Waste acceptance criteria (WAC) for disposal, a principal requirement for qualification of a produced waste package, are established predominantly on the radiological parameters of waste packages. Radiological parameter control is, therefore, considered to be the main component of a RAW control system. The WAC are country-specific; however, the IAEA recom­mendations for the establishment of WAC are accepted as the basis world­wide. Besides surface dose rates, maximum permitted activity concentrations (or total activity per entire waste package) of several radionuclides is usually defined in a WAC. The list of considered radionuclides is different for each country’s WAC for disposal. Besides common and simple measur­able radionuclides (such as Cs-134, Cs-137, Sr-90, etc.), a declaration of the activity concentration of 10-40 so-called critical radionuclides for disposal (alpha emitters, biologically important radionuclides, long-lived radio­nuclides usually with half-life over 30 years, etc.), is also required in a WAC. These radionuclides are often difficult to measure. To facilitate declaration of waste package compliance with a WAC, non-destructive (mainly gamma spectrometry) as well as destructive radiochemical procedures (with radio­chemical processing of the samples) are routinely applied.

Radiological control is applied in the entire life cycle of RAW. However, analogous to chemical parameters, the main effort is put on the radiological control of ‘as generated’ (raw) waste and then on the declaration of RAW package compliance with a WAC. For radiological control of ‘as generated’ waste, carefully selected combinations of non-destructive instrumental methods and radiochemical analysis with separation and subsequent deter­mination of difficult to measure radionuclides (some fission products, transuranium elements, etc.) are applied. The information obtained is widely used in waste processing planning for each waste stream and in the estima­tion (prognosis) of final waste package parameters. Results of radiochemi­cal analysis of input waste are also used for determination of radionuclide vectors, necessary for application of scaling factor methods (see below).

Most often a non-destructive check of the entire waste package is used for a declaration of final waste package compliance with a WAC. Gamma scanning, gamma tomography and in some cases also neutron tomography, all in combination with advanced data processing, are commonly used by both waste package producer as well as by disposal facility operator. The above-mentioned techniques allow determination of the major gamma — emitting radionuclides and along with using neutron tomography deter­mines the major actinides and fissile material. In general, non-destructive determination of minor radionuclides, critical for disposal, is very compli­cated, expensive, and in some cases even impossible. Destructive determina­tion with sampling of the waste form and waste package material and subsequent laboratory radiochemical analysis is not only technically com­plicated but can cause unacceptable damage to one or more of the waste isolation barriers in the waste package. The situation is more substantial for processed liquid waste, where critical disposal radionuclides can be expected with higher probability. The way around this situation is the application of scaling factors and a nuclide vector methodology [8]. The substance of this method is simple; however, implementation is more complicated and requires special software tools. Careful and precise radiochemical analysis of homogenized waste before the start of its processing is used to establish the nuclide vectors — a mathematical relationship between the activity con­centration of major or easy-to-determine radionuclides (usually strong gamma emitters) and the activity concentrations of minor (usually difficult — to-determine) radionuclides is developed. Using nuclide vectors and thor­ough knowledge of the waste processing procedure and waste package materials, it is possible to calculate and declare activity concentrations of minor radionuclides in a waste package using measured data on the activity of the major radionuclides, obtained by non-destructive gamma scanning of the entire waste package. Such a procedure should be, of course, qualified and approved by the regulator and disposal facility operator.

International safety standards have been developed for the transport of all forms of radioactive material [62] and are issued in the form of ‘transport regulations’. These ‘regulations’ have been adopted within all the regula­tions for transport of hazardous materials by all modes (land, air and sea) and are recognised and adopted in the national regulations of most coun­tries. As with all other facilities and activities associated with RAW manage­ment and contaminated environments, the radiation safety requirements are those set down in the international Basic Safety Standards [38].

The objective of the regulations is to establish requirements that must be satisfied to ensure safety and to protect persons, property and the environ­ment from the effects of radiation in the transport of radioactive material. This protection is achieved by requiring: (a) containment of the radioactive contents; (b) control of external radiation levels; (c) prevention of critical­ity; and (d) prevention of damage caused by heat. The regulations are satis­fied firstly by applying a graded approach to content limitations for packages and conveyances and to performance standards applied to package designs, depending upon the hazard of the radioactive contents. Secondly, they are satisfied by imposing requirements on the design and operation of packages and on the maintenance of packagings, including consideration of the nature of the radioactive contents. Finally, they are satisfied by requiring adminis­trative controls, including, where appropriate, approval by competent authorities. Confidence in this regard is achieved through the adoption of appropriate management systems involving quality assurance and compli­ance assurance programmes. The regulations are based on a classification of radioactive materials to be transported in a system of increasing hazard potential. The type of package and its testing are correspondingly higher as the hazard potential increases, with prescriptive testing and defined per­formance criteria for each category of package.

At the lower hazard level, low specific activity (LSA) material and surface contaminated objects (SCO) are defined quantitatively in the transport regulations. These materials can be transported in so-called ‘industrial pack­ages’ (IP) of types 1, 2 and 3, which must be designed and tested according to the specifications set in the regulations. The next generic class of materi­als is referred to as Type A and a schedule of radionuclide specific activity limits is provided in the regulations. Materials falling within these limita­tions can be transported in Type A packages for which design and testing requirements are prescribed in the regulations. The packages are designed to maintain their integrity during normal conditions of transport, providing the necessary shielding and containment, but are not expected to withstand severe transport accidents, the limitation on radioactive content ensuring that any consequences would not be severe. IP and Type A packages must conform to these design and testing requirements but do not require com­petent authority approval nor is notification required for international ship­ments of these package types. For transporting quantities of radioactive material greater than the limits for Type A packages and fissile material requires the use of Type B and C packages. These packages are designed to transport higher activity radioactive and fissile material and have to be designed with high integrity in terms of both shielding and containment features, which must be able to withstand the impacts of the most severe transport accident. Again, design and testing requirements are specified in the regulations, the latter including drop, puncture, crush and fire tests, representing the conditions that could be encountered in severe accidents, with Type C having to undergo impact testing simulating an aircraft acci­dent in order to qualify for transporting high activity radioactive material by air. Type B and C containers require competent authority approval, both from the country of origin for Type B and also from the countries en route during the shipment.

There is no direct correlation between RAW classes and transport cat­egories, as the classification is based on long-term safety (primarily dis­posal) consideration. Nevertheless, in general, low activity waste — generally VLLW from, for example, lightly contaminated building rubble from decom­missioning activities — would fall in the category of LSA for transport purposes, LLW and ILW would be/could be LSA, Type A or Type B and HLW would be transported as Type B material.

3.5 Conclusion

Significant progress can be observed in the development of internationally agreed standards on the management of RAW, radiation safety and trans­port safety in recent years. The role of each country is to implement these standards in the most efficient and appropriate manner, taking into account the specific characteristics and conditions of existing RAW or anticipated future arisings. One of the main challenges is for operators and regulators to apply a graded approach based on the existing and potential risks to the public and the environment and at the same time providing confidence in the demonstration of adequate levels of safety.

Recent advances in separations and immobilization sciences make reproc­essing more effective than ever for reducing the impact of nuclear waste on the environment. Waste form development has been focused in two primary directions that tend to be competitive: (1) reduction in complexity and cost of waste treatment, storage, and disposal and (2) improvements in the long­term performance of waste forms containing long-lived radionuclides. These endeavors allow for choices to be made in how regulatory dose limits are met, and where and how limited resources are spent.

5.2 Conclusion

Management of irradiated fuel continues to challenge the nuclear power industry. Despite improvements in packing efficiency in fuel storage pools, continued delays in establishing geological repositories for permanent dis­position have led utilities to move irradiated fuel into dry storage. The design for the Yucca Mountain repository in the US is in an advanced state, but this project has been suspended. Sweden and Finland have made sub­stantial progress in locating sites and in designing geological repositories for disposing of irradiated fuel based on the once-through fuel cycles prac­ticed in those countries. In countries such as France and Japan, irradiated fuels are processed to recycle the uranium and plutonium components for further energy production using mixed oxide fuel. Research is underway worldwide to develop advanced fuel cycle concepts that not only recycle the uranium and plutonium, but also the minor actinide component of the fuel. The goal of U/Pu recycle is to extend the supply of fuel; the primary goal of minor actinide recycle is to reduce the long-term radiological hazards associated with irradiated fuels from millions of years to a few hundred years. Regardless of the fuel cycle implemented, the choice of waste form is of critical importance to the safe disposition of the radioactive components. In reprocessing operations, consideration must be given to the waste forms used to immobilize volatile fission products, cladding, hulls, and other hardware components, undissolved solids, and the HLW stream.

5.3 Sources of further information

Yucca Mountain Repository License Application, http://www. nrc. gov/

Container failure by general corrosion is likely to result in sufficient opening of the container surface to allow substantial advective release of radionu­clides. The rate of release of radionuclides by advective release is expected to be higher by several orders of magnitude than diffusive release that may occur through tight cracks or small pits. Therefore, underestimating the general corrosion rates because of the uncertainties may lead to an inac­curate, delayed and low-magnitude radionuclide release from a failed con­tainer. When the uncertainties associated with the general corrosion rates are random in nature, the general corrosion rates are expressed in a uniform distribution, either in a linear scale or in a log scale, depending on charac­teristics of the uncertainties. Figure 7.4 is an example output of the failure probability of a carbon steel container with time, using the range of general

corrosion rates sampled from a log normal distribution extracted from Fig. 7.2 (David et al., 2002; Jung et al., 2011). The radionuclide release at a given time will begin from a finite number of containers that failed, from calcula­tions using the probability of failure and the total number of containers.

Container failure by localized corrosion may also limit the radionuclide release because of restricted flow through the small perforations due to the pits. Pit diameters were from micrometers to millimeters (0.4 microinch to milliinch) and pit density is 0.1-100/cm2 (0.6-645/inch2) as shown in Fig. 7.5 from selected metals and aqueous environments (Szklarsksa-Smialowska, 1986). In addition, the pits are usually filled with corrosion products or solid precipitates from groundwater. Therefore, radionuclide release through the restricted area is likely to be diffusive, which is generally slower than advec — tive release.

Planning factors such as ER under a life-cycle perspective and nontechni­cal issues are increasingly influential factors. Issues to be mentioned are the resources to aid good planning (with economics taken into account). How effectively are they being used? Again, how the experience from more advanced countries in the field of ER can be better transferred to less advanced ones? How to best incorporate ER in the whole life cycle of an operation and also how to optimize remediation programmes taking into account the life cycle of the projects? What are the best ways to engage stakeholders in the decision-making process? What should be communi­cated and how? What are the challenges in the different geographical regions of the world? How to clearly state to the public (and be convincing on it) that remediation does not mean returning the environment to back­ground levels; instead, new productive uses can be envisaged after ER? Who are the relevant stakeholders and how to best approach them? Ethics of ER remains crucial: will optimization justify higher expenditures in afflu­ent countries in comparison to less developed countries?

8.5 Conclusion

In the past, many nuclear activities were developed without proper consid­eration of environmental issues. Operations took place without established or well-addressed environmental laws and regulations. Through lack of good operating practices, contaminated sites have been created in many countries. Several contaminated sites have also been created by nuclear and radiological accidents.

Contaminated sites can ultimately lead to undesired health effects to the local residents. Environmental remediation strives to reduce the radiation exposure from contamination of land or other polluted media, such as surface water or groundwater.

In recent years a dramatic change in vision occurred: awakening aware­ness of environmental long-term problems has been bringing forth a move away from treating environmental problems only after they have occurred (typically at the end of service life of a facility or site). The current vision is to prevent environmental impacts from the beginning in the life cycle of a facility or activity. This life-cycle management aims to treat each stage of an operation not as an isolated event, but as one phase in its overall life. Thus, the planning covers not only each stage, but is a continuing activity, taking into account actual and projected developments. By implementing the elements of this vision, it is expected that the generation of contami­nated sites as well as the need for expensive remediation programmes will be minimized.

National institutions need timely and accurate information on available remediation strategies and technologies, management options as well as guidance in dealing with non-technical factors, e. g., communica­tions and stakeholder involvement. To resolve environmental liabilities and to avoid the generation of new contaminated sites, the IAEA and other international organizations help countries to adopt appropriate practices.

11.1.2 Legislation

After separation from the Soviet Union, Ukraine has implemented a well — developed legislation and regulatory framework for its nuclear industry. Radioactive waste management is carried out in accordance with the laws and other legal acts of Ukraine. These documents can be divided into three

levels:

• international agreements, laws and resolutions of the government of Ukraine;

• the system of special rules and regulations to ensure safety during RAW management, which are based on the laws and establish the procedure for certain types of activities;

• operational documentation, instructions, procedures and regulations.

The basic laws of Ukraine, which regulate the activity for RAW manage­ment are:

• The law of Ukraine ‘On the Use of Nuclear Energy and Radiation Safety’ (1995).

• The law of Ukraine ‘On Radioactive Waste Management’ (1995).

site clean-up

L. W E R M E, Consultant, USA

DOI: 10.1533/9780857097446.2.438

Abstract: The chapter describes the historical background to the current radioactive waste (RAW) situation in the Nordic countries. It discusses the current management and final disposal of low level waste (LLW) and intermediate level waste (ILW) and the siting processes for a repository for spent nuclear fuel. Early nuclear activities in Sweden led to contaminated nuclear facilities and uranium mining sites. The chapter describes the ongoing remediation of these sites.

Key words: nuclear waste, repository siting, uranium mining, site clean-up, legal framework.

13.1 Introduction

The atomic bombs dropped over Hiroshima and Nagasaki alerted Sweden to the potential of nuclear energy. Until then, the programmes for nuclear physics research had been very limited. In the autumn of 1945, however, the Swedish Defense Research Establishment (FOA) asked for funding for preliminary studies. The military thought that it would be useful for a small country to possess an atomic bomb as a deterrent (Jonter, 2002). The peace­ful aspects of nuclear energy were, however, most important. In 1945 a committee, Atomkommitten, was formed. Its task was to plan future nuclear research and to find applications for peaceful use of nuclear energy. The committee came to the conclusion that the government should develop this new power source in cooperation with industry and in 1947 AB Atomenergi was constituted with the government as the main shareholder (Larsson, 1987; Elam and Sundqvist, 2006).

Much of the initial research was concentrated on producing uranium and separating plutonium from irradiated uranium. The idea was that Sweden should become independent and self-sufficient in energy supply. The uranium was to be mined from the shale deposits in south Sweden. Any import of uranium was at that time out of the question. With a limited supply of uranium, the solution was a heavy water reactor with natural or low enriched uranium. The heavy water was to be imported from Norway.

The first Swedish reactor, R1, put into operation in July 1954, was con­structed underground at the campus of the Royal Institute of Technology (KTH) in central Stockholm. It was fuelled, however, with uranium bor­rowed from France. The agreement was that Sweden would return the uranium as soon as the Swedish uranium mines had gone into production.

R1 was a research reactor intended neither for energy production nor for plutonium production. Therefore, a second step was planned. In a sparsely populated coastal area with access to water, one or more reactors were to be built. The final location was Studsvik, where AB Atomenergi built its research centre. This was also the location of R2, a materials testing reactor, which was started in 1961 (see Fig. 13.1).

During the latter part of the 1950s, following a conference in Geneva, the nuclear weapons countries made available on the market enriched and natural uranium. The Swedish government issued a nuclear energy law in 1956, which allowed the development of nuclear power. This boosted the Swedish national nuclear programme and AB Atomenergi proposed the construction of two more reactors, R3 and R4. A group of private power companies had already in 1955 formed a consortium, Atomkraftkonsortiet (AKK). Their purpose was to follow the international development, propose reactor types and finally build a nuclear power plant (NPP) for the owners. AKK was first to propose light water reactors in Sweden.

The government policy was, however, still heavy water reactors and domestic supply of uranium. The programme was very optimistic, but it soon became obvious that the country did not have the means to carry it out. Of the originally foreseen five to six heavy water reactors built before 1965, only one was built and started in 1963 in Agesta in southern Stock­holm. This reactor was mainly used for district heating and operated until 1973. While Sweden concentrated on the heavy water line, light water reac­tors were developed in the US. The development of boiling water reactors (BWR) and pressurized water reactors (PWR) was rapid, while the Swedish national nuclear programme ran into difficulties. The programme included uranium production, fuel factories and reprocessing facilities. In 1965 the uranium production facility in Ranstad was opened. At the time, however, the cost for uranium from Ranstad was considerably higher than the world market price. Mining stopped in 1969 and the facility was closed in the early 1970s.

Vattenfall and AB Atomenergi had cooperated in building the Agesta reactor and were now planning a larger reactor in Marviken. The reactor design was changed several times and finally it was decided it should have a power of 400 MW and also be used for plutonium production. At the same time, AKK decided in 1959 to build a small BWR north of Oskarshamn. In 1965 AKK was transformed into Oskarshamns Kraftgrupp AB (OKG), and a BWR reactor was finally ordered in 1966 and in

operation in 1973. Following the order for the first reactor at Oskarshamn, 11 more reactors were ordered and put into operation during the following two decades.

The Marviken reactor was ready for test operation by 1968. That year Sweden had signed the non-proliferation treaty and there was no longer any reason for plutonium production. The result of a government investiga­tion published in 1968 led to the formation of ASEA-ATOM (1969), owned half each by ASEA1 and the state. The reactor design and nuclear fuel activities were transferred to this new company. AB Atomenergi continued as a research institute. This marked the end of the Swedish national nuclear programme. The Marviken reactor was never started. The legacy of the programme, however, was a uranium production facility and a pilot facility for reprocessing of nuclear fuel.

During the 1970s, the use of nuclear energy became increasingly contro­versial. In 1977, a new law required that the nuclear industry demonstrate how the nuclear waste was to be taken care of before any reactor could be fuelled. This led to the launching of the project Karnbranslesakerhet (KBS). The project finally resulted in the fuelling of all the reactors finalized after the law became effective. In 1981 a new law required that the nuclear power companies fund the future costs of nuclear waste management. The industry delegated a company, Swedish Nuclear Fuel Supply Company (SKBF, founded in 1973), jointly owned by Sydkraft AB (now E. ON Karnkraft Sverige AB), Vattenfall AB, OKG Aktiebolag and Forsmarks Kraftgrupp AB, to perform the necessary research and development work. The company name was later changed to the Swedish Nuclear Fuel and Waste Manage­ment Company (SKB).

At this time, Sweden has in operation three BWR in Oskarshamn, three BWR in Forsmark and two BWR and two PWR in Ringhals. The two reac­tors in Barseback were closed in 1999 and 2005, respectively.

The Finnish situation was different from that in Sweden. Finland had been on the losing side in the Second World War and had to cede 10% of its territory to the Soviet Union and was, furthermore, obliged to pay 300 million dollars in war reparations to the Soviet Union, following the 1947 Paris Peace Treaty. The loss of Karelia also meant the loss of important hydropower plants. After the war, Finland only had about two-thirds of its hydropower left compared to the situation before the war. Finland lacked both energy and economic resources to embark on a nuclear research pro­gramme. Furthermore, the Paris Peace Treaty prohibited Finland from research and development of nuclear weapons materials. That, and the lack of resources, meant that Finland did not invest in a research reactor as early as the other Nordic countries. The expanding Finnish industry, however, needed electricity and the potential of nuclear energy for electric power generation was recognized early (Anttila, 2000; Kojo, 2006).

After President Eisenhower launched the ‘Atoms for Peace’ initiative and the possibilities of having access to nuclear energy without a costly

XASEA (Allmanna Svenska Elektriska Aktiebolaget) bought the state’s share of the company in 1982. After ASEA merged with the Swiss Brown Boveri Corporation forming ABB (ASEA Brown Boveri), the name was changed to ABB Atom. Since 2000, the company is part of the Westinghouse Electric Company as Westinghouse Electric Sweden (2003).

domestic development programme, the Finnish Academy of Science sug­gested the formation of an Atomic Energy Committee in 1954, which the Finnish government then set up in 1955. The tasks of the committee included investigating the suitability of nuclear energy in Finland. Parallel to this, the industry established a company, Atomienergia Oy (Atomic Energy Ltd), primarily to satisfy the interests of the forestry industry. A milestone in Finnish nuclear research was taking into operation a TRIGA-type research reactor in 1962. This reactor is still in operation.

In 1965, the state-owned energy company, Imatran Voima (IVO, now Fortum Power and Heat Oy, FPH) put out to tender for a nuclear power plant to different suppliers, and in 1969 IVO decided to buy from the Soviet Union. Two PWR-type reactors were ordered and put into operation in 1977 and 1980 at Hastholmen near Loviisa. In 1969, the private industrial companies formed Teollisuuden Voima Oy (TVO) and the following year TVO decided to build two reactors. These were BWR reactors from ASEA ATOM and they were taken into operation in 1978 and 1980 in Olkiluoto. At present, TVO is having a third reactor constructed. Finland is, therefore, the only Nordic country expanding its nuclear energy capacity with new reactors.

There is no commercial electricity-generating reactor in Norway. In fact, over 99% of all electricity in Norway is produced by hydropower (OECD — NEA, 2005). Norway has, however, been very active in nuclear research. This started immediately after the Second World War, first at the Norwegian Defense Research Institute (FFI) from 1946 and later at the Institute for Nuclear Energy (IFA, now Institute for Energy Technology, IFE), which was founded in Kjeller in 1948. Norway’s and also Scandinavia’s first nuclear reactor, JEEP, was started at IFA as early as in 1951 (Oberlander et al, 2009; OECD-NEA, 2005). In all, there have been three research reactors at IFA, JEEP (1951-1967), NORA (1961-1967) and JEEP II (1967-). In addition to these, there is a fourth research reactor, the Halden boiling water reactor (HBWR) in Halden.

As was the case for Norway, there is no commercial electricity-generating reactor in Denmark. Three research reactors, DR 1, DR 2 and DR 3, have been operated at the Ris0 National Laboratory. They were started between 1957 and 1960 and are now all shut down, and DR 1 and DR 2 are fully decommissioned (Dansk Dekommissionering, 2006, 2009).

Radioactive waste can be classified according to different schemes for dif­ferent purposes such as operational segregation and treatment, e. g. com­pressible waste and combustible waste. At an international level, an agreed system of classification has been developed based on long-term safety con­siderations [3]. The scheme is used at a national policy and strategy level for exchange of information and for the purposes of international safety standards and international safety conventions. The scheme is based on linking waste types to corresponding disposal options. The waste classes defined are summarised in Box 3.1.

The classification scheme is illustrated graphically in Fig. 3.1. The ordi­nate is the radioactivity content of the waste and the abscissa the half-life. The diagram illustrates the need for greater levels of containment and isola­tion for higher activity and longer lived radioactive waste.

Conventional industries generally produce large volumes of residues con­taining naturally occurring radioactive materials (NORM), of the order of 104-106 t/a. This necessitates a different, pragmatic approach from typical RAW management, for which the principle of concentration and contain­ment is used [28] . For most NORM residues containment is not possible, and in many cases it is not waste but a useful recyclable residue. Therefore, for NORM residues the principle of dilution and dispersion should be pre­ferred wherever possible. It saves resources of other materials and it reduces waste volumes. Furthermore, one should keep in mind that concentration/ containment and dilution/dispersion are complementary, not contradictory, concepts. Processing of NORM waste consists of pile stabilization by various processes in order to increase the safety of storage and disposal sites. Large solid pieces of NORM waste, such as pipes from the oil industry, are frag­mented for handling and transport purposes. Liquid effluents are generated at all stages of the uranium production cycle that use process water and chemicals, including crushing, grinding, leaching, precipitation and tailings disposal and management. In addition, leaching of ore and mineralized waste rock by groundwater and surface water, respectively, can result in generation of acid mine water, which must also be contained and treated. The effluents contain radioactive and non-radioactive elements and com­pounds that, if not properly contained, can contaminate drinking water resources or enter the food chain, potentially harming the environment and endangering the health and well-being of human populations. Scales and sludges, which are generated in small volumes but which may have activity concentrations reaching very high levels, such as those originating from the oil and gas industry, are usually held in storage pending the establishment of suitable disposal facilities [28].

Criteria for exemption, without further consideration, of substances con­taining radionuclides of artificial origin are based on the premise that exemption will be the optimum option when the dose incurred by an indi­vidual is of the order of 10 pSv or less in a year [11]. For NORM, the situ­ation is quite different. Owing to the existence of significant and highly variable levels of background exposure to radionuclides of natural origin, exemption is likely to be the optimum option over a much wider range of doses, typically doses of the order of 1 mSv or less in a year.

The use, reuse and recycling of NORM residues and NORM contami­nated items — including, where appropriate, the dilution of NORM residues to reduce the activity concentration — is now starting to be recognized as a legitimate and desirable option for minimizing the quantities of NORM that need to be disposed of as waste. In particular, the beneficial (and safe) uses of phosphogypsum as a co-product of fertilizer production are now very much in the spotlight and, in some countries at least, there is already evidence of a shift in regulatory attitude towards this approach. However, when considering the use of NORM residues in the construction of dwell­ings, as a component of either landfill material or construction material, the possibility of increased radon exposure needs to be carefully taken into account.

• Cementation — the process of solidifying a liquid, sludge, solid, thermal residue, granular waste form, or calcine in cement matrix of crystalline calcium silicates, aluminates, and ferrate.

• Geopolymerization — the process of solidifying a liquid, sludge, solid, thermal residue, granular waste form, or calcine in an amorphous sodium aluminosilicate matrix.

• Bituminization — the process of solidifying a liquid, sludge, solid, thermal residue, granular waste form, calcine in bitumen.

• Forming — mixing a waste with cementitious, geopolymeric, bituminous, hydroceramic, or Ceramicrete-type additives and mixing in a form, i. e. can, vault, canister, and allowing the material to set or age.

• Pouring — similar to forming but the waste/additive mixture can be poured, extruded, or emptied into a form to set or age.

• Compositing — using metals, glass, cements, geopolymers, etc, to encap­sulate a waste that has already been solidified for special reasons such

RAW conditioning, immobilization and encapsulation 185

as heat dissipation, control of respirable fines in calcined or granular waste forms, and/or compressive strength requirements.

Often processes are coupled. For example, in France and the UK waste is calcined to remove excess nitrates before vitrification into a final waste form. This allows free-flowing oxides to enter the melter without nitrates being off-gassed or causing the particles to adhere to one another. Organic bearing wastes are often pyrolyzed to remove organics, if needed, before vitrification [14, 15].

Calcining is often performed before HIP, CIP, HUP, or CUP processes are performed, so that volatile species are not given off during the hot pressing or during the subsequent sintering. This ensures that the pressed waste form retains its integrity and form and does not crack during process­ing from off-gassing of hydrated or nitrated species.