NORM-containing waste is generally deposited in consolidated and over­covered piles or sludge beds, or purpose designed repositories with lined cells and protective capping [28] . As it is not feasible to move such large amounts of material, the waste tends to be disposed of on the site of its generation. Capping and some engineered structures may be used to prevent erosion and to limit the leakage of radioactive gases. In some cases, the waste has been disposed of by using it to backfill disused underground mines. There is growing evidence to suggest that bulk wastes contained in properly engineered surface reservoirs have very low radiological impacts. However, their environmental, safety and financial liability implications can be seriously underestimated. This has been demonstrated in the case of phosphogypsum stacks, where recent developments have suggested that the stacking option is not optimal and that more attention should be given to beneficial uses of the material [28]. Landfill disposal has been demonstrated as being an appropriate option for dealing with many types of NORM residue for which the quantities and activity concentrations are moderate, including most types of furnace dust with enhanced concentrations of 210Pb and 210Po. Normal landfill facilities are generally suitable, but the presence of non-radiological contaminants such as heavy metals may require the use of landfill sites specially designated for hazardous waste. NORM residues from the chemical extraction of rare earths from monazite are produced in significant quantities and have characteristically high activity concentra­tions. It has been demonstrated that such wastes can be suitably disposed of either in earthen trenches or in engineered cells, depending on the activ­ity concentration.

Forming geopolymers is a process that is very similar to cementation. Geopolymers are inorganic ceramic polymers made from aluminosilicates and cross-linked with alkali metal ions [126-128]. During fabrication, a low water content is used (H2O/M2O ~ 10-25 wt%) so that an amorphous geopolymer forms instead of crystalline zeolites which would technically form hydroceramic waste forms discussed below. A nominal composition of 4SiO2^Al2O3^M2O is used to represent the geopolymer matrix, although the Si:Al ratio varies according to the application from 1 to 3. For cements and concrete-like applications, a ratio of 2 : 1 is nominally used [129] . The alkali can be Na, K, or Cs. Geopolymers appear to be excellent low tem­perature binders and environmentally more acceptable than cement waste forms as the starting materials only need to be heated to ~700°C instead of clinkering at 1,400-1,500°C.

Geopolymers and geopolymeric cements, including but not limited to fly ash-based geopolymeric concretes, are ideal for environmental applications, such as the permanent encapsulation of radioactive species [130, 131] and other hazardous wastes [132]. Geopolymers can be used as sealants, capping, barriers, and other structures necessary at containment sites. Pilot-scale demonstrations have been performed in Europe on both mining wastes and uranium mill tailings [133-135]. Geopolymers were investigated for the disposal of radioactive wastes in Europe in the mid to late 1990s [136, 137] and the following applications have more recently been investigated.

• Geopolymers with Si:Al ratios of 1 : 1 and 2 : 1 for the stabilization of hazardous Resource Conservation and Recovery Act (RCRA) metals such as Ni, Se, Ba, Hg, Cd, Cr, Pb. A simulant RCRA spike was made that contained the RCRA components at 60x the concentration of the RCRA treatment standards known as the Universal Treatment Stand­ards or UTS limits [ 138] . The mixture was very acidic (pH < 1). The RCRA simulant was substituted for half of the 10 wt% water in the geopolymer formulation and the geopolymers met the Environment Protection Agency Toxicity Characteristic Leaching Procedure (EPA TCLP) test limits at less than the UTS limits even though the geopoly­mer contained 60x the UTS concentrations. It is not known whether or not the RCRA components interacted with the geopolymer, i. e. whether this was encapsulation or embedding (Table 6.9).

• Geopolymers derived from metakaolin and alkaline silicate solutions and having nominal Na/Al and Si/Al molar ratios of 1 and 2 were studied at ANSTO for the stabilization of [37Cs and [°Sr [139]. These geopolymers were studied by transmission electron microscopy and found to be amorphous on the ~1 nm scale after curing at 40°C. The Cs inhabited the amorphous phase, whereas Sr was incorporated only partly, being preferentially partitioned to crystalline SrCO3. This study implies that the geopolymer components do interact with some species and not with others, providing both encapsulation and embedding (Table 6.9 ).

• Special geopolymer formulations, marketed under the name DuraLith, have been patented [140] for stabilization of 129I and 99Tc. Testing [141] showed great promise for retention of technetium with rhenium used as a surrogate for the Tc, but not for iodine.

• Removal of radiolytic H2 production (and freeze-thaw problems) can be carried out by heating geopolymers at ~300°C without any serious effects on strength or leachability [142] .

• Geopolymers have demonstrated excellent fire resistance [142] .

Radioactive materials have been used widely since their discovery for a variety of scientific, medical and industrial uses. In some cases, either through ignorance, carelessness or accident, sites have been left contami­nated with residues of the operations. Such sites include, for instance, fac­tories where radium was used in luminescent paint (see Chapter 15) and thorium was used in thorium-coated gas mantles.

The optimum treatment of low-level liquid RAW involves a modular approach to the process, in which there is an autonomous technological module for each specific technology used in the LRAW treatment. At MosNPO ‘Radon’, long-term investigations were carried out into the effec­tiveness of different methods of low-level liquid RAW treatment, resulting in the development of a modular unit known as ‘Aqua-express’, involving h sorption and diaphragm methods of cleaning [20]. This unit, shown in Fig. 10.6, includes a filter-container with a nickel ferrocyanide absorbent (1), a cascade of sorption-filled filters (2), an ultra-filtration module (3) with membranes, the pore size of which does not exceed 50-100 nm. The unit’s capacity is 300-500 l/h of initial liquid RAW.

In 2002-2006 modular units of this sort were supplied to research centres in Bangladesh, Syria, Iran, Serbia and Uzbekistan. In 2006-2007 modular units were included in the projects of the Kazan department of the Federal State Unitary Enterprise (FSUE) ‘RosRAO’ and the United Institute of Energy and Nuclear Research in Belarus, and in 2007 they were delivered to Rostov SK Radon.

Currently, SF is stored in dry storage facilities located in the area of both NPPs in CASTOR-440/84-type approved casks or in pools at reactor sites. More than 9,000 SF assemblies from WWER 440 reactors and 1,000 spent assemblies from WWER 1000 reactors are stored in this way. More than 5,300 assemblies are expected to be spent by 2025 at Dukovany reactors and 2,600 assemblies by 2042 at Temelin reactors.

The multi-billion euro contract to build two new nuclear reactors at the current site of Temelin with the option for another three elsewhere has been launched recently in the Czech Republic; one in Dukovany and the second in Slovakia. The reactors will likely be built by US or Russian companies. It is expected that more than 8,000 fuel assemblies would be spent in the three new nuclear reactors in the Czech Republic during their 60 years of electricity production.

Other SF assemblies are from the research reactor located at the Nuclear Research Institute (NRI) Rez located near Prague. This experimental nuclear reactor has been in operation since 1957, with significant recon­struction performed in 1988-1989. Several hundred SF assemblies have been produced during this time. In 2004 the Czech Republic was included in the Global Threat Reduction Initiative (GTRI) programme. Highly enriched uranium (HEU) SF was shipped to the Russian Federation for reprocessing with the financial support of the US government and Department of Energy. To date, 457 spent fuel of IRT-2M and 208 EK 10 assemblies from the NRI research reactor have been sent to Russia. It is expected that after 20 years the corresponding activity will be returned back to the Czech Republic in the form of vitrified waste. NRI now also

participates in shipment of spent fuel assemblies from other Eastern Euro­pean countries’ research reactors to Russia, ‘Mayk’ Production Association. These reactors are of Russian provenance.

The main sources of radioactive waste are briefly described in the following sections. More detail is found in the following references (IAEA, 1977, 1988, 1992; Donald et al., 1997; Ojovan and Lee, 2005, 2007; Donald, 2010; Jantzen, 2011) as well as the annual symposium proceedings on the Scientific Basis for Nuclear Waste Management (Volumes I-XXXVI) published by the Materials Research Society (e. g., Lee et al., 2008).

1.1.1 Nuclear power plant operations and decommissioning

Waste generated during the operation of a NPP is generated mainly by treatment of water from the reactor or ancillaries including SF storage ponds and some decontamination operations. Standard effluent treatment technologies are based on evaporation (distillation), ion exchange, filtration or centrifuging. Typical process wastes from pressurised water reactors (PWR) are borated water concentrates, sludge or filter cartridges, and organic bead resin ion-exchangers (blow-down resins) from primary and secondary circuits, whereas those from boiling water reactors (BWR) are water concentrates and sludge containing different types of ion exchange or filter media as organic powdered resins, diatomaceous earth, activated carbon, cellulose and organic bead resins. Maintenance waste is mainly solid, comprising spent or damaged and contaminated equipment which cannot be repaired or recycled, and items such as contaminated clothes from operators, cardboard, bags, tools and plastic sheeting from mainte­nance work. Maintenance waste arises also from dismantling the internal structures of the reactor core including the used control rods. Liquid tech­nological wastes comprise mainly oils and small amounts of lubricants and organic solvents used for decontamination. Typically the main radioactive contaminants in operational waste are short-lived radionuclides such as 60Co, 90Sr, 134Cs and 137Cs, although long-lived radionuclides can be present in the internal elements of reactors.

Figure 1.5 shows schematically a nuclear fuel rod and an assembly. The fuel is in the form of ceramic Pu/U oxide pellets in the metal rod. When the fuel reaches the end of its useful life, it is removed from the reactor and is considered as SF. SF contains about 95% 238U, about 3% of fission products and transuranic isotopes, about 1% Pu and 1% 235U.

In the open NFC, the SF is considered as waste and can itself serve as a final waste form since it is a reasonably stable solid providing it is encap­sulated in an additional immobilising barrier such as a corrosion-resistant copper or lead container. The ceramic UO2 matrix of nuclear fuel retains the radionuclides and non-volatile fission products in its open fluorite crystal structure and its polycrystalline microstructure. The metal Zircaloy cladding of the fuel also, if intact, provides an additional barrier. About 30 tonnes of spent nuclear fuel (SNF) waste are typically produced per year by a typical 1 GW NPP.

DSS are a special category of institutional RAW coming from various industrial (non-destructive testing), medical, research and other applica­tions. Under normal conditions, they comprise firmly-contained single radionuclides with an activity that can vary over several orders of magnitude — from low dose rates brachytherapy and positron emission to­mography (PET) sources (typical activity of 10E-2-10E-4 TBq) to highly active teletherapy sources and radioisotope thermoelectric generators (typical activity of order of magnitude 10E-4 TBq). Various radionuclides, almost exclusively artificial, are used in sealed sources. The categorization of new sealed sources, described in detail in Ref. [6], is fully applicable also to DSS. A detailed registry of sealed sources is usually established according to national regulations and the most common procedure for management of spent sealed sources is to return them to the manufacturer, who is re­sponsible for further disposal.

The management of radioactive waste has been under discussion at the IAEA since its establishment in 1957. In 1961 the IAEA published Safety Series No. 5, dealing with the establishment of appropriate safety proce­dures and practices for the disposal of radioactive waste in the sea, and in 1965 the IAEA published guidance on radioactive disposal in the ground (Safety Series No. 15). By the late 1970s, it was clear that sea disposal was not an option favoured by many countries and since then land disposal has been preferred [53]. The first formal safety standard ‘Shallow Ground Disposal of Radioactive Wastes: A Guidebook’ [54] was published in 1981. As mentioned earlier, the IAEA suite of safety standards is made up of a safety fundamentals publication, safety requirements standards for differ­ent activities and facilities and supporting safety guides on meeting the

requirements. The Fundamental Safety Principles publication, SF-1 [37] defines ten safety principles (see Section 3.3) that must be met for all facilities and activities involving radioactive material and ionising radiation from uranium mining through reactor operation to radioactive waste dis­posal. The Safety Fundamentals are supported by the following general safety requirements of relevance to spent fuel and radioactive waste management:

• Basic Safety Standards No. 115 [38], currently under revision as an interim standard No. GSR Part 3 [55] based on the ICRP103 recom­mendations [7] ;

• Governmental, Legal and Regulatory Framework for Safety, No. GSR Part 1 [56] that replaces the previous publication No. GS-R-1 [57] of 2000;

• Safety Requirements on Predisposal Management of Radioactive Waste, No. GSR Part 5 [39] that replaces Safety Requirements No. WS-R-2 on Predisposal Management, Including Decommissioning [58];

• Safety Requirements on Disposal of Radioactive Waste No. SSR-5 [40] (that combines Safety Requirements on Geological Disposal of Radio­active Waste No. WS-R-4 [ 59] and the Safety Requirements on Near Surface Disposal of Radioactive Waste No. WS-R-1 [60] );

• Safety Requirements on Remediation of Areas Contaminated by Past Activities and Accidents Safety Requirements No. WS-R-3 [42] ;

• Safety Requirements on the Management System for Facilities and Activities, No. GS-R-3 [48];

• Safety Requirements on Safety Assessment for Facilities and Activities, No. GSR Part 4 [61];

• Safety Requirements on Decommissioning of Facilities Using Radio­active Material, No. WS-R-5 [41];

• Safety Requirements on Regulations for the Safe Transport of Radio­active Material No. TS-R-1 [62];

• Safety Requirements Preparedness and Response for a Nuclear or Radiological Emergency No. GS-R-2 [63] .

While the safety fundamentals and safety requirements set up provisions that must be complied with by the operators and licensees, the safety guides provide best practice for how to meet the principles and requirements. With respect to the pre-disposal and disposal of radioactive waste (i. e. GSR Part 5 and SSR-5) at present the following safety guides are of relevance: [9]

• Storage of Radioactive Waste, No. WS-G-6.1 [67];

• Management System for the Processing, Handling and Storage of Radi­oactive Waste, No. GS-G-3.3 [68];

• Management System for the Disposal of Radioactive Waste, No. GS-G.3.4. [69];

• Borehole Disposal Facilities for Radioactive Waste, No. SSG-1 [70];

• Management of Waste from the Use of Radioactive Material in Medi­cine, Industry, Agriculture, Research and Education, No. WS-G-2.7 [71];

• Application of the Concepts of Exclusion, Exemption and Clearance, No. RS-G-1.7 [44] .

In the area of decommissioning and management of waste generated during these activities, the following set of safety guides are in place:

• Decommissioning of Nuclear Power Plants and Research Reactors, No. WS-G-2.1 [45] that is currently under revision;

• Decommissioning of Medical, Industrial and Research Facilities, No. WS-G-2.2 [46] that is also currently under revision;

• Decommissioning of Nuclear Fuel Cycle Facilities, No. WS-G-2.4 [47] under revision;

• Safety Assessment for the Decommissioning of Facilities Using Radio­active Material, No. WS-G-5.2 [72] ;

• Release of Sites from Regulatory Control on Termination of Practices, No. WS-G-5.1 [43].

With respect to remediation and management or radioactive waste from mining and milling processing activities (past and current practices) the following IAEA safety guides apply:

• Remediation Process for Areas Affected by Past Practices and Acci­dents, No. WS-G-3.1 [73] ;

• Management of Radioactive Waste from the Mining and Milling of Ores, No. WS-G-1.2 [74] that is planned to be substituted by a new guide on Protection of the Public against Exposure to Natural Sources of Radiation including NORM (DS 421) [75];

• Occupational Radiation Protection in the Mining and Processing of Raw Materials, No. RS-G-1.6 [76].

A number of safety guides dealing mainly with spent fuel, safety assessment and safety case, as well as monitoring of disposal facilities are in a process of development and/or approval, such as:

• Storage of Spent Nuclear Fuel, No. SSG-15 [77];

• Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste, DS 284 [78];

• Near Surface Disposal, DS 356 [79];

• Geological Disposal, No. SSG-14 [80];

• Monitoring and Surveillance of Radioactive Waste Disposal Facilities, DS 357 [81];

• Safety Case and Safety Assessment for Radioactive Waste Disposal, No. SSG-23 [82];

• Control of Orphan Sources and Other Radioactive Material in the Metal Recycling and Production Industries, No. SSG-17 [83] .

Almost all of the radioactive gasses (H, C, Kr, Xe, Cl, and I) are released during voloxidation1 (if performed) and dissolution. They are carried in a relatively dry stream from the voloxidizer if present or in a wet stream containing significant nitrogen oxide species from the dissolver. These gaseous components can be selectively captured for immobilization. In the US, Cl and I must be captured for virtually all fuels and Kr must be captured for fuels cooled less than 30 years and 3H from fuels cooled less than -50 years. Figure 5.5 shows an example of the gaseous fission products evolved from various fuel recycling steps. Of primary interest are 3H, 1291,14C, and 85Kr.

Tritium removal from voloxidizer off-gas is performed using a desiccant (such as CaSOi) or a molecular sieve (such as Linde type 3A). The water content of air fed to the voloxidizer is controlled to obtain the desired decontamination factor in the tritiated water removal bed without signifi­cant increases in the tritium waste stream volume. Capture is performed near room temperature followed by release of the tritiated water at higher temperature. The captured (H, D,T)2O is then immobilized for decay storage and disposal. With a 12.3 year half-life, tritium immobilization does not require a robust waste form. Current process development activities assume that the tritium waste form is sufficiently low in long-lived radionuclides to qualify for near-surface disposal and the target waste form is generally considered to be a low-water cement. The other leading candidate is cemen­tation of the loaded sorbent.

Iodine-129 and 36Cl are significant dose contributors for nearly all reposi­tory environments because they are highly mobile, have long half-lives (15.7 x 106 years and 0.3 x 106 years, respectively), and are efficiently concentrated in the human body. Therefore, every reprocessing nation has strict toler­ances on the capture of 129I at a minimum. Various past studies have shown

Voloxidation is a potential process step employed primarily to remove 3H from the fuel meat prior to dissolution so that waste streams from all downstream solvent extraction processes are not 3 H contaminated. Tritium capture may not be necessary in all countries to the same level as required in the US (40CFR61 and 10CFR20), so processes with and without voloxida — tion will be considered.

that 94-99% of iodine reports to the dissolver off-gas. A large fraction of the iodine in the off-gas was found to be associated with organic compounds (e. g., methyl iodide). A range of technologies have been employed to capture iodine from the plant off-gas streams including (IAEA, 1980):

• silver saddles (AgNO3 on ceramic substrate) ^ Hanford and Savannah River

• silver faujasite (AgX) ^ Sellafield

• silver mordenite (AgZ) ^ Hanford

• AgNO3 on silica (e. g., AC-6120) ^ WAK and Mayak

• silver on alumina ^ LaHague, Rokkasho

• carbon ^ Hanford

• wet caustic scrub (2 m NaOH) ^ La Hague, Tokai, Krasnoyarsk, Mol, and Sellafield

• IODOX (20+ m HNO3)

• mercurex (mercuric and nitric acids) ^ Dounreay and West Valley

• cadmium faujasite (CdX)

Advancements in materials science have allowed for the development of improved solid getter materials for iodine. Chief among them are silver — loaded aerogels (Strachan et al., 2010a; Matyas 2012); metal organic frame­works (Nenoff et al., 2011; Sava et al., 2011) and chalcogenide-based glass aerogels (chalcogels) (Strachan et al., 2010a). However, these materials are currently in the development phase and are not ready for full implementation.

Iodine waste form development and waste management are closely coupled to the separations technique employed. For example, at La Hague

in France and Sellafield in the UK, iodine is managed by ocean disposal (isotope dilution) which leads to the most appropriate capture method of caustic scrubbing. Other than ocean disposal, the immobilization/manage — ment of iodine is still a significant technical challenge faced by the industry in general. Several waste forms have been proposed and are being devel­oped for the disposal of radioiodine.

Silver-loaded adsorbers (AgZ, AgX, AC-6120, alumina, etc.), for example, can be encapsulated in cements (Toyohara et al., 2002; Scheele et al., 2002) or low melting metals (Vance et al., 2005) or glasses (Garino et al., 2011; Perera et al., 2004), or hot pressed into a durable waste form (JAEA, 2007). Scheele et al. found that adding CaI2 to the grout would significantly reduce the leaching rate of 1 29 I by isotopic dilution in the pour water solution (Scheele et al., 2002). However, for some repository design concepts, the presence of cement is a disadvantage because of the impact of alkaline cement leach solution on the corrosion of HLW glass and SNF. For example, the Yucca Mountain repository design specifically excluded cement wher­ever possible. The loaded AgI containing ceramics or glass can be hot — pressed into a final waste form (Sheppard et al., 2006 ).

Alternatively, the iodine can be eluted from the capture media and immo­bilized. Pure halide waste can be immobilized in:

• bismuth oxide-based ceramics (Krumhansl and Nenoff, 2011)

• sodalite-like minerals (Strachan and Babad, 1979; Winters, 1980; Naka — zawa et al., 2000 )

• apatite-like minerals (Uno et al., 2001, 2004)

• glass by low temperature vitrification (Sakuragi et al., 2008 ; Mukunoki et al., 2009 ).

Table 5.2 summarizes several potential iodine waste forms along with their loading and anticipated performance. To date, the authors are not aware of any of these processes being utilized on an industrial scale.

For each topic in the performance evaluation of waste form and waste package (or canister), the associated risk in the disposal or extended storage system needs to be considered. Three risk-related questions are addressed: (i) what can go wrong?, (ii) how likely is it?, and (iii) what are the conse­quences? Various time-dependent or one-time behaviours of waste form and waste package are assessed with respect to these three questions. To answer the first question, features, events and processes (FEP) for the geo­logical disposal options (Nuclear Energy Agency, 1997), or equivalent (e. g., NRC, 2007; Dasgupta et al, 2002) are identified. Regarding the second ques­tion, the identified FEPs or their equivalents are evaluated with respect to given system designs considering normal conditions and accident conditions from man-made and natural hazards. For extended dry storage, monitoring, inspection and remediation will reduce the safety significance (likelihood or probability) of some FEPs. Once the likelihood or probability of a FEP, or its equivalent, exceeds a threshold value, its consequence may be assessed in terms of confinement failure, radionuclide release, nuclear subcriticality and radiation shielding, or other performance objectives of the total system or subsystems, thus addressing the third question. Implementation of the answers to these three questions are iterative in nature with modifications of design details, until risk assessment or design performance objectives are met. This iterative process also allows early identification of risk-significant issues related to different designs.

Based on the iterative process, the following FEPs associated with the behaviour of waste package or storage canister construction metals are considered significant in the two management cases.

• Long-term integrity of protective passive film for corrosion-resistant metals such as nickel-based alloys, titanium alloys or stainless steel. This allows low general corrosion rates for these metals, keeping the waste package or storage canister intact for a long time.

• Low oxygen or sulphur ion concentration in the reducing aqueous envi­ronment for corrosion-allowance metals such as copper or carbon steel. This also allows low general corrosion rates for these metals, keeping the waste package or storage canister intact for a long time.

• Low susceptibility to localized corrosion. Low ratios of chloride to nitrate ion concentrations prevent localized corrosion in nickel-based alloys and stainless steel. Low fluoride ion concentration prevents fast titanium dissolution without loss of adherent protective passive metal — oxide film. Carbon steel susceptible to pitting in the reducing environ­ment is minimal.

• Low susceptibility to SCC and/or hydrogen-induced cracking. The mag­nitude of residual stress or concentrations of chemical species such as carbonate ions in solution or salt deposits determine the susceptibility of a metal or alloys to SCC and/or hydrogen-induced cracking.

• Low initial manufacturing defects. This allows minimum early mechani­cal failure of waste package and canister due to manufacturing defects, for all metals considered.

Similarly, the following FEPs related to the waste form are potentially important for containing the SNF or HLW in the two management cases. [21]

Similarly, near-neutral pH of the aqueous environment also results in low dissolution rates of fission products and solubility limits for actinides in HLW glass dissolution (BSC, 2004).

• Conditions affecting performance of cladding. Conditions such as low hydrogen absorption, temperature, and residual stress in cladding mini­mize hydrogen-induced cladding failure.

• The high burnup of SNF encased by the cladding may increase or decrease the radionuclide release fraction, affecting radionuclide release in air.

In the following sections, selected specific subtopics from the above list are discussed in depth.