expected that periodic inspections will be carried out to confirm that the condition of the waste and its storage are not deteriorating adversely, and to confirm its continuing acceptability for safe storage, and ultimately retrieval, transport and disposal. Inspection will not be restricted to the waste packages but will cover the storage facilities and buildings, and the associated safety arrangements.

The design of storage facilities to take account of the needs for inspection, retrieval and transfer. All the radioactive waste accumulated or in storage to be routinely inspect able. Where only a fraction of the waste is to be inspected.

12.           Radioactive Waste In Hospitals/Nuclear Medical Centers

The radioactive waste at hospitals/nuclear medical centers mainly comprises of low level

(i)                 solid

(ii)               liquid and

(iii) gaseous waste .

Solid Waste: Solid waste mainly consists of used Molybdenum‑Technetium generators. empty vials, swabs, syringes, gloves, laboratory clothing, bench covers, absorbents etc.

Liquid Waste: Liquid waste includes washing from active labs., and excreta of patients injected/ingested with radiopharmaceuticals. Biological waste such as excreta or macerated material is regarded as liquid waste.

Gaseous Waste: Gaseous waste generally includes working with, tritium and tritiated water, iodine and xenon‑133.

13.           Management Of Radioactive Waste In Hospitals/Nuclear Medical       Centers

There are two principal ways to deal with the radioactive waste. In the first method, waste containing radioactive material  is stored under controlled conditions until it has decayed to background level so that disposal can be carried out. In the second method,  the activities are disposed of to the environment in such a way that natural processes transfer it back to man only in such amounts that, in combination with other sources of radiation, the resulting radiation doses are negligible.

13.1   Management of Solid waste

13.1.1                       Collection:

The collection of radioactive waste requires distribution of suitable containers (strong bins) throughout the working area to receive discarded radioactive material. Each container should be lined with heavy gauge plastic bag. The plastic bag should be marked with the name of radionuclide and date. The container should be brightly colored (e.g. yellow) with the radiation symbol clearly displayed so as to distinguish it from bins of inactive waste. Separate container should be used for each radionuclide at the point of origin. However, different radionucides having almost same half-life can be collected in one container (bin). Proper shielding should be provided for the containers to keep the radiation level within limits.

13.1.2         Storage:     The storage for decay is particularly important for radioactive waste resulting from medical/research uses of radionuclides. Many of the radionuclides used are of small activities and short lived.  The radioactive waste should be stored for decay purposes until the activity decays to the background level and can be considered inactive for final disposal as normal waste. The storage for decay is suitable for wastes containing radionuclides with half-lives of less than or equal to 100 days.

Keeping in view the working practices in our country, the following recommendations are made for storage of radioactive waste for decay purposes:

i.                    Tc-99m  should be stored for three months

ii.                  I-131, Tl-201, Ga-67 and Mo-99 (Tc-Generators)  should be stored for six months

iii.                In general, the radionuclides having half-lives up to  3 days should be stored for three months and radionuclides having half-lives up to 10 days should be stored for six months

iv.                Radionuclides not covered in i-iii should be stored for at least 10 half-lives as specified above 

13.1.3                   Disposal:     After ensuring that the radioactive waste has  completed its decay period,  the  radioactive waste can be disposed of as normal waste after  monitoring  its residual activity. All the radiation symbols, if any,  should be removed from the radioactive waste packages (plastic bags) before disposal as normal waste. The waste disposal record should be properly maintained.

13.2   Management of Liquid Waste

The liquid waste should be collected/stored in double stage delay tanks and discharged into the sewerage system when the activity approaches background level. Alternatively, all the liquid waste (active & non-active) should be collected in a large dilution tank and discharged to the sewerage system. Samples of liquid waste should be taken and analyzed before its release into the normal sewerage. In case of small laboratories dealing with RIA facility or research activities, the liquid waste should be collected in polythene bottles and disposed of in the normal sewerage after proper analysis. Before discharge; it should be ensured that all radioactive materials released into the sewer system are completely soluble and dispersible in water. Liquid, if it contains suspended solids or sediments, may need to be filtered prior to discharge. Non‑aqueous wastes which are immiscible with water should be completely excluded and stored  separately. Excreta from patients and samples such as urine and blood from patients who have received radioactive compounds should also be stored in the above mentioned tanks and discharged to normal sewerage accordingly. A complete and up‑to‑date record should be maintained of all the discharges.

The liquid waste generated due to the application of H-3 and C-14 or other long-lived radionuclides should not be disposed of into the normal sewerage system. This waste should be stored separately and sent to Pakistan Institute of Science and Technology (PINSTECH), Islamabad or Karachi Nuclear Power Plant (KANUPP) for proper disposal.

13.3   Management of Gaseous Waste

Particular care should be taken for the management of gaseous waste. The main problem of gaseous waste is the release of activity to the environment. Appropriate  filter should be used to trap the airborne radioactivity in the exhaust systems of the fume hood/labs.  The contaminated filters should be treated as solid radioactive waste.

GLOSSARY

activity. Of an amount of a radioactive nuclide in a particular energy state at a given

time, the quotient of dN by dt, where dN is the expectation value of the number

of spontaneous nuclear transitions from that energy state in the time interval dt:

The unit is

The special name for the unit of activity is becquerel (Bq): 1 Bq =1

(Although becquerel is a synonym for reciprocal second, it is to be used only

as a unit for activity of a radionuclide.)

In practice, the former special unit curie (Ci) is still sometimes used:

1 Ci = 3.7x     (exactly).

conditioning. Those operations that produce a waste package suitable for handling, transportation, storage and/or disposal. Conditioning may include the conversion of the radioactive waste to a solid waste form, enclosure of the radioactive waste in containers, and, if necessary, providing an overpack.

contamination. The presence of radioactive substances in or on a material or in the human body or other place where they are undesirable or could be harmful.

disposal. The emplacement of waste in an approved, specified facility (e.g. near surface or geological repository) without the intention of retrieval. Disposal also covers the approved direct discharge of effluents (e.g. liquid and gaseous wastes) into the environment, with subsequent dispersion.

disposal, geological. Isolation of waste, using a system of engineered and natural barriers at a depth up to several hundred metres in a geologically stable formation. Typical plans call for disposal of long lived and high level wastes in geological formations.

disposal, near surface. Disposal of waste, with or without engineered barriers, on or below the ground surface where the final protective covering is of the order of a few metres thick, or in caverns a few tens of metres below the Earth's surface. Typically, short lived, low and intermediate level wastes are disposed of in this manner. This term replaces 'shallow land/ground disposal'.

fuel, spent (used). Irradiated fuel not intended for further use in reactors.

fuel cycle (nuclear). Processes connected with nuclear power generation, including the mining and milling of fissile materials, enrichment, fabrication, utilization and storage of nuclear fuel, optional reprocessing of spent fuel, and processing and disposal of resulting radioactive wastes.

radioactivity. Property of certain nuclides to undergo spontaneous disintegration in which energy is liberated, generally resulting in the formation of new nuclides. The process is accompanied by the emission of one or more types of radiation, such as alpha particles, beta particles and gamma rays.

radionuclide. A nucleus (of an atom) that possesses properties of spontaneous disintegration (radioactivity). Nuclei are distinguished by their mass and atomic number.

repository. A nuclear facility where radioactive waste is emplaced for disposal. Future retrieval of waste from the repository is not intended.

reprocessing. Recovery of fissile and fertile material for further use from spent fuel by chemical separation of uranium and plutonium from other transuranic elements and fission products. Selected fission products may also be recovered. This operation also results in the separation of wastes.

segregation. An activity where waste or materials (radioactive and exempt) are separated or are kept separate according to radiological, chemical and/or physical properties which will facilitate waste handling and/or processing. It may be possible to segregate radioactive from exempt material and thus reduce the waste volume.

solidification. Immobilization of gaseous, liquid or liquid-like materials by conversion into a solid waste form, usually with the intent of producing a physically stable material that is easier to handle and less dispersable. Calcination, drying, cementation, bituminization and vitrification are some of the typical ways of solidifying liquid radioactive waste.

storage (interim). The placement of waste in a nuclear facility where isolation, environmental protection and human control (e.g. monitoring) are provided and with the intent that the waste will be retrieved for exemption or processing and/or disposal at a later time. transportation. Operations and conditions associated with and involved in the movement of radioactive material by any mode, on land, water or in the air.The terms 'transport' and 'shipping' are also used.

vitrification. The process of incorporating materials into a glass or glass-like form. Vitrification is commonly applied to the solidification of liquid high level waste from the reprocessing of spent fuel.

waste, heat generating. Waste which is sufficiently radioactive that the energy of its decay significantly increases its temperature and the temperature of its surroundings. For example, spent fuel and vitrified high level waste are heat generating, and thus require cooling for several years.

waste, high level (HLW). (1) The radioactive liquid containing most of the fission products and actinides originally present in spent fuel and forming the residue from the first solvent extraction cycle in reprocessing and some of the associated waste streams.

waste, long lived. Radioactive waste containing long lived radionuclides having sufficient radiotoxicity in quantities and/or concentrations requiring long term isolation from the biosphere. The term 'long lived radionuclide' usually refers to half-lives greater than 30 years.

waste, low and intermediate level. Radioactive wastes in which the concentration of or quantity of radionuclides is above clearance levels established by the regulatory body, but with a radionuclide content and thermal power below those of high level waste. Low and intermediate level waste is often separated into short lived and long lived wastes. Short lived waste may be disposed of in near surface disposal facilities. Plans call for the disposal of long lived waste in geological repositories.

waste, short lived. Radioactive waste which will decay to a level which is considered to be insignificant, from a radiological viewpoint, in a time period during which institutional control can be expected to last. Radionuclides in short lived waste will generally have half-lives shorter than 30 years.