The three generically accepted stages of glass corrosion are reviewed with focus on final stage termed alteration rate renewal (or resumption) stage when the glass may re-start corroding with the rate similar to that at the initial stage. It is emphasized that physical state and physical changes that occur in the near-surface layers can readily lead to an effective increase of leaching rate which is similar to alteration rate renewals. Experimental data on long-term (during few decades) corrosion of radioactive borosilicate glass K26 designed to immobilize high-sodium operational NPP radioactive waste evidence on resumption-like effects of radionuclides (137,134Cs) leaching. The cause of that was however related not to chemical changes in the leaching environment but rather to physical state of glass surface due to formation of small cracks and new pristine glass areas in contact with water.

Salt cake radioactive waste is a remnant solid salt concentrate after deep evaporation of radioactive evaporator concentrate at WWER NPP’s. The traditional cementing of borate-containing liquid radioactive waste, to which the salt cake belongs, leads to a significant increase in the volume of the final product. This work describes borosilicate vitreous wasteforms developed to immobilize radioactive salt cake waste and comprises data on both glass synthesis and characterization. The composition of glass selected for the purpose of immobilisation of the salt cake radioactive waste allows to include up to 40 wt. % of the oxides contained in the salt cake and to reduce the volume of the final product by more than 2 times compared with the cement compound. The batches were melted in a cold crucible melter at 1200 °C. The normalized cesium leaching rate of the vitrified wasteform product was within range 3.0·10-5 – 3.7·10-6 g/(cm2·day).

Radioactive waste with widely varying characteristics is generated from the operation and maintenance of nuclear reactors, nuclear fuel cycle facilities, research facilities and medical facilities and the through the use of radioisotopes in industrial applications. The waste needs to be treated and conditioned appropriately to provide wasteforms acceptable for safe storage and disposal. Conditioning of radioactive waste is an important step to prepare waste for long-term storage or disposal and includes the following processes:

• ▪ Immobilization which may or may not also provide volume reduction, including

  • a) Low temperature processes and

  • b) Thermal processes;

• ▪ Containerization for

  • a) Transport,

  • b) Storage, and

  • c) Disposal;

• ▪ Overpacking of primary containers

  • a) Prior to disposal and

  • b) In a disposal facility as part of disposal process.

Conditioning consists of operations that produce a waste package suitable for handling, transportation, storage and/or disposal and may be performed for a variety of reasons including standardization of practices and/or wasteforms, technical requirements for waste stability in relation to a repository design or safety case, technical requirements related to waste transportation, societal preferences, regulatory preferences, etc. This paper gives an overview of recent advances in conditioning of low- and intermediate-level radioactive waste.

The paper is based on the new IAEA Handbook “Conditioning of Low- and Intermediate-Level Liquid, Solidified and Solid Waste” which is one of eight IAEA handbooks intended to provide guidance for evaluating and implementing various characterisation and radioactive waste processing and storage technologies before final disposal

Mechanical damage of non-metallic nuclear wasteforms can be caused by electrical fields induced by decaying clusters of radionuclides surrounded by an insulating matrix. We assess the electric fields near clusters with decaying radionuclides 244Cm, 241Am, 238,239Pu and 137Cs in a glass matrix determining that matrix destruction can gradually occur via electric breakdown discharges and diffusion-controlled change in form of clusters. The most important parameters that control potential matrix destruction are the radioactive cluster (inhomogeneity) size, radionuclide specific radioactivity and effective electrical conductivity of the matrix.

Vitreous materials are the overwhelming world-wide choice for the immobilisation of HLW resulting from nuclear fuel reprocessing due to glass tolerance for the chemical elements found in the waste as well as its inherent stability and durability. Vitrification is a mature technology and has been used for high-level nuclear waste immobilization for more than 50 years. Borosilicate glass is the formulation of choice in most applications although other formulations are also used e.g. phosphate glasses are used to immobilize high level wastes in Russia. The excellent durability of vitrified radioactive waste ensures a high degree of environment protection. Waste vitrification gives high waste volume reduction along with simple and cheap disposal facilities. Although vitrification requires a high initial investment and then operational costs, the overall cost of vitrified radioactive waste is usually lower than alternative options when account is taken of transportation and disposal expenses. Glass has proven to be also a suitable matrix for intermediate and low-level radioactive wastes and is currently used to treat legacy waste in USA, and NPP operational waste in Russia and South Korea. This report is also outlining IAEA activities aiming to support utilisation of vitreous materials for nuclear waste immobilisation.

Dismantling of old reactors and the management of radioactive graphite wastes are becoming increasingly important issues for a number of IAEA Member States. Exchange of information and research cooperation in resolving identical problems between different institutions contributes towards improving waste-management practices, their efficiency, and general safety. The IAEA Coordinated Research Project (CRP) under the title ’Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal’ was conducted during 2010-2014 and has involved 24 organisations from ten Member States [1]. The CRP has explored both innovative and conventional methods for graphite characterisation, retrieval, treatment, and conditioning technologies and produced an IAEA technical document [2] which has identified a number of unresolved scientific and technical issues such as the need to:

• 1. Improve the scientific understanding required on creation, chemical form, location and release behaviour (transport models) of radionuclides;

• 2. Improve predictive models of radioisotope behaviour;

• 3. Ensure that sampling programmes are statistically representative of the totality of the graphite to be disposed of;

• 4. Establish an accurate radionuclide inventory;

• 5. Consider novel alternative dismantling and treatment strategies.

The CRP promoted the exchange of technical information on R & D activities and will facilitate practical application for treatment and conditioning of graphite waste. The collaboration continues under the IAEA International Decommissioning and Predisposal Networks (IDN and IPN).

Data on a viscous flow model based on network defects – broken bonds termed configurons – were analysed. An universal equation has been derived for the variable activation energy of viscous flow Q(T) of the generic Frenkel equation of viscosity η(T)=A∙exp(Q/RT) which is known to have two constant asymptotes – high QH at low temperatures and low QL at high temperatures. The defect model of flow used by e.g. Doremus, Mott, Nemilov, Sanditov states that higher the concentration of defects (e.g. configurons) the lower the viscosity. We have used the configuron percolation theory (CPT) which treats glass–liquid transition as a percolation-type phase transition. Additionally the CPT results in a continuous temperature relationship for viscosity valid for both glassy and liquid amorphous materials. We show that a particular result of CPT is the universal temperature relationship for the activation energy of viscous flow: Q(T)=QL+RT∙ln[1+exp(-Sd/R) exp((QH-QL)/RT)] which depends on asymptotic energies QL (for the liquid phase) and QH (for the glassy phase), and on entropy of configurons Sd. This equation has two asymptotes, namely Q(T<<Tg) = QH, and Q(T>>Tg) = QL. Moreover we demonstrate that the equation for Q(T) practically coincides in the transition range of temperatures with known Sanditov equation.

Alkaline earth aluminosilicate glasses (AeAS) with different MoO3 additions have been produced and assessed. MoO3 solubility increases with the equimolar substitution of smaller to larger alkaline earths and reaches 5.34 mol% in magnesium aluminosilicate glass (MAS). All visibly homogeneous glasses are X-ray amorphous, while the partially crystallised glasses exhibit some small X-ray diffraction peaks which are probably due to corresponding molybdates. The addition of MoO3 decreases glass transition and crystallisation temperatures and creates two broad Raman bands which are assigned to vibrations of MoO42‒ tetrahedra. The intensities of these bands increase along with MoO3 incorporation until the maximum solubility is reached. Electron microscopy shows that these separated particles are spherical, with sub-micron diameters and are randomly dispersed within glass. The separated phases are formed through liquid-liquid separation and thereafter crystallisation. Overall AeAS glasses look quite promising for molybdate immobilisation with MAS glasses being particularly attractive.

The pH-dependence of glass corrosion rates has a well-known U-shaped form with minima for near-neutral solutions. This paper analyses the change of U-shaped form with time and reveals that the pH dependence evolves even for solutions that have pH not affected by glass corrosion mathematically corresponding to a zero surface to volume ratio. The U(t) dependence is due to changes of concentration profiles of elements in the near-surface layers of glasses in contact with water and is most evident within the initial stages of glass corrosion at relatively low temperatures. Numerical examples are given for the nuclear waste borosilicate glass K-26 which is experimentally characterised by an effective diffusion coefficient of caesium DCs = 4.5 10-12 cm2/day and by a rate of glass hydrolysis in non-saturated groundwater as high as rh = 100 nm/year The changes of U-shaped form need to be accounted when assessing the performance of glasses in contact with water solutions.

An overview is given of an International Atomic Energy Agency Coordinated Research Project (CRP) on the treatment of irradiated graphite (i-graphite) to meet acceptance criteria for waste disposal. Graphite is a unique radioactive waste stream, with some quarter-million metric tons worldwide eventually needing to be disposed of. The CRP has involved 24 organizations from 10 Member States. Innovative and conventional methods for i-graphite characterization, retrieval, treatment and conditioning technologies have been explored in the course of this work, and offer a range of options for competent authorities in individual Member States to deploy according to local requirements and regulatory conditions.

Ordering types in the disordered structure of amorphous materials and structural changes which occur at glass-liquid transition are discussed revealing medium range order and reduction of topological signature of bonding system.

An overview of selection criteria for waste processing and disposal technologies is given. A systematic approach for selection of an optimal technology is proposed. Optimal selection of a technical processing and disposal option is case specific to the waste management needs. Waste streams considered are from nuclear applications, research, power generation, nuclear fuel cycle activities and decommissioning of nuclear facilities as well as for NORM-containing waste.

An overview is given on research activities on cementitious materials for radioactive waste management systems based on the IAEA Coordinated Research Project (CRP) held in 2007-2010. It has been joined by 26 research organizations from 22 countries which shared their research and practical activities on use of cementitious materials for various barrier purposes. The CRP has initially formulated the research topics considered within four specific streams: A) Conventional cementitious systems; B) Novel cementitious materials and technologies; C) Testing and waste acceptance criteria; and D) Modelling long term behaviour.

The CRP has analysed both barrier functions and interactions envisaged between various components with focus on predisposal stage of waste management. Cementation processes have achieved a high degree of acceptance and many processes are now regarded as technically mature. A large body of information is currently available on proven waste conditioning technologies although novel approaches are continuing to be devised.

Most of the existing technologies have been developed for conditioning of large amounts of operational radioactive waste from nuclear power plants and other nuclear fuel cycle facilities. However new waste streams including those resulting from legacy and decommissioning activities required improved material performance and technologies.

The most important outcome of CRP was the exchange of information and research co-operation between different institutions and has contributed towards general enhancement of safety by improving waste management practices and their efficiency. The paper presents the most important results and trends revealed by CRP participants. The research contributions of participating organizations will be published as country contributions in a forthcoming IAEA technical publication.

Cementitious materials are widely used in waste management systems with different aims and requirements for long term performance. Both conventional and novel cementitious materials are used to create reliable immobilising elements for safe storage and disposal of wastes. The barrier elements as well as interactions envisaged between various components are important to ultimately ensure the overall safety of a storage/disposal system. The behaviour and performance of cementitious materials including waste package components, wasteform and backfilling were analysed within the IAEA Coordinated Research Project which involved 26 research organizations from 21 Member States MS). The paper presents briefly the main research outcomes for conventional cementitious systems; novel materials and technologies; testing and waste acceptance criteria; and modelling long term behaviour.

Thermochemical processing of reactor graphite waste is based on self-sustaining reaction 4Al + 3TiO2 + 3C = 3TiC + 2Al2O3 which chemically binds 14C from the irradiated graphite in the titanium carbide. Thermochemical processing was investigated to analyse the behaviour of rare earth elements (REE), where REE = Y, La, Ce, Nd, Sm, Eu and Gd. Both thermodynamic simulations and laboratory scale experiments were used. The REEs in the irradiated reactor graphite are formed as activation products of impurities and spread over the graphite bricks surfaces as well as arise from fission of nuclear fuel. REEs can be used also to substitute for waste actinides as well as to increase the durability of carbide-corundum ceramics relative to waste actinides.

Thermodynamic calculations and X-ray diffraction analysis of ceramic specimens synthesized revealed that durable REE's aluminates with perovskite, β-alumina and garnet structures are formed by interaction of REE oxides with the Al2O3 melt during the selfpropagating reaction of ceramic formation.

The porous carbide-corundum ceramics synthesized have a high hydrolytic durability, e.g. the normalised leaching rates of 137Cs, 90Sr and Nd are of the order of 10–7 – 10–8 g/(cm2·day).

Glass composite materials (GCM) are versatile wasteforms for immobilising various types of both radioactive and hazardous wastes. We review current research on the utilisation of GCMs for hazardous and radioactive waste immobilisation. Compared to homogeneous glassy materials GCMs can incorporate larger amounts of waste elements and, in the case of glass matrix composites, they can be produced using lower processing temperatures (by viscous flow sintering) than those of conventional melting.

The acoustic emission (AE) technique was used to detect and characterise the processes associated with generation of stress waves during melting and solidification of granite E93/7 at a pressure of 0.15 GPa. The AE signals recorded as a result of partial melting of the granite at a temperature of 780°C and subsequent solidification during cooling were distinguished from the equipment noise and their parameters used to characterise the AE sources associated with the phase transformations during melting and solidification of the granite. The mechanisms generating AE during granite melting were differentiated by AE signals with their highest peaks in the frequency spectrum at 170 and 268 kHz. The transformation of the liquid into glass during solidification of the partially melted granite generated AE waves in an essentially broad range of frequencies between 100 and 300 kHz. This preliminary work demonstrates the potential of the AE technique for use in applications related to deep borehole disposal of radioactive wastes.

Alkali-borosilicate glasses are widely used in nuclear industry as a matrix for immobilisation of hazardous radioactive wastes. Durability or corrosion resistance of these glasses is one of key parameters in waste storage and disposal safety. It is influenced by many factors such as composition of glass and surrounding media, temperature, time and so on. As these glasses contain radioactive elements most of their properties including corrosion resistance are also impacted by self-irradiation.

The effect of external gamma-irradiation on the short-term (up to 27 days) dissolution of waste borosilicate glasses at moderate temperatures (30¢X to 60¢XC) was studied. The glasses studied were Magnox Waste glass used for immobilisation of HLW in UK, and K-26 glass used in Russia for ILW immobilisation. Glass samples were irradiated under ×-source (Co-60) up to doses 1 and 11 MGy. Normalised rates of elemental release and activation energy of release were measured for Na, Li, Ca, Mg, B, Si and Mo before and after irradiation. Irradiation up to 1 MGy results in increase of leaching rate of almost all elements from both MW and K-26 with the exception of Na release from MW glass. Further irradiation up to a dose of 11 MGy leads to the decrease of elemental release rates to nearly initial value. Another effect of irradiation is increase of activation energies of elemental release.

Acoustic emission (AE) signals were continuously recorded and their intrinsic frequency characteristics examined in order to evaluate the mechanical performance of cementitious wasteform with encapsulated Al waste. The primary frequency in the power spectrum and its range of intensity for the detected acoustic waves were potentially related with appearance of different micromechanical events caused by Al corrosion within the encapsulating cement system. In addition the process of cement matrix hardening has been shown as a source of AE signals characterised with essentially higher primary frequency (above 2 MHz) compared with those due to Al corrosion development (below 40 kHz) and cement cracking (above 100 kHz).

Long-term field tests of cemented aqueous radioactive wastes in an experimental mound type surface repository were carried out at Moscow Scientific and Industrial Association “Radon” from 1965 to 2004. Aqueous radioactive wastes of different compositions containing short-lived radionuclides including 90Sr and 137Cs at concentrations from 0.34 to 1.8 MBq/L were immobilized using cementation technology. Water solution to cement ratio was 0.66, grout mixing time 10-15 minutes, and cement paste hardening time 7 days. 73 cement blocks with a volume of 0.027 m3 were disposed of for long-term tests in a simple mound type surface repository. The atmospheric precipitates, which contacted radioactive cement blocks, were collected and analyzed for the content of radionuclides. In August 2004 the experimental repository was opened, cemented blocks, underlying and covering materials were retrieved for analyses. XRD analyses showed that along with amorphous tobermorite gel the main crystal phases in cements are calcite and portlandite. Both visual inspection and radiometric analyses demonstrate that cemented blocks are in good condition and that the cement paste has retained radionuclides from the wastes. Thus after 39 years of storage in the mound type repository the cemented aqueous wastes are reliable immobilized.