Vitrification is the process of choice for separated highly radioactive wastes in virtually every reprocessing nation. (Donald et al., 1997; Ojovan and Lee, 2005; Vienna, 2005; Donald, 2010) Vitrification is:

• a proven process,

• tolerant to a wide range of waste compositions,

• a fast continuous process,

• generates no fine particulates, and

• the US Environmental Protection Agency (EPA, 2009) best demon­strated available technology (BDAT). Vitrification produces a waste form of good performance that is well understood (including many natural and ancient man-made analogs).

While vitrification into a borosilicate glass is the reference process, the next generation waste forms for HLW with potential benefits over vitrifica­tion are being developed. For example, glass composite materials (GCMs) including glass ceramics may allow for significantly higher waste loading than possible in typical borosilicate glasses (Ojovan and Lee, 2011). There are three primary limitations to the loading of HLW in glass: (i) decay heat, (ii) solubility of waste components (e. g., MoOtt, and (iii) noble metals. GCMs could allow for higher heat as the crystalline portions may be much more thermally stable. They also are expected to tolerate significantly higher concentrations of components that are sparsely soluble in the glass melt. The noble metals limit would depend on the processing methods, but, will not be more restrictive for GCMs. Crum et al. (2012) developed durable, radiation — resistant glass ceramics that could be processed using existing melter tech­nologies containing roughly double the waste loading of typical glasses. Figure 5.6 shows graphically the potential for the increases in waste loading.

Other potential waste forms for HLW are crystalline ceramic waste forms, which show promise for high loading and high chemical durability (Burakov et al., 2010). Development of the synthetic rock (Synroc) types of waste forms began in the 1950s; the term Synroc was coined by Ringwood et al. in 1979 when the most concerted waste form development and testing on these forms began (Ringwood et al. , 1979). Ranges of silicate, aluminate, and phosphate ceramics were developed in the 1960 to 1980s. Excellent reviews of these waste forms already exist (Lutze and Ewing, 1988; Donald, 2010; Burakov et al., 2010). A number of recent advancements in these materials have shown that complicated processes such as alkoxide

precipitation and hot isostatic pressing could be replaced with a simpler melt-cast-type process (Vance et al., 1996; Advocat et al., 1997; Stefanosky et al., 2009).

With the advanced separations methods currently available, it is possible to subdivide the HLW raffinate into streams with similar chemical proper­ties such as lanthanides, alkali and alkaline-earths, transition metal fission products, etc. Each of these streams could be separately immobilized in a form specifically design for the waste chemistry and disposal environment. A cost-benefit analysis was performed to evaluate the value of separating the HLW raffinate into constituent streams (Gombert et al., 2009 ). It was concluded that, aside from the noble metals, there was not a strong cost driver to further segregate the HLW. In the case of noble metals, there was a case for treating them separately under some circumstances. Thus, any further separations would be implemented for reasons other than cost.

In the absence of severe localized corrosion conditions, nickel-based alloys containing chromium are protected against fast corrosion by a chromium — rich oxide adherent film commonly known as a ‘passive film’ at the exposed surface. Typical examples are the thin, adherent passive oxide films observed on sample surfaces after short-term polarization tests and long-term immersion tests (Orme, 2005; NWTRB, 2002). Film thicknesses were in the range of a few nanometers (10-9 meters, nm, 3.9 x 10-10 inch) and tended to be rich in chromium (III) oxides (Cr2O3 and/or NiCr2 O4). A thick outer layer was also observed on top of the inner chromium-rich oxide layer. The outer layer was typically porous and consisted mostly of nickel oxide and the oxides of some other alloying elements. The chromium-rich oxide is considered to protect the bare metal against rapid corrosion in the long term, i. e. geological timeframes. A cross-sectional view of the passive film formed on the surface of an annealed nickel-based alloy is presented in Plate II (between pages 448 and 449). It is important to understand whether or not the passive layer persists for a long period of time (Ahn et al., 2008a). A number of issues have been studied to determine whether the protective layer remains stable in the long term. For example, if the protective layer grows continuously, the stress may build up at the interface of the bare metal and the protective layer, and the protective layer may spall off. However, the subsequently exposed bare metal would repassivate. Certain metalloids such as sulphur may be segregated at the interface during the anodic dissolution of the bare metal surface. A potential mechanism of the breakdown of the passive film induced by enrichment of sulphur at the metal-passive film interface is presented in Fig. 7.1 (Marcus, 19952 . When the surface concentration of the segregated sulphur exceeds a critical value, the protective layer will become unstable. The bare metal exposed as a result of the unstable protective layer may repassivate after dissolution of the accumulated sulphur layer. Other impurity elements such as silicon in the alloys or solutions may also affect the long-term stability of the protective layer. Microbially-influenced corrosion may also destabilize the protective layer. However, the bare metal surface formed after the desta­bilization of the protective layer could repassivate.

An important related issue is the accuracy in measuring very low general corrosion rates. General corrosion rates on the order of nm/year are

difficult to measure accurately. The accuracy is important because the rates must be extrapolated to a very long time period to calculate the extent of general corrosion and assess when the package would fail.

Once the passive film becomes unstable without repassivation during disposal, either high general corrosion or localized corrosion such as crevice corrosion or pitting corrosion would occur. For localized corrosion to be initiated, if there is no existing (propagating) pitting or crevice corrosion, the corrosion potential needs to reach the breakdown potential for highly corrosion-resistant alloys (such as nickel-based alloys) (Ahn et al., 2008b, 2013; ASM International,1993). This condition is determined by the severity of the evolved groundwater chemistry. More conservatively, at the corrosion potential below the repassivation potential, even the existing (propagating) pitting or crevice corrosion would be arrested. The breakdown potential or repassivation potential generally increases with higher concentration ratios of oxyanions such as nitrates to chloride (Dunn et al., 2005). Even if the localized corrosion occurs, it is not expected to open up entire areas of a container surface. The cathodic capacity of the outside of an active crevice or pit, from the separated cathodic area from the active area, would limit localized corrosion propagation fronts (Shukla et al., 2007). Some studies show only pit growth rather than uniform dissolution in the crevice area of highly corrosion-resistant alloys (Ahn et al., 2008a). Based on the cathodic capacity limitation, a maximum of 20% of the surface area is likely to be open (He et al., 2011).

One of the points that should be explored is to what extent a regulatory framework is of utmost importance for the implementation of ER pro­grammes. But this is not enough. On top of that, regulatory requirements must be well understood by all the sides involved, something that is espe­cially challenging when one takes into account the philosophical elements embodied in the radiation protection principles. Despite international rec­ommendations, final, mandatory decisions are taken in political and judici­ary environments that do not necessarily possess the proper technical background, often leading to total removal of the contamination and exces­sive (and unnecessary) expenditures (greenfield rather than brownfield and redevelopment). It may be useful to discuss the regulatory differences in different countries (on the basis of economics or social-cultural-political environments). How best to transfer the experience from one country to another? How to establish a better flow of information between scientific community, regulators and industries? Are the industries aware of, and do they possess, a good understanding of the rationale behind the regulatory requirements? How do industries perceive the existing regulatory frame­work? How to improve co-operation between regulators and other players? To what extent should international guidance (e. g., from the IAEA) be tailored to individual countries? How much flexibility should be allowed for reference doses? Prescriptive (e. g., radioactive concentrations) v. non­prescriptive (e. g., cost-benefit analyses) approaches is a crucial issue.

RAW of nuclear energy sector

NPPs are currently the main radioactive waste producers in the Ukraine, producing liquid and solid RAW. The main sources of primary liquid radioactive waste (LRAW) from NPPs are pipelines leakages, water from spent nuclear fuel storage pools, solutions remaining after sorbent regen­eration and spent decontamination solutions. The product of primary LRAW reprocessing (except for spent filtering materials and sludge) is concentrated salt solutions, which are exposed to deep evaporation, result­ing in a fusion cake.

The volume of LRAW at Ukrainian NPPs per 109 kW h of the generated electricity is evaluated to be (Shestopalov et al, 2008):

• evaporation bottoms: about 13 m3

• filtering materials and sludge: about 8 m3

• fusion cake: about 2.4m3

• oil and mixed solutes: 0.45 m3

The isotope composition of LRAW is mostly: 90Sr, 90Y, 134Cs, 137Cs, which come from untight heat-emitting elements, and 58Co, 60Co, 54Mn, 59Fe, 51Cr, 124Sb, which are formed by neutron activation of pipes and contour corro­sion products.

Specific activities of LRAW are within the following range:

• evaporation bottoms and fusion cake: from 1010 to 1011 Bq m-3;

• filtering materials and sludge: from 109 to 1010Bqm-3.

Table 11.3 presents the data related to LRAW accumulated at Ukrainian NPPs as of mid-2008 (NatRep, 2008).

The sources of solid radioactive waste (SRAW) from NPPs are worn-out equipment, apparatus and instruments; dismantled equipment and pipe­lines, construction materials and debris; used individual protection means; elastron, electric — and heat insulation materials; ventilation system spent filters, and sludge from treatment facilities. In addition, NPPs store spent ionizing radiation sources.

The isotope composition of SRAW waste is mostly: 137Cs, 134Cs, 60Co, 90Sr, 95Nb, 54Mn, 51Cr, 59Fe.

The volume of SRAW at Ukrainian NPPs per 109 kW h of the generated electricity is evaluated to be (Shestopalov et al., 2008):

• low level waste (LLW): about 30 m3

• intermediate level waste (ILW): about 1.0 m3

• high level waste (HLW): about 0.1 m3

Table 11.4 presents the data related to SRAW accumulated at Ukrainian NPPs as of mid-2008 (NatRep, 2008).

Non-nuclear sector RAW

The State Interregional Specialized Plants (SISP) of the Ukrainian State Association ‘Radon’ (UkrSA ‘Radon’) deal with collection, transportation, storage and disposal of RAW from Ukrainian enterprises, medical and research institutions, including ionizing radiation sources. SISP are located near Dnipropetrovsk, Kyiv, Lviv, Odessa and Kharkiv (Fig.11.1).

Table 11.5 presents the data related to radioactive waste accumulated at SISP of the UkrSA ‘Radon’ as of mid-2008 (NatRep, 2008).

Two research reactors in Ukraine (Fig. 11.1) store liquid and solid radio­active wastes on site:

• reactor WWR-M of the Nuclear Research Institute of the National Academy of Sciences of Ukraine (Kyiv);

• reactor IR-100 of the National Institute for Nuclear Energy and Indus­try (Sevastopol).

RAW in the Chernobyl Exclusion Zone

As a result of the accident at unit 4 of ChNPP, a large quantity of radioac­tive material was released and distributed over a huge territory. Most of these materials remain inside the unit above which the ‘Shelter’ object has been built, and within a local area (the so-called ‘Shelter’ object site) that surrounds the ruined unit and ChNPP site. According to a first estimate, the

nuclide composition of this radioactive waste corresponds to the nuclide composition of irradiated nuclear fuel of RBMK-1000 reactors with an average burn-up of 11,000 MW day t-1. At this rate, the ratio of activities of gamma — and beta-emitting nuclides to alpha-emitting nuclides is approxi­mately 100 to 1.

Two main radioactive waste sources can be defined in the Chernobyl Exclusion Zone (ChEZ):

• radioactive waste generated during the operation of the four ChNPP units (Table 11.6);

• radioactive waste resulting from the accident at unit 4. Accidental wastes are located in the ‘Shelter’ object (Table 11.7) and in radioactive waste disposal points (RWDP) and radioactive waste temporary storage points (RWTSP) (Table 11.8).

deposits, underground water) of the radioactively contaminated landscape. Table 11.8 gives the characteristics of waste in RWDP and the RWTSP.

Outside the ChEZ, in the Kyiv, Zhytomir, and Chernigov regions, there are several waste storage facilities built as a result of decommissioning and remediation activities in these areas. Their total volume of waste is 171,000 m3 and the total activity of these materials can be between 1 and 2 x 109Bq.

Another concept is that of very deep (permanent) disposal. In this concept (Gibb, 2000), the waste is located at depths of 3 km or more and as such any transport of radionuclides through the geosphere is extremely limited (Fig. 1.24) . Further, if located in suitable (granitic) rock, the radiogenic heat from HLW can cause reaction with the surrounding rock and lead to creation of a sarcophagus or granite coffin which seals in the waste permanently.

The US BRC was positive about the deep borehole disposal concept and the US is planning a demonstration programme. However, this is untried

technology that requires a case to be made for its safety that potentially could take many decades to come to fruition.

Safety is the prime consideration during management of RAW on account of the potential that exists for exposure of people to radiation. Such expo­sure can be to workers involved in the handling and management of RAW or to members of the public, due to any radioactive material associated with the waste being released into the publicly accessible environment. Similarly, environments contaminated with radioactive materials can cause exposure of persons to radiation. Both the management of RAW and contaminated environments can also lead to plant and animal species being exposed to radiation. Exposure can arise at the present time and can also occur in the future, and its magnitude can vary from insignificant to very high depending on the nature of the RAW and the circumstances of exposure. Exposure can also arise during normal anticipated circumstances associated with waste management and contaminated environments and from accidents or disruptive events.

The same philosophical basis for radiation safety has been adopted for all facilities and activities that can give rise to radiation exposure. However,

the manner in which this philosophical basis has been developed and applied to waste management and contaminated environments is influ­enced by the often long timescales involved and the desire to dispose of the waste, i. e. to no longer have to exercise active control and management over the materials. There is also need to differentiate those materials containing radioactive material but at such low levels that the material does not need to be managed as radioactive waste.

The effects of exposure to radiation have been studied throughout the twentieth century and a sound knowledge base has been developed [ 1]. Studies continue to refine and update this knowledge base, but in general the effects are known. Lower levels of radiation dose cause an increase in the incidence of cancer in the exposed populations and at higher levels of radiation dose in excess of a threshold in the region of 1 Gy, deterministic health impacts start to occur. The latter effects range from chromosomal aberrations to organ damage and skin burns to death at doses beyond a few Gy. The rate of cancer incidence increases with increasing radiation dose in a stochastic manner; in the lower range of doses, no increase in the natural incidence of cancers is detectable, at higher levels of dose in larger popula­tions, an excess incidence is discernible. The basic approach to radiation safety is both to prevent short-term deterministic health effects and to ensure that the longer-term risk of cancer induction is not significant.

Exposure to radiation can arise from radioactive material emitting pen­etrating radiation located outside the body of a person or other species, which due to proximity impinges on the body. Alternatively, exposure can arise from radioactive material being incorporated into the body, generally by inhalation or ingestion. Other diffusive transfer mechanisms generally apply to incorporation into plants.

All these factors have to be considered in developing and applying a safety regime for the management of RAW and contaminated environments. The fundamental approach is to reduce the volume of waste to the extent reason­ably possible, to solidify it into an immobile form, and to provide measures to contain and isolate the waste from the accessible environment. The con­tainment is intended to keep the radionuclides within the containment boundary by chemical or physical fixation within the waste matrix and by physical containing barriers, and to provide shielding for any penetrating radiation emitted from the radionuclides within the waste [2]. The isolation function aims to keep the radionuclides away from people and the environ­ment and also to protect the waste and its protective features from disturbing and degrading influences such as fire, water, physical disruption, etc. The timeframes required for such containment and isolation are influenced by the radioactive half-lives of the radionuclides contained in the waste.

Many activities involving radioactive material processing, handling and use also give rise to contaminated effluents. Treatment of the fluids generally involves cleaning by filtration, solvent extraction, ion exchange or by evaporation. The aim of these cleaning processes is to reduce the radio­activity levels in the effluent to the extent that they can be safely discharged into the environment.

This chapter outlines the general international principles of radiation, waste and transport safety. Examples of their application in various coun­tries are given in Part II.

Some solid and liquid wastes may contain bio-hazardous or infectious ma­terials. Further to radiological protection, other precautions for handling these wastes should be respected. When processing bio-hazardous wastes, their infectious features, and tendency to putrefaction, to insect attacks and to microbial degradation must be controlled. Clearance of bio-hazardous waste from radiation regulatory control is unlikely to mean that this waste is also exempt from bio-hazardous waste regulatory control. The goals of treatment of bio-hazardous waste are the following: (a) biologic detoxifica­tion; (b) prevention of biological degradation; (c) volume reduction.

An important step in the treatment of bio-hazardous waste is neutraliza­tion of biological hazard. It can be done by sterilization. A number of steri­lization methods are regularly used in hospitals and they can be applied for treating bio-hazardous RAW with some adaptation. Some other methods are aimed at volume reduction of the waste. Available treatment methods for bio-hazardous and medical radioactive waste have been described in detail [24]. Lidded containers lined with plastic bags are used for collection of wastes displaying biological hazards. Special consideration should be given to sharp objects. When possible, these items should be collected in puncture-resistant packages, properly labelled and treated separately. Most microbiologically contaminated laboratory wastes are suitable for steam autoclaving, but this method should not be used where the radioactive content of the waste is volatile during steam treatment. This method is not considered appropriate for most non-microbial pathogens, animal carcasses or parts. Chemical disinfection is useful for laboratory ware or similar ma­terials, but it is not suitable for pathological waste and animal carcasses or parts. Gamma irradiation is an attractive option for sterilization since it is appropriate for pathological waste, animal carcasses and parts. After deacti­vation or procedures aimed at preventing decomposition of its biological components, bio-hazardous waste can usually be treated using the same methods applied to non-biological radioactive materials in order to meet the waste acceptance criteria. Incineration is the preferred method for treating bio-hazardous RAW of animal or human origin, as well as organic chemical waste [11,21]. Incineration provides complete combustion of waste, produc­ing totally sterile residues, with any emissions from the stack being kept to acceptable environmental standards. Thermochemical treatment has been proven to be an effective method to treat animal carcasses producing totally sterile slag residue, with minimal off-gas emissions, the composition of which can be kept in line with acceptable environmental standards [11, 22, 23].

In cases where incineration is not available or the volumes of human and animal wastes are so low that it is desirable to treat them as they are pro­duced, it may be feasible to use maceration/pulverization to render these

materials liquid, so that they can be discharged via a liquid RAW route, including any necessary chemical deactivation to treat the biological hazard. Compaction and shredding are not considered viable for treatment of bio­hazardous solid waste. The primary reason for this restriction is that any microorganisms contained within the waste may be spilled or released during these processes and contamination may be widely dispersed.

Various thermal and non-thermal processes can be used to make various types of waste forms. The cross-referencing of the various processing tech­nologies that can be used to produce various waste forms is given in Table 6.2 . In Table 6.2 the technologies are designated as continuous processes which are more applicable to large volume wastes, or batch processes which are more applicable to small to medium volume wastes. Table 6.2 provides a crosswalk of the processes/technologies (rows) that can be used to form various waste forms (columns). A list of advantages, disadvantages, and types of waste form that can be made with a particular technology are also given in Table 6.2. Further discussion of the individual technologies (rows) can be found in Chapter 4 of Reference 11.

The processing and pre-processing technologies by which a waste form can be made are briefly described below as thermal or non-thermal tech­nologies in keeping with Table 6.2.

Table 7.1 summarizes the dissolution rates for oxidizing and reducing dis­posal environments (Ahn et al. , 2011a) used in a performance assessment model (Markley et al., 2011). A range of environmental conditions are con­sidered, mostly near-neutral pH and ambient temperature. The variation of pH and temperature can be adjusted in terms of dissolution rate as user — defined parameters. For this base case, radionuclide release is estimated combining the reducing and oxidizing environments, to simulated residual radiolysis of water by actinides in the reducing environment. Figure 7.8 shows the estimated dose from the radionuclide release for this combined case.

Considering all radionuclide release fractions from the UO2 matrix, an exercise was conducted to estimate the doses to workers or members of the public from airborne fragments of the SNF matrix caused by SNF oxidation and SNF drop/collision (after Kamas et al., 2006). The most significant dose contributor in the release fraction is aerosol SNF fines (i. e., small solid particles). Tritium, noble gases, iodine, crud, ruthenium, caesium, strontium and SNF fines were part of the source term considered. In Fig. 7.11. the

Table 7.1 Summary of SNF dissolution rates in oxidizing and reducing environments

Parameter name Value Description and basis

Spent MOX fuel is also included in the table and the reference numbers quoted are from the reference by Ahn et al. (2011a).

7.11 Example dose estimate for (a) oxidation and (b) collision (/drop) of SNF assemblies (after Kamas et al., 2006). Used with permission from American Nuclear Society (ANS).

radionuclide release fraction of the aerosol SNF fines, 2.0 x 10-6 for the drop/collision case and 1.2 x 10-3 for the SNF oxidation case, were used to estimate the dose to workers or members of the public (Ahn et al., 2011b; Kamas et al. , 2006). A site boundary was defined, for the dose to workers within the boundary and to members of the public outside the boundary.

The left figure is for SNF oxidation under normal operations. The wake effects are a modification of the radionuclide transport path right outside any storage building if any building shadow exists. Consequently, radionu­clide transport will stop. Within a short distance from the building, the radionuclide transport will not be reached. The right figure is for drop/col — lision cases. In both cases, arbitrary dose rate units are used for the log scale. The oxidation case gives a dose rate ten times higher than the collision case in the same log-scale unit.

Towards the end of the 1950s, along with the formation of the Eastern Urals Radioactive Track, the process of RAW accumulation began in the

territories where scientific and research institutes were located, mainly in the large cities in the central part of the USSR. Furthermore, the develop­ment of the nuclear industry helped to solve a wide range of medical and technological problems with the help of radioactive materials. Two new systems were almost simultaneously implemented by government resolution.

The first of these systems was ‘Isotope’, an All Union association estab­lished in 1958 with the aim of revolutionizing isotopic production, radiation techniques, and medical and general purpose equipment.

Isotope had the following functions:

• delivery of isotopic products for external and domestic markets;

• addressing radiation safety problems during handling of ionizing sources

(IS);

• supplying medical and scientific establishments with the required iso­topic products;

• development of radioisotopic technologies.

The establishment of this organization helped to solve many of the prob­lems associated with the introduction of new nuclear technologies and their influence on the national economy.

At the same time, a centralized system for the collection and disposal of RAW and spent ionizing sources (SIS) was created, with 35 different

organizations involved (16 from the Russian Federation, 5 from Ukraine, and 1 from each member republic of the USSR). For example, in February 1960 the Council of Ministers of the USSR created an organization called ‘Radon’ in Moscow, which was designed to act as a central facility for RAW processing and disposal serving organizations in Moscow itself, the Moscow region and 10 adjacent regions. It began practical operation at the start of 1961, when the special vehicles column made its first journey to the Kurchatov Institute.

The introduction of these new specialized facilities for RAW and SIS collection and disposal stabilized the accumulations of RAW in scientific and production establishments across the USSR, as RAW removal began to be effectively and routinely carried out. For example, industrial, medical and research establishments in the central regions of the USSR sent the following quantities for further disposal:

• up to 2,500 m3 of solid radioactive waste (SRAW) with an activity up to

1015 Bq,

• up to 300 m3 of liquid radioactive waste (LRAW) with an activity up to

1011 Bq,

• up to 20,000 units of SIS with an activity up to 1016 Bq.

The sources were predominantly composed of 60Co (more than 90% of the overall activity) and 137Cs (up to 6%). Over almost 50 years, more than 100,000 m3 of RAW was removed from the Moscow area. These specialized enterprises also improved radiation control systems, developed monolithic matrix technologies as a product of RAW processing, and drew up new models and algorithms for safe RAW processing.