Imperial College London

DrMichaelOjovan

Faculty of EngineeringDepartment of Materials

Visiting Professor
 
 
 
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Royal School of MinesSouth Kensington Campus

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Summary

 

Publications

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237 results found

Tournier RF, Ojovan MI, 2021, Comments about a recent publication entitled “Improving glass forming ability of off-eutectic metallic glass formers by manipulating primary crystallization reactions”, Scripta Materialia, Vol: 205, ISSN: 1359-6462

Comments about a recent publication entitled "Improving glass forming ability of off-eutectic metallic glass formers by manipulating primary crystallization reactions." 'Acta Materialia, 200 (2020) 710-719 by Y. Q. Zeng, J. S. Yu, Y. Tian, A. Hirata, T. Fujita, X. H. Zhang, N. Nishiyama, H. Kato, J. Q. Jiang, A. Inoue, M. W. Chen. The liquidus temperatures Tl of Ni60Pd20P20-y-x SiyBx glass formers are equal to the disappearance temperatures Tn+ of liquid medium-range order resulting from the formation of an undercooled phase behind the glass phase. At this temperature, a volume fraction of about 15%, corresponding to the percolation threshold of configurons, melts accompanied by an endothermic enthalpy.

Journal article

Kochkin B, Malkovsky V, Yudintsev S, Petrov V, Ojovan Met al., 2021, Problems and perspectives of borehole disposal of radioactive waste, Progress in Nuclear Energy, Vol: 139, Pages: 1-9, ISSN: 0149-1970

An overview is given of status of projects for the disposal of radioactive waste in very deep boreholes in crystalline rocks which demonstrates all main pros and cons of this technology. New opportunities offered by drilling long horizontal drillholes in ductile formations can provide the basis for projects that have the potential to overcome many of the disadvantages of deep boreholes. The concept of disposal in horizontal drillholes brings together the technologies of borehole and mined repositories using the advantages of both, and therefore deserves an expert discussion at international level.

Journal article

Tournier RF, Ojovan MI, 2021, Dewetting temperatures of prefrozen and grafted layers in solid ultrathin films viewed as melt-memory effects, Physica B: Condensed Matter, Vol: 611, Pages: 1-10, ISSN: 0921-4526

Undercooled phase exists behind glass phase with superheated medium-range order between Tg and Tn+ > Tm. The ordered volume fraction stays equal to the percolation threshold F ≅ 0.15 of broken bonds up to Tn+. The difference ΔTg between Tg(bulk) of films with thickness (h > h0) and Tg(h) of ultrathin films of thickness (h < h0) is a linear function of (h-ho). Dense layer with minimum thickness hr is grafted against substrate by isothermal annealing, rinsed to reduce film thickness below hr/F, and finally dewetted at Tg. Similar thickness prepared and annealed near Tm and heated above Tm contains residual crystallized layer dewetting at Tn+. The prefrozen layer reproduces the glassy grafted layer in a crystallized state up to Tn+. Melting heat and melting temperature Tm are linear functions of h for h < h0. Prefrozen layers are due to melt-memories leading to new scenarios of crystallization.

Journal article

Tournier RF, Ojovan MI, 2021, Building and breaking bonds by homogenous nucleation in glass-forming melts leading to transitions in three liquid states, Materials, Vol: 14, ISSN: 1996-1944

The thermal history of melts leads to three liquid states above the melting temperatures Tm containing clusters—bound colloids with two opposite values of enthalpy +Δεlg × ΔHm and −Δεlg × ΔHm and zero. All colloid bonds disconnect at Tn+ > Tm and give rise in congruent materials, through a first-order transition at TLL = Tn+, forming a homogeneous liquid, containing tiny superatoms, built by short-range order. In non-congruent materials, (Tn+) and (TLL) are separated, Tn+ being the temperature of a second order and TLL the temperature of a first-order phase transition. (Tn+) and (TLL) are predicted from the knowledge of solidus and liquidus temperatures using non-classical homogenous nucleation. The first-order transition at TLL gives rise by cooling to a new liquid state containing colloids. Each colloid is a superatom, melted by homogeneous disintegration of nuclei instead of surface melting, and with a Gibbs free energy equal to that of a liquid droplet containing the same magic atom number. Internal and external bond number of colloids increases at Tn+ or from Tn+ to Tg. These liquid enthalpies reveal the natural presence of colloid–colloid bonding and antibonding in glass-forming melts. The Mpemba effect and its inverse exist in all melts and is due to the presence of these three liquid states.

Journal article

Ojovan MI, 2021, The Modified Random Network (MRN) model within the Configuron Percolation Theory (CPT) of glass transition, Ceramics, Vol: 4, Pages: 121-134, ISSN: 2571-6131

A brief overview is presented of the modified random network (MRN) model in glass science emphasizing the practical outcome of its use. Then, the configuron percolation theory (CPT) of glass–liquid transition is concisely outlined, emphasizing the role of the actual percolation thresholds observed in a complex system. The MRN model is shown as an important tool enabling to understand within CPT the reduced percolation threshold in complex oxide systems.

Journal article

Ojovan M, 2021, Glass formation, Encyclopedia of Glass Science, Technology, History, and Culture, Editors: Richet, Publisher: Wiley

Book chapter

Tournier RF, Ojovan MI, 2021, Undercooled phase behind the glass phase with superheated medium-range order above glass transition temperature, Physica B: Condensed Matter, Vol: 602, Pages: 1-17, ISSN: 0921-4526

Rapidly quenched glass formers are amorphous and transformed into glass phases by relaxing enthalpy during the first heating. Two liquids give rise, at first, to an intermediate Phase 3 below T3 < Tg respecting the entropy constraints and then, the enthalpy increases towards that of the glass phase up to Tg. The negative activation energy shows that Phase 3 is hidden behind the glassy phase acting as an intermediate invasive phase during the second cooling. Phase 3 carries a medium-range order above Tg which can be superheated above the melting temperature up to Tn+. The two-liquid state model predicts the thermodynamic properties as well as the relaxation times from liquids 1 to 2. The configuron model is successfully applied to 54 glasses explaining the transitions by percolation and an ‘ordered’ fraction equal to the critical threshold Φc = 0.15 ± 0.01 from Tg to Tn+.

Journal article

Ojovan MI, 2020, On viscous flow in glass-forming organic liquids, Molecules, Vol: 25, Pages: 1-13, ISSN: 1420-3049

The two-exponential Sheffield equation of viscosity η(T) = A1·T·[1 + A2·exp(Hm/RT)]·[1 + C·exp(Hd/RT)], where A1, A2, Hm, C, and Hm are material-specific constants, is used to analyze the viscous flows of two glass-forming organic materials—salol and α-phenyl-o-cresol. It is demonstrated that the viscosity equation can be simplified to a four-parameter version: η(T) = A·T·exp(Hm/RT)]·[1 + C·exp(Hd/RT)]. The Sheffield model gives a correct description of viscosity, with two exact Arrhenius-type asymptotes below and above the glass transition temperature, whereas near the Tg it gives practically the same results as well-known and widely used viscosity equations. It is revealed that the constants of the Sheffield equation are not universal for all temperature ranges and may need to be updated for very high temperatures, where changes occur in melt properties leading to modifications of A and Hm for both salol and α-phenyl-o-cresol.

Journal article

Malkovsky VI, Yudintsev SV, Ojovan M, Petrov VAet al., 2020, The influence of radiation on confinement properties of nuclear waste glasses, Science and Technology of Nuclear Installations, Vol: 2020, ISSN: 1687-6075

Self-irradiation can affect durability of glasses used to immobilize high-level nuclear waste (HLW). The stability of glasses can also be indirectly affected by the radiolytic changes in contact water leading to decrease in its pH although this is unlikely to occur for disposal systems where the interaction of groundwater with glass is delayed to times when radiation dose rates are decreased to levels insignificant to cause such effects. Besides, interaction of the water influenced by radiation with other repository protective elements (container and bentonite) will suppress the radiolytic changes. Literature analysis shows practical absence or very weak effect of self-irradiation on structure and characteristics of borosilicate glasses with typical content of nuclear waste. Data for aluminophosphate glass used in Russia have showed that, after γ-irradiation with a dose of 6.2·107 Gy, the leaching rates of elements were decreased approximately twice relatively to pristine samples.

Journal article

Zubekhina BY, Burakov BE, Ojovan MI, 2020, Surface alteration of borosilicate and phosphate nuclear waste glasses by hydration and irradiation, Challenges, Vol: 11, Pages: 1-11, ISSN: 2078-1547

We examined the degradation of nuclear waste borosilicate and phosphate glasses containing strong alpha-emitter 238Pu at a specific activity of 6.33 × 105 MBq/g in comparison with similar non-radioactive, non-radioactive irradiated and radioactive samples containing beta- and gamma-emitters, namely radionuclides 134Cs and 137Cs. For irradiation and leaching experiments, we used borosilicate and phosphate glasses, which are well-known and currently used to immobilize high-level radioactive waste. The main focus was the observation of the surface of altered glasses. Comparative analysis of hydrolytic surface alteration of borosilicate and phosphate nuclear waste glasses reveals that the behavior of radioactive samples differs significantly from that of non-radioactive glasses.

Journal article

Luzhetsky AV, Petrov VA, Yudintsev SV, Malkovsky VI, Ojovan MI, Nickolsky MS, Shiryaev AA, Danilov SS, Ostashkina EEet al., 2020, Effect of gamma irradiation on structural features and dissolution of nuclear waste Na–Al–P glasses in water, Sustainability, Vol: 12, Pages: 4137-4137, ISSN: 2071-1050

Structural properties and water dissolution of six sodium–aluminum–phosphate (NAP) glasses have been investigated before and after irradiation by a gamma-ray source based on 60Co. Two of these samples were of simple composition, and four samples had a complex composition with radionuclide simulants representing actinides, fission, and activated corrosion products. Samples of the simple composition are fully vitreous, whereas samples of the complex composition contained up to 10 vol.% of aluminum–phosphate, AlPO4, and traces of ruthenium dioxide, RuO2. Based on the study of pristine and irradiated glasses, it was established that the radiation dose of 62 million Gray had practically no effect on the phase composition and structure of samples. At the same time, the rate of leaching of elements from the irradiated samples by water was decreased by about two times.

Journal article

Ojovan MI, Louzguine-Luzgin DV, 2020, Revealing structural changes at glass transition via radial distribution functions., The Journal of Physical Chemistry B: Biophysical Chemistry, Biomaterials, Liquids, and Soft Matter, Vol: 124, Pages: 3186-3194, ISSN: 1520-5207

Transformation of glasses into liquids is discussed in terms of configuron (broken chemical bond or transformation of an atom from one to another atomic shell) percolation theory with structural changes caused. The first sharp diffraction minimum (FSDM) in the pair distribution function (PDF) is shown to contain information on structural changes in amorphous materials at the glass transition temperature (Tg). A method to determine the glass transition temperature is proposed based on allocating Tg to the temperature when a sharp kink in FSDM occurs. The method proposed is more sensitive compared with empirical criterion of Wendt-Abraham; e.g., for amorphous Ni the kink that determines Tg is almost twice sharper. Connection between the kink in fictive temperature behavior of PDF and Wendt-Abraham criterion is discussed.

Journal article

Sanditov DS, Ojovan MI, Darmaev MV, 2020, Glass transition criterion and plastic deformation of glass, Physica B: Condensed Matter, Vol: 582, ISSN: 0921-4526

We develop the notion that amorphous substances undergo reversible configurational structural changes accompanied by local expansion and compression (atom delocalization) near the glass transition temperature. They are similar in nature to configurational changes in the structure of glasses in the case of reversible frozen (plastic) deformation and its thermally stimulated relaxation. We assume that the glass-liquid transition is associated with the process of atom delocalization caused by bond breaking and formation of elementary excitations e.g. configurons. We discuss the possibility of detection of configuron formation and atom delocalization near glass transition based on temperature dependence of X-rays or neutron first sharp diffraction (pair distribution function) minimum.

Journal article

Kashcheev VA, Musatov ND, Ojovan MI, 2020, Advanced vitreous wasteforms for radioactive salt cake waste immobilisation, MRS Advances, Vol: 5, Pages: 121-129, ISSN: 2059-8521

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).

Journal article

Ojovan MI, 2020, On alteration rate renewal stage of nuclear waste glass corrosion, MRS Advances, Vol: 5, Pages: 111-120, ISSN: 2059-8521

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.

Journal article

Farid OM, Ojovan M, Rahman ROA, 2020, Evolution of cations speciation during the initial leaching stage of alkali-borosilicate-glasses, MRS Advances, Vol: 5, Pages: 185-193, ISSN: 2059-8521

Alkali-borosilicate glasses (ABS) are used as host immobilization matrices for different radioactive waste streams and are characterized by their ability to incorporate a wide variety of metal oxides with respectively high waste loadings. The vitreous wasteform is also characterized by very good physical and chemical durability. The durability of three ABS compositions were analyzed by investigating their leaching behavior using the MCC1 test protocol and these data were used to investigate the waste components retention in the altered layer and the evolution of the interfacial water composition during the test. The results indicated that the Mg species evolution is exceptional with respect to other alkaline elements and dependent on glass matrix composition and leaching progress, while transition elements speciation is fairly constant throughout leaching process and independent on glass compositions. Si and B species are changing during leaching process and are affected by waste composition. For modified wasteform sample, evolution of Mg, Si and B species is respectively constant, whereas at highest waste loading, these elements have fairly constant speciation evolution within the first 2 weeks of leaching.

Journal article

Jantzen CM, Ojovan MI, 2019, On selection of matrix (wasteform) material for higher activity nuclear waste immobilization (review), Russian Journal of Inorganic Chemistry, Vol: 64, Pages: 1611-1624, ISSN: 0036-0236

Selection of wasteform materials for higher activity nuclear waste containment is considered. Utilization of materials such as glasses, ceramics, glass composite materials and cements is discussed as practiced in different countries. Emphasis is on multiple parameter approach on selecting the wasteform where the durability is not solely the most important characteristic.

Journal article

Hyatt NC, Ojovan M, 2019, Special Issue: Materials for Nuclear Waste Immobilization, Materials, Vol: 12, ISSN: 1996-1944

Nuclear energy is clean, reliable, and competitive with many useful applications, among which power generation is the most important as it can gradually replace fossil fuels and avoid massive pollution of environment. A by-product resulting from utilization of nuclear energy in both power generation and other applications, such as in medicine, industry, agriculture, and research, is nuclear waste. Safe and effective management of nuclear waste is crucial to ensure sustainable utilization of nuclear energy. Nuclear waste must be processed to make it safe for storage, transportation, and final disposal, which includes its conditioning, so it is immobilized and packaged before storage and disposal. Immobilization of waste radionuclides in durable wasteform materials provides the most important barrier to contribute to the overall performance of any storage and/or disposal system. Materials for nuclear waste immobilization are thus at the core of multibarrier systems of isolation of radioactive waste from environment aimed to ensure long term safety of storage and disposal. This Special Issue analyzes the materials currently used as well as novel materials for nuclear waste immobilization, including technological approaches utilized in nuclear waste conditioning pursuing to ensure efficiency and long-term safety of storage and disposal systems. It focuses on advanced cementitious materials, geopolymers, glasses, glass composite materials, and ceramics developed and used in nuclear waste immobilization, with the performance of such materials of utmost importance. The book outlines recent advances in nuclear wasteform materials including glasses, ceramics, cements, and spent nuclear fuel. It focuses on durability aspects and contains data on performance of nuclear wasteforms as well as expected behavior in a disposal environment.

Journal article

Orlova AI, Ojovan MI, 2019, Ceramic mineral waste-forms for nuclear waste immobilization, Materials, Vol: 12, Pages: 1-45, ISSN: 1996-1944

Crystalline ceramics are intensively investigated as effective materials in various nuclear energy applications, such as inert matrix and accident tolerant fuels and nuclear waste immobilization. This paper presents an analysis of the current status of work in this field of material sciences. We have considered inorganic materials characterized by different structures, including simple oxides with fluorite structure, complex oxides (pyrochlore, murataite, zirconolite, perovskite, hollandite, garnet, crichtonite, freudenbergite, and P-pollucite), simple silicates (zircon/thorite/coffinite, titanite (sphen), britholite), framework silicates (zeolite, pollucite, nepheline /leucite, sodalite, cancrinite, micas structures), phosphates (monazite, xenotime, apatite, kosnarite (NZP), langbeinite, thorium phosphate diphosphate, struvite, meta-ankoleite), and aluminates with a magnetoplumbite structure. These materials can contain in their composition various cations in different combinations and ratios: Li–Cs, Tl, Ag, Be–Ba, Pb, Mn, Co, Ni, Cu, Cd, B, Al, Fe, Ga, Sc, Cr, V, Sb, Nb, Ta, La, Ce, rare-earth elements (REEs), Si, Ti, Zr, Hf, Sn, Bi, Nb, Th, U, Np, Pu, Am and Cm. They can be prepared in the form of powders, including nano-powders, as well as in form of monolith (bulk) ceramics. To produce ceramics, cold pressing and sintering (frittage), hot pressing, hot isostatic pressing and spark plasma sintering (SPS) can be used. The SPS method is now considered as one of most promising in applications with actual radioactive substances, enabling a densification of up to 98–99.9% to be achieved in a few minutes. Characteristics of the structures obtained (e.g., syngony, unit cell parameters, drawings) are described based upon an analysis of 462 publications.

Journal article

Farid O, Ojovan M, Massoud A, Abdel Rahman ROet al., 2019, An Assessment of Initial Leaching Characteristics of Alkali-Borosilicate Glasses for Nuclear Waste Immobilization, Materials, Vol: 12, Pages: 1462-1462

<jats:p>Initial leaching characteristics of simulated nuclear waste immobilized in three alkali- borosilicate glasses (ABS-waste) were studied. The effects of matrix composition on the containment performance and degradation resistance measures were evaluated. Normalized release rates are in conformance with data reported in the literature. High Li and Mg loadings lead to the highest initial de-polymerization of sample ABS-waste (17) and contributed to its thermodynamic instability. Ca stabilizes non-bridging oxygen (NBO) and reduces the thermodynamic instability of the modified matrix. An exponential temporal change in the alteration thickness was noted for samples ABS-waste (17) and Modified Alkali-Borosilicate (MABS)-waste (20), whereas a linear temporal change was noted for sample ABS-waste (25). Leaching processes that contribute to the fractional release of all studied elements within the initial stage of glass corrosion were quantified and the main controlling leach process for each element was identified. As the waste loading increases, the contribution of the dissolution process to the overall fractional release of structural elements decreases by 43.44, 5.05, 38.07, and 52.99% for Si, B, Na, and Li respectively, and the presence of modifiers reduces this contribution for all the studied metalloids. The dissolution process plays an important role in controlling the release of Li and Cs, and this role is reduced by increasing the waste loading.</jats:p>

Journal article

Sanditov DS, Ojovan M, 2019, Relaxation aspects of the liquid-glass transition, PHYSICS-USPEKHI, Vol: 62, Pages: 111-130, ISSN: 1063-7869

Journal article

Ojovan MI, Burakov BE, Lee WE, 2018, Radiation-induced microcrystal shape change as a mechanism of wasteform degradation, Journal of Nuclear Materials, Vol: 501, Pages: 162-171, ISSN: 0022-3115

Experiments with actinide-containing insulating wasteforms such as devitrified glasses containing 244 Cm, Ti-pyrochlore, single-phase La-monazite, Pu-monazite ceramics, Eu-monazite and zircon single crystals containing 238 Pu indicate that mechanical self-irradiation-induced destruction may not reveal itself for many years (even decades). The mechanisms causing these slowly-occurring changes remain unknown therefore in addition to known mechanisms of wasteform degradation such as matrix swelling and loss of solid solution we have modelled the damaging effects of electrical fields induced by the decay of radionuclides in clusters embedded in a non-conducting matrix. Three effects were important: (i) electric breakdown; (ii) cluster shape change due to dipole interaction, and (iii) cluster shape change due to polarisation interaction. We reveal a critical size of radioactive clusters in non-conducting matrices so that the matrix material can be damaged if clusters are larger than this critical size. The most important parameters that control the matrix integrity are the radioactive cluster (inhomogeneity) size, specific radioactivity, and effective matrix electrical conductivity. We conclude that the wasteform should be as homogeneous as possible and even electrically conductive to avoid potential damage caused by electrical charges induced by radioactive decay.

Journal article

Ojovan MI, Robbins RA, Garamszeghy M, 2018, Advances in conditioning of low-and intermediate-level nuclear waste, Scientific Bases for Nuclear Waste Management, Pages: 983-990

© 2017 Materials Research Society. 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: â- Immobilization which may or may not also provide volume reduction, including a) Low temperature processes and b) Thermal processes; â- Containerization for a) Transport, b) Storage, and c) Disposal; â- 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.

Conference paper

Sanditov DS, Ojovan M, 2017, On relaxation nature of glass transition in amorphous materials, Physica B: Condensed Matter, Vol: 523, Pages: 96-113, ISSN: 0921-4526

A short review on relaxation theories of glass transition is presented. The main attention is paid to modern aspects of the glass transition equation qτg = C, suggested by Bartenev in 1951 (q – cooling rate of the melt, τg – structural relaxation time at the glass transition temperature Tg). This equation represents a criterion of structural relaxation at transition from liquid to glass at T = Tg (analogous to the condition of mechanical relaxation ωτ = 1, where the maximum of mechanical loss is observed). The empirical parameter ะก = δTg has the meaning of temperature range δTg that characterizes the liquid-glass transition. Different approaches of δTg calculation are reviewed. In the framework of the model of delocalized atoms a modified kinetic criterion of glass transition is proposed (q/Tg)τg = Cg, where Cg ≅ 7·10−3 is a practically universal dimensionless constant. It depends on fraction of fluctuation volume fg, which is frozen at the glass transition temperature Cg = fg/ln(1/fg). The value of fg is approximately constant fg ≅ 0.025. At Tg the process of atom delocalization, i.e. its displacement from the equilibrium position, is frozen. In silicate glasses atom delocalization is reduced to critical displacement of bridge oxygen atom in Si-O-Si bridge necessary to switch a valence bond according to Muller and Nemilov.An equation is derived for the temperature dependence of viscosity of glass-forming liquids in the wide temperature range, including the liquid-glass transition and the region of higher temperatures. Notion of (bridge) atom delocalization is developed, which is related to necessity of local low activation deformation of structural network for realization of elementary act of viscous flow – activated switch of a valence (bridge) bond. Without atom delocalization (“trigger mechanism”) a switch of the valence bond is impossible and, consequently, the viscous

Journal article

Ojovan M, Robbins RA, 2017, Vitreous Materials for Nuclear Waste Immobilisation and IAEA Support Activities., MRS Advances, Vol: 1, Pages: 4201-4206, ISSN: 2059-8521

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.

Journal article

Ojovan M, Wickham AJ, 2017, Processing of Irradiated Graphite: The Outcomes of an IAEA Coordinated Research Project, MRS Advances, Vol: 1, Pages: 4117-4122, ISSN: 2059-8521

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).

Journal article

Ojovan M, Wickham AJ, Steinmetz HJ, O'Sullivan Pet al., 2017, Updating irradiated graphite disposal: Project ‘GRAPA’ and the international decommissioning network., Journal of Environmental Radioactivity, Vol: 171, Pages: 34-40, ISSN: 0265-931X

Demonstrating competence in planning and executing the disposal of radioactive wastes is a key factor in the public perception of the nuclear power industry and must be demonstrated when making the case for new nuclear build. This work addresses the particular waste stream of irradiated graphite, mostly derived from reactor moderators and amounting to more than 250,000 tonnes world-wide. Use may be made of its unique chemical and physical properties to consider possible processing and disposal options outside the normal simple classifications and repository options for mixed low or intermediate-level wastes. The IAEA has an obvious involvement in radioactive waste disposal and has established a new project ‘GRAPA’ – Irradiated Graphite Processing Approaches – to encourage an international debate and collaborative work aimed at optimising and facilitating the treatment of irradiated graphite.

Journal article

Ojovan M, Schmidt OV, Kascheev VA, Poluektov PPet al., 2016, Modelling aqueous corrosion of nuclear waste phosphate glass. P.P. Poluektov, O.V. Schmidt, V.A. Kascheev, M.I. Ojovan. J. Nucl. Mater., 484, 357–366 (2017), Journal of Nuclear Materials, Vol: 484, Pages: 357-366, ISSN: 0022-3115

A model is presented on nuclear sodium alumina phosphate (NAP) glass aqueous corrosion accounting for dissolution of radioactive glass and formation of corrosion products surface layer on the glass contacting ground water of a disposal environment. Modelling is used to process available experimental data demonstrating the generic inhibiting role of corrosion products on the NAP glass surface.

Journal article

Ojovan M, Lee WE, 2016, About U-shaped Glass Corrosion Rate/pH Curves for Vitreous Nuclear Wasteforms, Innovations in Corrosion and Materials Science, Vol: 7, Pages: 30-37, ISSN: 2352-0949

Background: We analyse here the well-known U-shaped form of pH-dependence of glass corrosion rates U(t) focusing on changes of U-shaped form with time.Methods: We account for the two main mechanisms of glass corrosion which play dominant role at different times - ion exchange and hydrolysis.Results: We reveal that the pH dependence evolves even for solutions ostensibly kept at constant pH. The U(t) dependence is caused by changes of concentration profiles of elements in the nearsurface layers of glasses in contact with water. The change is most evident within the initial stages of glass corrosion at relatively low temperatures. Numerical examples are given for the Russian nuclear waste borosilicate glass K-26 characterised in experiment by an effective diffusion coefficient for 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.Conclusion: The changes of the U-shaped form of glass corrosion rates need to be accounted for when assessing the performance of glasses in contact with water solutions.

Journal article

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