Regardless of the extensive work performed at an international level, the main responsibility for safety of spent fuel and RAW management remains with the operator/licensee and the control/oversight with the nationally des­ignated regulatory authority. The safe development, operation and closure (siting, design, construction, operation and decommissioning or closure and subsequent control) of facilities is to a large extent dependent on an ade­quate regulatory framework and its effective implementation in practice. The regulatory framework will take into account the specificity of the country, interfaces between operators and regulators, and also other stakeholders.

In general three legal and regulatory approaches have been adopted, namely:

• prescriptive regulations — with very detailed and specific safety require­ments (e. g., Russian Federation and the USA);

• non-prescriptive regulations — with goal oriented safety requirements (e. g., UK);

• a combination of the two above (e. g., Czech Republic, Bulgaria, etc.). In addition specific provisions may apply in a country, e. g.:

• categorisation of spent fuel as a radioactive waste or resource;

• application of the clearance concept (e. g., that is excluded in France);

• issuance of licences, permissions or both;

• issuance of one licence for the whole life cycle of the facility or multiple licences for each stage of the facility development (siting, design, con­struction, commissioning, operation, closure and decommissioning/ closure and institutional control).

All of the above aspects depend on the national policy and strategy adopted for radioactive waste management and nuclear development, the legal and regulatory framework within the country, past and current practices, finan­cial mechanisms in the long term and the capacity to perform the required oversight of the facilities at present and in the future.

Although it is beyond the scope of this chapter to describe the quantities, treatment, and disposal of secondary wastes, it should be pointed out that these wastes make up the largest volume fractions and relatively insignifi­cant radionuclide fractions of the waste produced in reprocessing. Mature technologies are employed to characterize, package, ship, and dispose of these secondary wastes worldwide.

Carbon steel is a corrosion-allowance metal that is expected to have a rela­tively low corrosion rate in a mild, near-neutral pH, and reducing environ­ment such as granite and clay (Jung et al., 2011). One localized corrosion process in carbon steel is pitting corrosion. This pitting process is empiri­cally represented by a ‘pitting factor’, which is defined as a ratio of pit penetration depth to the uniform corrosion depth. Therefore, degradation by pitting corrosion of a carbon steel container can be represented by adjusting the magnitude of the general corrosion rates.

Even in an underground repository, which is planned in the long term to have a reducing environment, it will initially have an oxidizing condition due to the excavation conducted before closure (which provides oxygen). The general corrosion rate of carbon steel in an oxidizing environment is very high: in the range of 10-100pm/year (3.94 x 10-4 to 3.94 x 10-3 inch/ year) at room temperature in simulated mild initial groundwater (Jung et al., 2011). During 30-year oxidizing conditions applicable for carbon steel corrosion, a general corrosion rate of 50 pm/year (1.97 x 10-3 inch/year) will
result in a small penetration depth of 0.15 cm (0.02 inch). Estimating the carbon-steel general corrosion rate using laboratory data and analogue data indicates corrosion rates in the range of 0.1-10 pm/year (3.94 x 10-6 to 3.94 x 10-4 inch/year) in the reducing environment (Yoshikawa et al., 2008; David et al., 2002). Figure 7.2 shows the data collected from Yoshikawa et al. (2008) from Japan and David et al. (2002) from France.

Carbon steel is susceptible to pitting corrosion in an oxidizing environ­ment. In addition, carbon steel is expected to corrode at a higher general

Period (year)

1000 [22]

7.2 Measured corrosion rates of carbon steel in simulated solutions and correlation with archaeological analogue data up to 1,000 years: the first analysis was conducted in Japan (a) (used with permission from Elsevier), and the second in France (b) (used with permission from Maney Publishing).

rate in an oxidizing compared to a reducing environment. However, the effect of pitting corrosion can be accounted for by using an enhanced general corrosion rate. The pitting factor, which is a ratio of pit propagation depth to general corrosion penetration depth, will approach unity during the oxidizing period as the general corrosion proceeds deeper. The pitting does not enhance the corrosion penetration at this point. An example case is shown in Fig. 7.3 (Johnson and King, 2000).

What technologies have proved to be effective? What needs to be improved? What are the innovations and how promising are they? Is there room or need for extensive R&D programmes? How willing are international or local contractors to use innovative technologies? What are the risks? The business factor can be measured in terms of inexpensive vs costly technolo­gies (e. g., in-situ measurements vs. laboratory analysis). How to optimize the relationship between costs and accuracy of measurements? ER plan­ning depends heavily on predictions about the behaviour of pollutants in different compartments (soil, groundwater, etc.) as well as on the design of remediation solutions. What are the challenges in mathematical models? How effectively and wisely are these being used? How to best educate and give training to potential modellers? What role does statistics play in pre — and post-clean-up characterization in reducing costs while ensuring safety?

Table 11.9 summarizes the data on total amount of radioactive waste accu­mulated in Ukraine by the middle of 2008. In the future, new streams of radioactive waste will originate in Ukraine. It is assumed that in accordance with international agreements, vitrified high-level RAW will arrive to Ukraine from the Russian Federation starting in 2013. These wastes are the product of reprocessing and subsequent vitrification of spent nuclear fuel from Ukrainian VVER-440 reactors.

According to Shestopalov et al. (2008), in the course of decommissioning and dismantling of existing Ukrainian VVER reactors, the following amounts of reprocessed solid waste will appear (in thousand m3): LLW: 34.4; ILW: 4.6; HLW: 2.4.

In the future, the rate of RAW accumulation in Ukraine will increase. This is due to Ukraine’s plans to increase the annual electric power produc­tion at NPP from 88 x 109kWh in 2005 to 220 x 109kWh in 2030 (Energy Strategy, 2006 ).

Managing uncontrolled releases and contaminated site clean-up are con­sidered in detail in Chapter 8 and Part III so only a summary derived from them is provided here. Agreed international safety requirements cover such situations (IAEA, 2003c). To ensure that protective measures can be quickly and efficiently implemented to mitigate the adverse effects of an accident or other forms of long-term contamination at a nuclear site requires good planning, clear strategies and a good managerial team. Preparations for environmental remediation should ideally be done in two phases: prelimi­nary planning, which should be available as part of normal operation, or emergency preparedness for each nuclear facility; and detailed remediation planning, which takes into account site (and accident where applicable) specific information.

According to the IAEA Safety Glossary (IAEA, 2007) remediation is defined as any measure that may be carried out to reduce the radiation exposure from existing contamination of land areas through actions applied to the contamination itself (the source) or to the exposure pathways to humans. An important element in the overall remediation concept is that complete removal of the contamination is not implied. Remediation aims to achieve optimised protection of the public, workers and the environment. The goal of remediation activities is the timely and progressive reduction of hazard and eventually, if possible, unrestricted release of the site. However, there are situations where this goal may not be achievable and then it must be demonstrated that as a minimum any unacceptable risks to human health and the environment have been removed. When choosing a remedia­tion option, a range of factors must be considered, such as the impacts on health, safety and the environment; and technical, social and financial factors. National remediation strategies are needed to specify, prioritise and to ensure remediation situations are managed in a manner commensurate with the risks associated with the contaminated areas and taking into account possible effects on neighbouring countries.

In general, remediation of a contaminated area involves preparation and approval of a remediation plan; remediation operations; and management of RAW resulting from the remediation activities. It needs to be based on collection and assessment of all available information of current and past activities at the site. Therefore an appropriate assessment of both the radio­logical and non-radiological impacts of the situation must be performed and the benefits and detriments associated with possible remedial measures, including the associated restrictions and institutional arrangements follow­ing remediation must be identified based on established reference levels as part of the decision-making process.

The remediation plan has to be subject to the approval of the regulatory body prior to its implementation and must state, as a minimum: the goal for the remediation; reference levels for remediation; the nature, scale and duration of the remedial measures to be implemented; the waste disposal or storage site, as appropriate; any post-remediation restrictions; and the monitoring and surveillance programmes and arrangements for institu­tional control for the remediation area. During the implementation of remedial measures, consideration must be given to (i) radiation safety, transport safety and waste safety, general health and safety issues and envi­ronmental issues so as to minimise hazardous impacts, and (ii) the potential for prolonged exposure after the termination of remediation activities.

The area has to be monitored and surveyed regularly during remediation so as to verify the levels of contamination; to ensure compliance with the requirements for site release and for waste management, and to detect any unexpected levels of radiation. Before an area can be released for unre­stricted use, a survey must be performed to demonstrate that the end-point criteria and conditions, as established by the regulatory body, have been met. The organisation responsible for the surveillance and verification of activities must be clearly identified. An appropriate programme, including any necessary provisions for monitoring and surveillance, has to be estab­lished to verify the long-term effectiveness of the completed remedial meas­ures. As part of the overall management system, arrangements for archiving, retrieval and amendment of all important records concerning the initial characterisation of the area, the choice of options for remediation and the implementation of remedial measures, including all restrictions and the results of all monitoring and surveillance programmes, must be established and maintained in all cases.

Radioactive waste can arise from a number of activities and facilities. It can occur in a very broad range of physical and chemical forms, and can have a similarly wide range of associated radiological properties. These factors influence the possible mechanisms of radiation exposure to persons and other species and the potential magnitude of such exposure.

One of the major sources of RAW generation is the nuclear sector, including both the commercial nuclear power industry and the military nuclear weapons manufacturing component. Whilst having completely dif­ferent objectives, the waste types generated have many similarities, arising from uranium mining and processing, enrichment, nuclear fuel manufac­ture, reactor operation, reprocessing and decommissioning. Production and use of radioactive sources for industrial, medical and other applications is another significant source of radioactive waste generation. The sources can be reactor produced, so the waste types have some similarities to the nuclear sector, or can be accelerator produced. Radioactive sources are generally of small physical size (i. e. < centimetres) but can vary significantly in terms of radiological properties — half-life, radioactive content and radiation type emitted. Radiation sources are widely used in medicine, industry and research. A number of scientific research and development activities use or generate radioactive material and can give rise to a broad and diverse range of RAW. The other area in which RAW arises is that involving naturally occurring radionuclides, generally associated with mineral extraction and processing. Numerous ore bodies and mineral deposits contain elevated levels of naturally occurring radionuclides, often linked to the phosphate industry, coal mining and oil extraction. Water treatment for domestic use can give rise to sludges with concentrations of naturally occurring radionu­clides that warrant management as radioactive waste.

As indicated, RAW can take many different forms, a factor influencing safety and hence the way in which the waste is managed. The waste material itself can be radioactive, it can contain radioactive material or it can be contaminated on its surfaces by radioactive material. A considerable amount of waste is generated in the form of solids, varying from granular mineral forms to solid rock to civil rubble to equipment, metals, plastic and paper. Also, contaminated liquids and gases are generated whose treatment can give rise to solid waste such as ion exchange resins, cemented or bitu — menised chemical sludges and filters used to clean liquid or gas streams contaminated with particulate and volatile radioactive species.

The majority of nuclear activities commence with mining and processing uranium/thorium-bearing ores, likewise such waste can arise from other mining and mineral processing activities. The radioactive species contained in these ores originate from the primordial radionuclides with radioactive half-lives of the order of thousands of millions of years. These species, iso­topes of uranium and thorium, each head decay chains of radionuclides with radioactive half-lives varying from microseconds to thousands of years. A decay product of particular interest is radon, the radioactive noble gas whose physical characteristics influence its instant mobility and related radiological hazard potential. Like any mining and mineral processing activity, the residues are waste rock, process tailings, chemical sludges and used plant equipment and buildings. Many process fluids are used, as mines often have to be de-watered and both mines and processing buildings are normally ventilated. Thus the spectrum of physical waste types generated takes the form of solids, liquids and gases. The amounts of waste generated are large; hundreds to thousands of tonnes of rock are mined to produce a single tonne of uranium. Mine sites are also generally quite large (i. e. up to tens of square kilometres) in area and due to the bulk nature of the materials handled, stored and processed, large areas of land and buildings become radioactively contaminated during operations. On the other hand, the radio­active concentration of the materials involved is not high — on the order of becquerels per kilogramme, although various adventitious concentration mechanisms can cause these concentrations to multiply thousands of times.

Spent nuclear fuel (SNF) may be considered either as waste (SFW), which will eventually be packaged and disposed of [25], or reprocessed to recover uranium and plutonium followed by conditioning of residue in the form of high level waste (HLW) containing mainly fission and activation products, and so-called minor actinides (Np, Am, Cm) [11, 25] . There is no specific treatment step of SFW except that fuel elements are stored at the reactor site for some time to allow their intense radioactivity to decay and associ­ated heat to decrease. SFW elements can then be moved to longer term storage facilities (dry or wet), before deep geological disposal. If not declared as waste, SNF elements could be shipped to a reprocessing plant, but only after a suitable storage (cooling) period. The decay/cooling storage period at reactor sites usually varies from three to five years or even longer; afterwards, the spent fuel can be transferred to ‘away from reactor’ storage for up to 50 years or more, depending on the national policies with regard to reprocessing or disposal.

HLW formed after reprocessing of SNF contains fewer long-lived acti­nides than SNF due to extraction of plutonium. The world industrial repro­cessing practice (as used, for example, at the La Hague Reprocessing Plant in France) demonstrates that the volume of HLW after conditioning is less than the total volume of SNF assemblies [26] . Both LLW and ILW gener­ated by reprocessing are treated by methods described in this chapter. HLW contains some long-lived fission products such as Tc-99 and I-129, and minor actinides (Np, Am, Cm). An additional technological procedure — partitioning — can be introduced into reprocessing technology for extraction of minor actinides to reduce the HLW radiotoxicity. The extracted minor actinides could then be transmuted by fission using fast neutrons. The par­titioning and transmutation (P&T) approach can reduce the radiotoxicity of SNF by a factor of 100 or more [27] resulting in less dangerous waste for disposal.

• Calcination — heating at elevated temperature to convert all cations to the oxide form (removes waters of hydration, hydroxides, nitrates in the presence or absence of air, i. e. rotary pyrolytic calciners). May be coupled with other high temperature processes.

• Drying — heating at 110°C to remove bound water in preparation for solidification, embedding or other high temperature processes.

• Vitrification — the process of solidifying a liquid, sludge, solid, thermal residue, granular waste form, or calcine in a glass (borosilicate, iron phosphate, aluminosilicate).

• Metal formation — melting a metallic waste with or without other metal additives.

Table 6.2 Waste form processing technologies

Processing Process technology mode

Treatment and waste stream scale

Thermal Joule Heated Continuous technologies Melter (JHM)

Large

Advanced Joule Continuous Heater Melter (AJHM)

Large

Cold Crucible Continuous Induction Melter (ССІМ)

Large

Waste forms produced

Disadvantages

Proven technology; typically operates with a ‘cold cap’ to minimize volatility of species of concern

Increased capacity, throughput, and melt rate

compared to JFIM

Allows processing of corrosive glasses; no refractories; no electrodes; water cooled; can be stirred if needed; increased capacity compared to JFIM and AJFIM; can operate at higher temperatures than JHM and AJHM; operates with a ‘cold cap’ to minimize volatility

Electrode and refractory erosion may be a problem; solubility control of certain species (Cr, Mo, and S04) critical

Operates with minimal or no ‘cold cap’ with associated increases in volatility of species of concern

Higher temperature operation can increase volatilization of species of concern but ‘cold cap’ coverage minimizes these impacts

Thermal In-Container Batch Depends on technologies Vitrification (ICV); also known as ‘Bulk Vitrification’ container size (could be mediurr to large) Self-Sustaining Vitrification (SSV) Batch Small Cold Press and Sinter (Cold Uniaxial Pressing, CUP: Cold Isostatic Pressing, CIP) Batch Small Hot Isostatic Pressing (HIP) Batch Small

Inexpensive and simple for low activity wastes or contaminated soils; not

applicable to HLW

Inexpensive; can be used to process small amounts of wastes at remote locations Higher waste loadings;

Minimum disposal volume

Zero off-gas emissions; higher waste loadings; minimum disposal volume; mature flexible

technology; no major secondary wastes; mature industrial process

Inhomogeneous waste forms produced; no temperature control so radionuclide vaporization is high; little or no convection in melt

May require some pre-processing, i. e. grinding of the waste and pre­mixing

Usually small scale; may require pre-calcining or pre-treating waste to an oxide to avoid shrinkage of form

Processes small quantities; can overpressurize if large amounts of volatiles (e. g. nitrates/ hydrates) are present; may require pre­calcining or pre-treating waste to an oxide (shrinkage handled by bellows like canisters)

Processing Process Treatment and technology mode waste stream scale Thermal technologies Hot Uniaxial Batch Small Press (HUP)

Cyclone Furnaces Continuous Large

Waste forms Advantages Disadvantages

produced

Borosilicate glass (lab scale only); GCM’s, crystalline ceramic, simple oxides, metal matrix, zeolites, hydroceramic Borosilicate glass, GCM’s, other glasses (LaB’s, FeP, AIP, chalcognide, etc.), crystalline ceramics, simple oxides, metal matrix

Higher waste loadings;

minimum disposal volume, mature flexible technology; mature industrial process

Suitable for soils containing low volatility radionuclides

Usually small scale; may require pre calcining or pre-treating waste to an oxide for shrinkage control

Secondary recovery process needed to treat off gases

Fluidized Bed Continuous

Steam Reforming (FBSR)

Electric Arc Batch Furnaces
Thermal technologies	Medium/Large
Plasma Furnaces Batch	Small

Pyrolysis (not incineration); immobilizes halides, sulfates,"Tc sequestered in sodalite; moderate temperature; >85% volatile species contained; wastes processed without neutralization; destroys organics and nitrates; industrially proven technology; no secondary liquid waste stream Established Industrial Practice; Similar technology is used for ICV

Plasma generating electrode erosion; efficient for the destruction of organics

Product is granular and requires a high integrity container (НІС) or

encapsulation in a binder to make a glass ceramic material, a geopolymer, or a hydroceramic; Radionuclide partitioning amongst the phases needs to be further studied

No large-scale radioactive practice; high temperatures; volatilization of radionuclides

Large-scale practice in Belgium

(Belgoprocess; high temperatures; volatilization of radionuclides [206-208])

Processing Process technology mode	Treatment and waste stream scale Small
Microwave Batch Heating
Cement Continuous or Batch
Non-thermal technologies	Large

Waste forms Advantages Disadvantages

produced

Suitable for mixed wastes; Can be used as a heat source in other equipment (e. g fluidized bed)

Simple technology; design formulation for best waste retention; fly ash and slag additives keep ssTc and Cr in reduced oxidation state to prevent leaching

Limited to small scale; process scale up; inhomogeneous heating (need a susceptor material); no large-scale practice

Formulations waste specific; some sequestering of radionuclides in hydration products vs. grain boundaries needs more study; radiolytic

production of H2 in high radiation; pH of pore water alkaline and promotes leaching

Non-thermal Geopolymer Batch

technologies

Hydroceramics Batch	Small
Ceramicrete Batch	Small
Continuous or batch
Bitumen	Large to small

Minimal water so radiolytic H2 generation is limited, fire resistant, pore water less alkaline than cements

High capacity for high sodium or calcium containing wastes; stabilize halides and sulfates.

Very dense; room temperature curing; high waste loading

Simple; low operating cost; leach-resistant characteristics

Formulations waste specific; distribution of radionuclides among the phases needs more study; batches are thick and require extrusion

Require hydrothermal set; requires more water than geopolymers so radiolytic H2 generation; batches are thick and require extrusion; wastes with >25wt% nitrate must be pre-treated

High heat of hydration; bubble formation which can be vibrated out of mixture during set

Flammable; requires heat to make bitumen molten; poor performance with salts; thick even when molten; requires extrusion

184 Radioactive waste management and contaminated site clean-up

• Pyrolysis — process of destroying organics in the absence of air (more environmentally compliant than incineration which destroys organics in the presence of air). Pyrolysis can be carried out in calciners, drums, or by fluidized bed steam reforming (FBSR).

• Hot isostatic pressing (HIP) — a manufacturing process used to reduce the porosity of metals and increase the density of many ceramic materials by subjecting the waste/additive mixture to both elevated temperature and isostatic gas pressure in a high pressure containment vessel.

• Cold isostatic pressing (CIP) and sintering — a manufacturing process used to reduce the porosity of metals and increase the density of many ceramic materials by subjecting the waste/additive mixture to isostatic liquid pressure in a flexible but impervious form such as a balloon before sintering at high temperature.

• Hot uniaxial pressing (HUP) — a manufacturing process used to reduce the porosity of metals and increase the density of many ceramic materi­als by subjecting the waste/additive mixture to uniaxial mechanical pres­sure from above and below in containment form while simultaneously subjecting the form to elevated temperature.

• Cold uniaxial pressing (CUP) and sintering — a manufacturing process used to reduce the porosity of metals and increase the density of many ceramic materials by subjecting the waste/additive mixture to uniaxial mechanical pressure from above and below in containment form before sintering at high temperature either with or without the containment form.

This section presents the performance of cladding in aqueous disposal environments and dry storage environments. Hydrogen-induced cracking of cladding may be a major detrimental degradation mechanism for both disposal and storage conditions. Crack opening area allows radionuclide release under both conditions. Oxidation (or general corrosion) of cladding is very slow and localized corrosion is unlikely to occur in near-neutral pH disposal environments (Ahn, 1996b). Oxidation of cladding is only possible in the presence of residual water and/or oxygen in dry storage canisters. Initially defective cladding may be further cracked (unzipped) by the pres­sure imposed on it by corrosion products of the SNF matrix or zirconium
itself. Longer longitudinal cracks that develop from the initial cracking/ unzipping will increase radionuclide release under both conditions.