Waste Glasses

Waste Glasses$ E Vernaz, CEA – Marcoule, Bagnols sur Cèze Cedex, France C Veyer, AREVA NP–E&P, Montigny le Bretonneux, France S Gin, CEA – Marcoule, Bagnols sur Cèze Cedex, France r 2016 Elsevier Inc. All rights reserved.

1 2 2.1 2.2 2.3 2.4 2.5 2.6 3 3.1 3.1.1 3.1.2 3.1.3 3.2 3.3 3.3.1 3.3.2 3.3.3 3.3.4 4 4.1 4.2 4.2.1 4.2.2 4.2.3 4.3 4.3.1 4.3.2 4.3.3 4.3.4 4.3.5 4.3.6 4.3.7 4.3.8 4.3.9 4.4 5 5.1 5.1.1 5.1.2 5.2 5.2.1 5.2.2 6 References

Introduction Glass and Vitreous State Phenomenological Approach of Glass Glass Transition Temperature Glass Structure at Atomic Scale Polymerization and De-Polymerization of the Glass Network Structure of R7T7-Type Containment Glass Crystallization Mechanisms Waste Glass Definition and Characterization The Waste Streams to Vitrify Nature and composition of HLW solutions Other kinds of nuclear waste Hazardous waste Glass Formulation Glass Characteristics of Interest Microstructural homogeneity Physical properties Thermal stability and crystallization potential Chemical durability Long Term Behavior of Nuclear Waste Glasses Glass Crystallization and Long Term Thermal Stability Glass Resistance to Self-Irradiation Investigations of glasses doped with a short half-life actinide Atomistic modeling of glass self-irradiation External irradiation of glasses Nuclear Glass Alteration by Water Basic mechanisms of glass alteration Initial rate of glass dissolution Alteration rate in saturated conditions and final rate of glass dissolution Essential role of the ‘Passivating Reactive Interphase’ (PRI) Influence of glass composition Influence of groundwater and environmental materials Influence of glass fracturing Modeling glass long term behavior Natural an archeological analogues Conclusions on Glass Long Term Behavior Vitrification Processes Existing Processes for Radioactive Waste Vitrification The French two-step continuous vitrification process Liquid-fed ceramic melters Emerging Processes for Radioactive Waste Vitrification Cold crucible induction melters Incineration/vitrification processes Conclusions and Outlook on Waste Glasses

2 3 3 4 4 5 6 6 6 7 7 7 8 8 11 11 12 13 13 15 15 16 16 17 19 19 19 19 19 20 20 20 21 21 21 22 23 24 24 25 27 27 29 30 31

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Change History: April 2015. E. Vernaz nothing is changed until Section 4.2. About ‘Glass Long Term Behavior’ (Sections 4.2.1, 4.3.1, 4.3.2, 4.3.3, and 4.3.4), research is going on; and some new references were added. Sometimes a sentence was added to introduce the new result. About ‘Glass processing’: Minor changes in 5.1.1 – AVM is now decommissioned, the number of glass canister produced in La Hague was updated. In Section 5.2.1, the changes are a bit more important as the cold crucible melter is now in operation at La Hague. Figure 23 was added. There is no change in the figures until Figure 22 ( so Figures 4–6, 9, 11, 12, 19, 21, and 22 are to be mentioned in the text exactly as before in the above pdf). Only Figure 23 has been updated as Figure 24.

Reference Module in Materials Science and Materials Engineering

doi:10.1016/B978-0-12-803581-8.00758-X

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Waste Glasses

Abbreviations FPs fission products MA minor actinides HLW high-level waste BO bridging oxygen NBO non-bridging oxygen NMR nuclear magnetic resonance EXAFS extended X-ray absorption fine structure LWR light water reactor PUREX plutonium and uranium refining by extraction MOX (fuel) mixed OXide (fuel) CEA Commissariat à l'Energie Atomique et aux énergies alternative (the French Atomic and Alternative Energy Commission). GCR gas-cooled reactor

1

AVM atelier de vitrification de Marcoule (the first industrial vitrification plant in France) SEM scanning electron microscope TEM transmission electron microscope ICP-MS inductively coupled plasma–mass spectroscopy ICP-AES inductively coupled plasma-atomic emission spectroscopy SIMS secondary ion mass spectrometry STEM scanning transmission electron microscope GRAAL (model) acronym for ‘glass reactivity with allowance for the alteration layer’ UP1 the first French reprocessing plant located in Marcoule

Introduction

Fission products (FPs) and minor actinides (MAs) produced during fuel irradiation in the reactor only represent about 5% of the weight of used nuclear fuel, but about 98% of its radioactivity. When the fuel is reprocessed, these FPs and MAs end up in concentrated solutions (called high level waste – HLW – solutions) that are stored in tanks fitted with stirring systems and cooling capacities to evacuate the heat resulting from radioactive decay. Figure 1 presents an example of such a storage tank, made of stainless steel, in a commercial reprocessing plant. Such a storage principle can be safe for several decades. However it requires active monitoring and maintenance, and cannot reasonably be extended for the durations required for complete decay of the activity (thousands of years). As soon as the mid-50s, the major western countries have started designing plans for their nuclear waste, and work on FP immobilization has been initiated at Oak Ridge (USA), Harwell (UK), Chalk-River (Canada), and Saclay (France). Several materials were considered at first, with a rapid convergence on glass or glass-ceramics compositions. For instance, the first attempts at the CEA in 1957 targeted crystals of mica-phlogopite (M2Mg6(AlSi3)2O20F4, M being an alkali or an alkaline earth), but this was soon abandoned due to the impossibility of digesting all the elements of the concentrated solution within one specific mineral. During these first tests, a glassy component was frequently observed at the bottom of the crucible, with often better durability than that of the targeted mineral. Since, at the same time, some favorable results had been obtained at Chalk River, where a confining glass had been obtained by melting natural aluminosilicates impregnated with FP solutions at 1350 1C, glass was selected for further investigations in France.1 A new application for glass was born: glass for the containment of radioactivity. At least, the idea was born, but, from the idea to industrial deployment a long path remained to be covered, to optimize glass compositions adapted for each type of FP solution, and to develop processes operable in very radioactive environments. Similar exploratory work was performed in the other countries, which ended up twenty years later in the quasi unanimous selection of borosilicate glass as the preferable matrix.2

Figure 1 Example of a HLW storage tank for concentrated fission product solutions.

Waste Glasses

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3

Glass and Vitreous State

Glass is one of the oldest materials known to man. During pre-historic times, man used natural glasses (volcanic) to make knives or arrowheads. The first glass actually melted by man could date back to 4500 BC. Although, in common language, the term ‘glass’ often refers to a fragile and transparent material, the scientific approach regarding vitreous state is both much wider (for instance, nuclear glasses are not transparent) and more difficult to define. This chapter aims at providing the basis to understand vitreous state, by considering the aspects of the formation of a glassy structure, the major glass properties (viscosity, durability, thermal stability) and the fine atomic scale structure of glass.3

2.1

Phenomenological Approach of Glass

Why are most of natural rocks crystallized, while only a small number of them display amorphous structures (absence of diffraction peaks as evidenced by XRD)? Most minerals compounds, when in the molten state, form liquids with a low viscosity (some centipoisesi). Upon cooling, these liquids easily crystallize when they reach their melting temperature. Some of these liquids, however, are very viscous in the range of their melting temperature (typically 105–107 Poise). Such liquids, if they are kept below their liquidus temperature (in this case, they are supercooled liquids) will tend to crystallize very slowly. If the cooling rate is faster than the crystallization rate, crystallization will not occur. During cooling, the viscosity of the supercooled liquid increases progressively until the material rigidifies: the liquid ‘vitrifies’ or transitions from supercooled liquids to the ‘vitreous state.’ A phenomenological definition of ‘glass’ could then be ‘glass is a rigidified supercooled liquid.’ This definition is nevertheless too restrictive, because a glass can be obtained by other routes, (sol–gel for instance). Several alternative approaches can be proposed:

• • • •

structural approach: absence of order in the distribution of elementary structural units at scales larger than 10–30 Å , thermodynamic approach: glass is in a metastable state. It is nevertheless not unstable because the energy gap that must be overcome to bring it to its more stable crystallized state is generally significant due to the high viscosity, physical approach: glass is a non porous, impermeable, isotropic, non cleavable, elastic, solid with a fragile rupture behavior (absence of plastic deformation before failure), kinetic approach: glass is a material which transitions continuously and reversibly from liquid to solid state with temperature (Figure 2).

This figure provides an example of a typical evolution of viscosity between 250 and 1500 1C. One can distinguish an elastic solid domain below 500 1C, a plastic domain between 500 and 1000 1C, where the glass can be worked (blown, made to fibers, molded, etc.) and a liquid domain above 1000 1C.

Figure 2 Typical evolution of glass viscosity with temperature.

i

The Poise (P) is a viscosity unit commonly used in the glass industry 1 Poise ¼ 0.1 Pa.s1, 1 cP ¼103 Pa.s1.

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Waste Glasses

Figure 3 Evolution of the specific volume V of a glass or a crystal during cooling.

2.2

Glass Transition Temperature

Figure 3 compares the evolutions of the specific volumes (inverse of densities) for a glass and a crystal with temperature. During cooling of a liquid, a sharp step is observed in the evolution of density when the material crystallizes at melting temperature. If the material does not crystallize, the specific volume continues to decrease smoothly below this melting temperature until a change of slope is observed at a temperature Tg. This temperature is called the glass transition temperature. At that temperature, the material transitions from a supercooled liquid to a solid whose expansion coefficient (the slope of the curve) is roughly a third of that of the liquid. Similar glass transition patterns are observed for other thermodynamic parameters such as specific heat. The glass transition temperature corresponds to a glass viscosity of 1013 poises. One can then propose another definition: ‘glass is an amorphous solid that displays the glass transition behavior.’

2.3

Glass Structure at Atomic Scale

Numerous compounds can be made into vitreous structures: oxide (silicates, borates, phosphates, etc.), chalcogenide (sulfides, selenides, tellurides, etc.), ionic compounds (BeF2, ZrF4-BaF2, AlF3, ZnCl2, etc.), specific metallic alloys subjected to overhardening (Pd4Si, FeB, ZrCo, CaMg, etc.), and also organic compounds such as glycerin, polyethylene, or glucose. For instance, the shiny caramel of a tiered cake is obtained by cooling molten sugar fast enough to obtain a vitreous state; candy floss is an organic glass fiber... whose Tg is around 55 1C !.4 The most common glass compositions are silica-based oxide glasses. Even if pure silica can vitrify by itself (application to optic fibers), all-days glasses are enriched with other oxide-based components. They include, for instance, window glass (alumina– sodium–calcium–silicate), Pyrexs (a high silica borosilicate), crystal-glass (silicate glass with a high content of lead oxide), optical glass (such as ‘flint’ glass with high barium oxide), nuclear glass (alumina borosilicate), etc. The constitutive oxides of an oxide glass can be categorized into three families:

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Glass network forming oxides (network formers): these oxides are able to form glass by themselves; SiO2, B2O3, GeO2, and P2O5, are the most common. In silica glass, the basic structural units are silica tetrahedra [SiO4] sharing corners. Their linking creates a continuous network with some degree of disorder (groupings in cycles of 5 or 6 tetrahedra for instance), while, in a crystalline form such as quartz, the tetrahedra are perfectly ordered. Network modifiers: these oxides cannot give a glass by themselves. When they are coupled with network formers they are inserted into the vitreous structure and modify the properties of the material. Typically, these oxides are alkali or alkaline earth oxides (Na2O, Li2O, Cs2O, CaO, BaO, etc.). As an example, the introduction of sodium oxide in a silicate network induces the breakdown of strongly covalent Si–O bonds and their replacement by Si–O…Na þ bonds with a more ionic character.

O

O j  Si j O



O



O j Si j O

 O

þ

Na2 O -

O 

O j Si j O

Naþ  O

O Naþ



O j Si j O

 O

Waste Glasses

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The oxygen atoms bonding two silicon atoms are said to be ‘bridging oxygen’ (BO) atoms. The introduction of two sodium atoms in the structure creates two ‘non-bridging oxygen’ (NBO) atoms. The same mechanisms may be observed with CaO, but here only one Ca þ þ is needed to compensate the negative charge of the two NBO atoms. The presence of NBO atoms weakens the vitreous structure and allows decreasing melting temperature (the glass becomes less refractory) and viscosity (the glass can be poured more easily). This explains why alkali oxides are often used as fluxes. NBO atoms also loosen the network, and then help incorporating more elements in the structure. A counterpart is a decrease in chemical durability. In the end it should be recalled that the amount of modifiers must remain limited if one wants to obtain a glass from a molten liquid. If the number on NBO atoms is too high, the liquid becomes very fluid and tends to crystallize easily upon cooling. Intermediate oxides: these oxides cannot give a glass by themselves. However, when mixed with network modifiers, they behave like network formers within the vitreous structure. These oxides are typically Al2O3, Fe2O3, ZnO, ZrO2, PbO, TiO2, etc. Alumina is specifically important for the glass industry, since it improves chemical durability (for container glass, parti
 cularly). In order to behave like a network former and to form a tetrahedron similar to that of silica, the AlO4=2 ion needs a positive charge to maintain local electro-neutrality. This will be achieved by fixing an alkali ion (or one alkaline earth ion for
 two AlO4=2 ). This results in the following structure:

O 

•

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O j Si j O

Naþ 

O

O Na

þ

O j  Si j O

0

O þ

Al2 O3

O B j B B O Si - 2 B B B j @ O



O 

O j Na Al  j O

1 C C C OC C C A

Introducing alumina into an alkali silicate glass thus ‘hardens' the glass, since NBO atoms are eliminated according to the stoichiometry (1Na þ for 1Al or 1Ca þ þ for two Al). This will allow keeping a structure slightly looser than that of a pure silica glass (since the Al–O bonds are slightly weaker than the Si–O bonds), while leaving only a small number of NBO atoms. The introduction of calcium and some alumina into the formulation allowed to transition from the poorly durable alkali glass used for medieval stained glass to the window glass used at present. The specific role of boron: The behavior of boron oxide in the presence of alkalis differs from the general pattern described for silica. Adding alkalis to pure B2O3 or to a mixture of SiO2 and B2O3 leads at first to the formation of BO atoms, as a result of the formation of BO4 tetrahedra from the BO3 triangles initially present in the B2O3 glass (in this case the alkali cation is used to compensate the charge of the [BO4/2] structural unitii) . It is only for higher alkali contents that NBO atoms are formed, in the environment of boron or silicon atoms. Consequently, the addition of alkalis to a borate or borosilicate glass initially induces an increase in viscosity. It is only for higher additions that viscosity starts to decrease. This deviation in the behavior of boron when compared to silica is often termed ‘boron anomaly.’

Boron, in right amount, play a key role In nuclear borosilicate for several reasons: (1) it helps decreasing the glass melting temperature without drop of the durability as [BO4], units decrease the number of NBO by fixing an alkali ion; (2) it helps digesting a lot chemical element that would be lowly soluble in pure silica; (3) it prevents glass crystallization; and (4) it participates to the good glass long term behavior by allowing the formation of a very fine gel structure (without boron) and by decreasing the final pH.

2.4

Polymerization and De-Polymerization of the Glass Network

Based on the above classification, it is observed that the progressive addition of network modifiers to silica SiO2 leads to network de-polymerization by the formation of NBO atoms. At a certain limit, the liquid is strongly de-polymerized, its viscosity becomes very low, and it is not possible to obtain a glass by quenching any more. On the contrary, adding boron or intermediate oxides favors network polymerization by fixing alkalis or alkaline earth as charge compensators. The very good performances of nuclear borosilicate (moderate melting temperature coupled with good durability) come from an adequate balance between boron and intermediate elements (Al, Fe, Zr, etc.) on one side and the alkali elements on the other side, allowing a high polymerization rate as most of alkali are found as charge compensator rather than forming NBOs. ii The elementary structural units of glass are often written [SiO4], [BO4], [AlO4] to indicate the tetrahedral environment of Si, B, or Al; however, for more detailed structural descriptions, the more rigorous notation [SiO4/2], [BO4/2], [AlO4/2] is preferred. In addition to indicating that the base atom is surrounded by four oxygen atoms, this notation stresses the fact that the oxygen atoms are shared between two tetrahedra and, for B and Al, that a negative charge exists, which will need compensation by a neighboring cation.

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Waste Glasses

Figure 4 Nucleation and growth of a crystalline phase within in a glass.

2.5

Structure of R7T7-Type Containment Glass

The atomic structure of nuclear waste glasses has been studied using spectroscopic techniques such as NMR or EXAFS, at first on simplified glass compositions (SiO2–B2O3–Na2O–Al2O3), and then on compositions with increasing complexity. These studies show that all the intermediate cations in the R7T7 glass composition are in positions of network formers. Their substitution to Si4 þ induces charge deficits which are compensated by a network modifier cation in order to ensure charge 2   neutrality. The sum of negative charges created by the (AlO 4=2 ), (ZrO6=2 ), (FeO4=2 ), and (BO4=2 ) groups in R7T7 amounts to 5.19 mol per 100 mol of elements (computed from elemental mole%). On the other hand, the sum of positive charges created by alkalis and alkaline earths is 9.44 mol%. As a result:

• •

All the intermediate cations behave as network formers. A somewhat limited number of network modifiers remain available to create NBOs.

These considerations show that the network of this glass composition is homogeneous and very well co-polymerized, owing to the good incorporation of intermediate cations in the network. The network is most probably composed of numerous mixed Si–O–M bonds (with M¼ Al, B, Zr, Zn, etc. and even some rare earths). This strong polymerization, as well as all the structural data, are confirmed by molecular dynamics modeling.

2.6

Crystallization Mechanisms

Crystallization in a supercooled liquid or a glass occurs via a nucleation-growth mechanism which starts by the formation of small ordered germs. The germination can be enhanced by the presence of rough interfaces in the melt. Figure 4 gives an example of nucleation and growth curves for a glass. Several crystalline phases can be formed during cooling of a supercooled liquid, according to its composition. The chemical composition of the formed crystals can be very different from those of the liquid or of the glass. For a given phase, graph (n) plots the number of supercritical germs per unit volume and time as a function of temperature. graph (g) plots the growth rate for this phase as a function of temperature. If those two curves do not overlap, the given crystal cannot form spontaneously during cooling. In this case, when the glass is in the growth zone, there are no germs liable to grow, and when it is in the nucleation zone, the temperature is too low to allow the formed nuclei to grow. Crystallization will then be possible only after a two stages heat treatment, one stage for nucleation, the second stage for growth. This is a favorable situation for the production of glass-ceramics with homogeneous and well controlled crystallization.5,6 On the other hand, if the two curves overlap, a risk of uncontrolled crystallization appears in the overlap region, where germs can be formed and grow. The relative position of the nucleation and growth curves in the temperature field will then be an important parameter determining the sensitivity of the glass to crystallization; it must be studied for each of the phases liable to be formed in a given glass composition.

3

Waste Glass Definition and Characterization

Vitrification is not a simple encapsulation process (as bitumen for instance) but consists in making a new material where the waste components are contained at the atomic scale within the matrix and can only be released by destruction of the network bonds.

Waste Glasses

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One major requirement is then that the selected matrix be able to incorporate all of the waste stream components in its structure. By using the flexibility brought about by the disordered and relatively loose structure of a glass, it is possible to design glass compositions able to integrate a very wide range of elements within their structure, and which are tolerant to compositional variations in the waste stream. This approach constitutes waste vitrification, where waste components are usually mixed with suitable additives and molten to give a glass wasteform. A recent development of vitrification has been the design of glass-ceramics which combine the flexibility of glass formulation to digest most of the waste components with the possibility of targeting (a) well defined crystalline phase(s) for specific waste components that may not be soluble in large amounts in glass, as molybdenum for instance. Vitrification usually involves a small number of processing steps, with a robust design, compatible with operation in a highly radioactive or hazardous environment.

3.1

The Waste Streams to Vitrify

A large variety of nuclear waste compositions have been considered for vitrification since the first attempts in the late 1950s, including not only high-level, but also intermediate or low-level effluents. More recently, this approach has been extended to other types of inorganic hazardous waste for which other types of immobilization were not considered suitable.

3.1.1

Nature and composition of HLW solutions

FPs and MAs produced in the reactor by fission or neutron capture display a wide range of atomic numbers. One can find for instance alkalis (rubidium, cesium), alkaline earths (strontium, barium), a wide range of transition elements (zirconium, molybdenum, etc.), noble metals (ruthenium, rhodium, and palladium), chalcogenides (Se and Te), non-metals (As, Sb, etc.), lanthanides, and actinides. The amount and composition spectrum of the FPs and MAs varies with fuel initial composition, enrichment, and burnup. The separation process used for LWR oxide fuel at the commercial reprocessing plants of La Hague (France), Rokkasho Mura (Japan), or Sellafield (UK) is a hydrometallurgical process based on PUREX, where, after nitric dissolution of the fuel and a series of solvent extraction steps used for U and Pu recycling, most FPs and MAs from the fuel end up in a concentrated nitric solution (high level waste solution – HLW) which constitutes the major target feed for vitrification. In addition to the isotopes extracted from the fuel the HLW stream holds some chemicals added during reprocessing. The current PUREX-based process for LWR oxide fuel has been designed to minimize these additions, by using essentially chemicals which will not add to the waste load. However it is not possible to avoid the addition of some selected chemicals, mainly sodium, during ancillary operations such as solvent purification or equipment cleaning operations. Some impurities resulting from a slight corrosion of the piping (namely Fe, Cr, and Ni) or solvent degradation (phosphate) also end up in the HLW solution. In addition to commercial LWR fuel, other fuels have been reprocessed worldwide in the past, and other processes have been used or are considered for separating uranium and/or plutonium from the fuel. Fuel alloy metals (Al and Mo) or dissolved cladding material (Al, Mg, and stainless steel) may follow the HLW stream. The chemicals used to perform the dissolutions or separations can also be very diverse (mercury, fluorine, ferrous sulfamate, bismuth phosphate, etc.). In several instances the HLW solution (which is, initially, acidic) has been neutralized by adding massive amounts of caustic to prevent corrosion of the tanks, thus precipitating most of the FPs and MAs as hydroxides. This is for instance the case in the US, at Hanford or Savannah River; The HLW solutions are thus quite complex and not unique, with a large number of constituents. Typically, in France, concentrated HLW solutions are nitric solutions (1–2 N) with high bg activities (several tens of TBq per liter) including suspended solids such as colloids (zirconium phosphate, cesium phosphomolybdate) and some metallic fines (insoluble residue, cladding fines generated during the fuel shearing operation). Table 1 gives examples of solutions derived from reprocessing various types of fuel from past or present reactors in France.

3.1.2

Other kinds of nuclear waste

In addition to HLW, vitrification is more and more considered for waste of lower radioactivity content, although other, less costly, immobilization methods, such as grouting, are more common. Despite the cost, vitrification has the advantage of providing durable matrices, with significant volume reduction. For liquid effluents, a vitrification plant is for instance being designed and built at the Hanford site, USA, as part of the Waste Treatment Plant (WTP) to immobilize the low activity fraction of the waste from 177 underground tanks. This low activity fraction consists mainly of concentrated sodium nitrate and sodium hydroxide solutions, with some aluminum, chrome, sulfur, and other minor metals. In Russia, at the Radon facility near Moscow, borated low activity power plant effluents are also vitrified. Vitrification is also applied to low activity solid combustible waste, in conjunction with incineration: the combustible fraction is incinerated and the inorganic fraction, under the form of ashes, is vitrified. This is for instance the case at Ulchin, S. Korea, where a facility based on French technology has been recently commissioned to simultaneously incinerate and vitrify dry active waste and resins from nuclear power plants. At the Zwilag waste management facility, in Switzerland, a plasma-torch powered facility has been built to incinerate, melt and vitrify low activity waste conditioned in drums.

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Table 1

Example of HLW compositions to vitrify in France

Reactor type

PWR

Gas-cooled, graphite moderated, natural uranium reactor

Fuel type Burnup (MWj t  1) HLW solution (l/tU) Oxide contents (g/L) Free acidity (N) Oxide composition (g/L) Fe Al Cr Ni Na Mg Zn Mo Sn P F PF oxides (g/L) Actinide oxides

UO2 33 000 660 75–100 0.95

SiCrAl 4000–6000 100–120 90–100 1.0

240 0.8

13 0–2 2,3 1.9 20 – 0–1 – – 1.26 – 52.23 3.83

7–14 19–38 0–1.5 0–1.3 7–11 3–7 – – – 1–2 2–8 45.1 3.65

3.14 2.76 0.65 0.31 14 1.6 – 163.2 0.67 41.16 – 11.46 3.62

UMo–MoSnAl 1500–3500

In the end, vitrification has also sometimes been considered for actinide-bearing waste or residues. In the USA, for instance, a significant program has been launched in the 1990s in the frame of the START-II program, to study the possibility of vitrifying plutonium-rich residues from the weapon industry that could not be easily made into MOX fuel.

3.1.3

Hazardous waste

Vitrification, alone or combined with incineration, provides a convenient method for immobilizing or rendering hazardous waste innocuous. For instance, in the USA, vitrification has been considered as Best Demonstrated Available Technology (BDAT) by the US-EPA (US Environmental Protection Agency) for waste containing toxic metals such as arsenic. In situ vitrification of soils has been applied in some contaminated sites in the US. Owing to the potential for immobilizing inorganic, non-volatile, toxic metals into a stable and durable matrix, vitrification is used in several facilities worldwide, including France and Japan, to immobilize the slag and ashes from municipal waste incineration or waste-to-energy conversion facilities. The product is completely inert and can be disposed of in a conventional disposal facility. Efforts are under way in several institutions to improve the product for reuse, mostly in the building industry (as road base or tiles for instance). Another significant application of vitrification is the destruction of asbestos: once vitrified, asbestos becomes a compact, harmless substance. Such a facility for asbestos vitrification has been operating in South-Western France since 2003.

3.2

Glass Formulation

Glass formulation is aimed at defining a compositional domain within which the matrix will display a number of required characteristics related to technological feasibility, durability, containment properties, and all the properties required for the intended use or destination of the product. For high-level waste glass, for instance, the product will have to ensure the long-term safety in the proposed geological disposal environment. Glass formulation then consists in reaching the best compromise between a large number of constraints: glass formation domain (solubility of the various waste components, waste loading, homogeneous glass), technological feasibility (melting temperature limits, viscosity allowing to pour the product, minimum volatilisation, minimum corrosion of the melter), stability and containment properties (thermal stability, resistance to self-irradiation, chemical durability, mechanical properties, etc.) (Figure 5). For radioactive waste immobilization, borosilicate systems represent the best compromise between the various constraints, and they have been selected as the reference matrix compositions for High-Level Waste in most countries (France, USA, UK, Japan, Germany, Belgium, etc.). Other matrices for High-level waste include phosphate glasses in Russia. Glass formulation involves several successive (and often iterative) steps illustrated below by the methodology used at the French CEA for high-level waste borosilicate glass formulation: in this instance, it was necessary to design a matrix with very good containment and long term properties, to immobilize a very radioactive, heat-generating, waste stream, but whose composition was expected to be very stable throughout the years. In other countries, or for other waste streams, the practical organization of formulation studies may differ, but these steps are always necessary to ensure consistent and reliable wasteform properties.

Waste Glasses

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Ability to accommodate the waste Solubility (Cr, Ru, Rh, Pd, Ce, Pu, SO4, Cl) Phase separation (Mo, SO4, Cl, P) Devitrification (Mo, P, F, Mg, …) Maximize the waste loading

Process / Technology

Glass performance

Ease of processing Melting temperature Viscosity, reactivity, residence time, Electrical cond., thermal cond. Additives needed

Properties for storage/disposal Thermal stability Chemical durability Resistance to self irradiation Mechanical properties

Figure 5 Waste glass formulation is a compromise.

Figure 6 Basic quaternary glasses for GCR on left (SAN type glasses) and LWR on right (SON type glasses).

•

Establishment of glass formation diagrams

The glass-forming domain is determined by establishing quaternary phase diagrams with the expected four major components of the glass, to identify the glass-forming regions. For waste solutions dominated by FP oxides, the four major components can be summarized as silica SiO2, boron oxide B2O3, sodium oxide Na2O, and FP oxides (considered collectively, with a chemical spectrum corresponding to the expected spectrum for the solution). This leads to ‘SON’ type glass formulations such as those selected for the LWR FP solutions at La Hague. When the waste solution is dominated by aluminum, the quaternary diagram considers SiO2, Al2O3, B2O3, and Na2O. This leads to ‘SAN’-type glass formulations, such as those selected for the GCR FP solutions processed at the AVM facility at Marcoule (Figure 6). The glass-forming domain is established by melting a large number of glass compositions in small laboratory crucibles and performing visual or microscopic observations. As can be seen in Figure 7, boron helps to dissolve the entire FP spectrum into the glass, prevent crystallization and lower viscosity. The chosen boron content is however limited to the minimum needed to have a sufficient domain of homogeneous glass while keeping the best durability (for instance 18 wt% B2O3 is sufficient in the quaternary systems shown in Figure 7). Once this glass-forming region has been determined, additional constraints are used to further limit this domain, for instance: limitation of the melting temperature, which will result in limits on the refractory elements such as silica or alumina, limitation of waste loading by thermal considerations (for high-level waste), limitation of network modifiers to keep an acceptable chemical durability. Each of these constraints establishes one side of the diamond in which the glass composition will be chosen.

•

Optimization of glass formulation Optimization consists in adding or substituting various additives (such as Li2O, CaO, ZnO, ZrO2, etc.) to improve the matrix properties. Among all properties of interest, have to be mentioned the waste load (considering the final glass homogeneity), the viscosity of the molten glass, and the chemical durability. Depending on the choice for industrial process, other properties may be targeted in addition. As an example, small amounts of lithium may be substituted for sodium to decrease melt viscosity, without affecting the chemical durability. Aluminum may be substituted for a small amount of silica and calcium to improve durability.

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Waste Glasses

Figure 7 Effect of boron oxide content on the vitrification domain in quaternary glasses. Domains of glass formation into the quaternary diagram SiO2–Na2O–OxPF–B2O3 for two B2O3 contents.

Table 2

Chemical composition range of the French R7T7 glass

Chemical composition range of R7T7 glasses produced in the AREVA–La Hague plant workshops Oxides

SiO2 B2O3 Al2O3 Na2O CaO Fe2O3 Nio Cr2O3 P2O5 Li2O ZnO Oxides (PF þ Zr þ actinides) Fines suspension Actinide oxides SiO2 þ B2O3 þ Al2O3

•

Specified interval for the industry (wt%) Min

Max

42.4 12.4 3.6 8.1 3.5

51.7 16.5 6.6 11.0 4.8 o4.5 o0.5 o0.5 o0.6 2.4 2.8 18.5

1.6 2.2 7.5 460

Average composition of industrial glasses (wt%)

45.6 14.1 4.7 9.9 4.0 1.1 0.1 0.1 0.2 2.0 2.5 17.0 0.6 64.4

It has to be pointed out that those optimizations result from sharp compromises and high-level expertises in mixing the appropriate oxides in a synergic way (for instance high aluminum content enhance the durability but also the risk of crystallization of undesirable phases; controlled substitution of lithium to sodium improve viscosity without impacting durability, but induce crystallization of soluble phases at higher contents, etc.). At the end of this phase, a reference composition is proposed for a given effluent composition and a given melting technology. Dozens of compositions are published in the literature, for a very wide range of waste compositions, in many countries. Table 2 gives the composition of the French reference glass for the HLW solutions derived from processing 33 GWj T1 LWR fuel at La Hague: the so-called ‘R7T7’ composition, named from the vitrification facilities R7 and T7 at La Hague plant. Validation of the reference formulation

Once a reference composition has been defined at lab scale for a specific simulatediii waste stream, it needs to be validated for actual industrial application. To this effect, several additional programs are performed: J complete characterization of the reference composition and determination of all its important properties J validation of technological feasibility in industrial conditions by performing long duration demonstrations on an inactive industrial-scale pilot facility, and confirmation of the maintain of the characteristics defined at lab scale on the product made during scale one demonstrations J validation of the representativeness of the inactive glasses made in the laboratory or on the platform by making a glass of the same composition with actual radioactive waste in a hot laboratory and determining its characteristics. The fabrication of active samples is also necessary to study the effect of self-irradiation and the specific behavior of some radioactive elements. iii

During the development stage of reference matrix, since the main consideration is based of chemical incorporation, radioisotopes issued from FPs and MAs are replaced by their inactive representatives.

Waste Glasses

•

11

Sensitivity to chemical composition

In industrial situations, one must expect day-to-day variations in the composition of the solution to be vitrified, and also process upsets or variability which may affect the product. It is then necessary to determine what is the flexibility of the formulation toward these variations, in order to provide a wide enough operational domain, while keeping acceptable glass properties. For this, a sensitivity study is launched, to study systematically the effect of selected composition variations on glass properties, and establish acceptable limits for these components. For instance, in order to be pourable, the glass must display a viscosity lower than about 100 poise at the melting temperature. This will limit the possible increase in refractory elements such as SiO2 or Al2O3 or the decrease in fluxing components such as Na2O or Li2O. On the other hand, in order to keep an acceptable chemical durability, the increase in alkali oxides will be limited. In such a multi-dimensional space, with more than 10 constituents of interest in a formulation, sensitivity studies can become very cumbersome and involve a large number of crucible melts and the associated characterization measurements. In order to minimize the number of experiments while obtaining a reliably acceptable composition domain, statistical tools, propertycomposition models, and experiment design are implemented. These allow identifying acceptability limits and modeling matrix properties within these limits. For illustration, Table 2 displays the range of acceptable composition for the R7T7 glass produced La Hague. One can see that the acceptable variations may be small for some components, and much larger for other components.

3.3

Glass Characteristics of Interest

Glass characterization is the determination of all the physical and chemical properties of the product. Generally, and as observed in many laboratories, the products made with simulated waste by substituting the inactive counterparts for the radioactive components of the waste display properties that are very representative of those of the actual radioactive product. Glass characterization is thus generally performed on a simulated, non-radioactive glass for convenience, and validated in the end on radioactive glasses (see Section 3.2).

3.3.1

Microstructural homogeneity

Since it is generally more difficult to characterize and demonstrate a favorable long-term behavior for a chemically heterogeneous material, the objective is usually to produce a homogenous glass, or a glass that contains, after cooling of industrial-size blocks, only a small amount of harmless precipitates or insoluble phases. Nevertheless, the notion of homogeneity is relative and depends on the resolution of the measuring instrument used. Upon visual observation, HLW glasses such as the R7T7 composition are homogenous, black and shiny (Figure 8). The black color, analogous to that of basaltic glasses, is the result of the wide diversity of chemical elements included in the glass, and more particularly transition metals and rare earths, which absorb light over a wide range of wavelengths. By visual observation some small bubbles and some cracking associated to stress relaxation during cooling can be observed on industrial-size blocks. Conservative durability modeling accounts for the impact of these physical heterogeneity. Under the microscope (optical or SEM), a small volume of chemical heterogeneities can be observed, essentially noble metal inclusions (Pd, Rh, or Pd–Te alloy), ruthenium oxide, nickel, zinc, and iron chromites (Figure 9). This type of inclusions represents a small volume of the matrix, and has been shown to have no effect on the long term behavior of R7T7 glass. For other waste glass compositions, it is necessary to make sure that the precipitated phases do not alter the matrix properties, such as chemical durability. For instance, for Fe- and Cr-rich glasses, spinels have been observed, which do not alter glass chemical durability. On the other hand, for Al-rich glasses, it is necessary to avoid the formation of nepheline (aluminosilicate) crystals upon cooling, since these tend to be detrimental to resistance to aqueous leaching. In the end, the presence of precipitates or insoluble alloys in the molten glass may promote settling in the melter and have detrimental effects on the process (pouring difficulties, electrical short-circuiting in some types of melters). For the La Hague process, stirring allows processing melts with significant amounts of insoluble noble metals. Another form of heterogeneity is phase separation, whereby the glass separates (during melting or during cooling) into two or more different liquid or glassy phases. This type of phenomenon has to be carefully controlled to avoid the formation of a phase that is ‘weaker’ than the other, inducing some loss of general durability.

Figure 8 Obsidian glass and nuclear waste glass (cannot be distinguished with the naked eye).

12

Waste Glasses

50 µm

Mo

Al

Si

Fe

Cs

20 µm

Palladium-Tellure

Oxyde de ruthénium Chromites Figure 9 SEM observation on a zone of platinoïde concentration into the glass.

Table 3

Comparison of physical properties of selected nuclear and industrial glass compositions

Density (kgm3) Viscosity at 11001C (P) Tg (1C) Expansion coefficient (106K1) Thermal conductivity (Wm1K1) Young’s modulus (1010Pa) Fracture toughness Klc (106 Pa m1/2)

Typical HLW glass compositions

Pyrex glass

Window glass

2.50–2.75 50–150 510 8.3–9.9 1.0 8.4–8.6 0.75–0.95

2.28 80 000 565 3.2 1.09 6.1 0.85

2.46 4000 527–547 9.3 1.05 7.3 0.70–0.80

Nevertheless, in some specific instances, it may be acceptable to formulate microcrystalline or phase-separated glasses, if it can be demonstrated that this does not alter chemical durability. Such an approach has been retained in France to formulate glasses to immobilize molybdenum-rich solutions resulting from the processing of U–Mo fuel. Lower resolutions of observations (such as the one reachable by TEM, as well as structural techniques) are not relevant to link the heterogeneities at that scale to a significant effect on durability, but are of great use for understanding sakes.

3.3.2

Physical properties

The physical properties of nuclear waste glasses are usually quite comparable to those of classical industrial glasses,7 as illustrated on Table 3 below. Most of them have to range in between a minimum and a maximum value to suit with industrial processes. The density of waste glasses is slightly higher than those for the industrial glasses, owing to the presence of heavy metals. Viscosity at 1100 1C is much lower. This comes from the fact that nuclear glasses are formulated to be poured at 1100 1C while the

Waste Glasses

13

melting temperature of industrial glasses is significantly higher. It should also be noticed that the presence of noble metals or others heterogeneity in the nuclear glass can significantly modify its rheological behavior. Glass transition temperature occurs in similar ranges, although it is slightly lower for nuclear waste glass. The thermal expansion coefficient of nuclear waste glasses is similar to that of window glass, and significantly higher than that of Pyrex: Pyrex is formulated specifically to resist thermal shocks. Thermal conductivity is similar for all three types of glass. It should be noted that this low value is significant in nuclear waste glass which hold heat-generating FPs: temperature at the center of the glass block will need to be controlled. Young’s Modulus, which characterizes rigidity, is slightly higher for nuclear waste glass, while fracture toughness is similar. Electrical resistivity, not included in the table is another significant parameter if the glass is heated by direct electromagnetic induction or within electrodes (in the case of Joule Melter technology, this property can be neglected). Electrical resistivity decreases when temperature increases and when the alkali content of the glass increases. Electrical resistivity depends mainly on ionic diffusion in the material. It decreases from around 1.5  104 at 500 1C to a few O cm at 1200 1C. Other parameters not included are the thermal conductivity and the redox behavior. Thermal conductivity impacts directly the energetic efficiency in the melter. If too high, the thermal losses out of the melter make the process energetically (and economically) inefficient; if too low the energy transmitted to the glass is non homogenous. Regarding redox behavior, the presence of multivalent elements (Ce oxide for instance) in the glass composition may induce foaming at high temperature due to their reduction. The monitoring of oxygen partial pressure in the molten glass, and if necessary the use of redox buffer (FeII/FeIII) as well as physical means can overcome this inconvenience.

3.3.3

Thermal stability and crystallization potential

Devitrification is the process by which the glass looses part or all of its glassy nature through crystallization. It depends on the glass composition and its thermal history. For instance, increasing the levels of FPs, noble metals, molybdenum, phosphorus, chrome, nickel, iron or magnesium can favor crystallization in a nuclear waste glass. Furthermore, the time needed to reach the glass transition temperature (Tg) from melting temperature is also influent (this depends mainly on glass thermal conductivity and specific heat, canister geometry, and process parameters such as pouring rate). However, it is considered that once Tg is reached in cooling conditions, the devitrification process is kinetically frozen (cf. Section 4.1). Devitrification studies are based on subjecting glass samples to short duration heat-treatments (around 15 h) at stabilized temperatures and observing the heat-treated samples under the microscope to detect, observe, and quantify the crystals formed. Several indicators are determined:

• • • •

starting crystallization temperature: it is the temperature below which crystalline phases can be observed in the bulk of the sample after about 10 h of isothermal heat-treatment Crystallization temperature range: range in which these crystals are observed maximal crystal growth rate (generally expressed in mm mn1) Crystallization potential: it characterizes the ability of a composition to devitrify. It is the maximum percentage of crystals that can form after a heat treatment. This can vary from 0% (pure borate glass) to 100% (lithium disilicate glass). It is about 4% for R7T7 glass.

XRD and X-ray microprobe are used to identify the crystalline phases formed within a glass while image analysis and quantitative XRD are used to evaluate the percentage of phases formed. The amount of crystals that form in an actual large size industrial glass block is different from the maximal values found at lab scale: indeed, the thermal profile in the glass canister involves a continuous decrease of temperature, and is different in the various parts of the canister (cooling close to the canister wall is faster than in the center of the glass block). Several approaches can be used to bracket an estimation of the amount of crystals in the industrial glass block. In France, for instance, the maximum possible amount of crystallization is determined on laboratory samples, by subjecting them to a heat treatment designed to promote crystallization (5 h at 610 1C – nucleation temperature – and 100 h at 780 1C – maximum growth temperature – this cannot happen in a real glass block) and it is postulated that the amount of crystals in the glass block cannot be higher than the fraction determined in this way (which is in fact quite small for the R7T7 glass). In the US, where the glass blocks are quite large, crystallization studies are performed by establishing systematic TTT (Time Temperature–Transformation) diagrams and by considering the cooling profile at the center of canister (CCC – Canister Centerline Cooling curve), which is the slowest cooling part of the canister. Whatever the approach, the important aspect is the fact that crystallization must not be detrimental to glass durability. It is thus necessary to avoid depleting the glass matrix of elements that are favorable to durability, such as silica or alumina.

3.3.4

Chemical durability

Water is the major cause of glass alteration and radionuclide (or hazardous metal) dispersion during the life of the product, and more particularly during geologic disposal. The resistance of glass to aqueous alteration is generally called chemical durability. It is

14

Waste Glasses

the essential property required for a containment matrix. It is also essential for classical industrial glass compositions (container glass, window glass ...) and as such is also considered during their formulation. The process by which glass constituents are washed into water is called leaching, and it is the combination of a variety of mechanisms described in Section 4. Chemical durability is assessed by various ‘leach tests’ during which glass samples are contacted with aqueous solutions, under a large range of experimental conditions. The leaching behavior of glasses varies according to glass composition, test conditions, and time. There is no unique evaluation of chemical durability, and the performance of different glasses compositions can only be compared under the conditions of a given test. In order to fully evaluate the behavior and performance of a given glass, it is advisable to understand the various mechanisms that are involved during its interaction with water and the environment, and to design testing to obtain the parameters that allow modeling this performance. Several types of leach tests can be listed:

•

•

•

Static tests, during which glass samples, either monolithic or crushed to powders, are exposed to a solution and left standing there for the duration of the test. During these tests, the components dissolved from the glass progressively accumulate into the solution and are left to interact between one-another and with the sample. Two of these tests have been normalized in the US: the so-called ‘MCC-1’ test on monolithic samples (also known as ASTM-C1220) and the ‘PCT’ test on crushed glass samples (also known as ASTM-C1285). Both these tests are available with several variants (temperature, volume of solution, type of solution, etc.) Flow-through tests, during which the sample is exposed to a continuous flow of fresh leachant to prevent the accumulation of reaction products into the solution, and, thus, to test the ‘initial’ alteration. Among those tests, one can list the ‘Soxhlet’ test (also known as ISO 16797:2004), by which a monolithic sample is exposed to a continuous flow of condensed water at 100 1C re-circulated in a distillation apparatus, or the ‘flow-through’ test (also known as ASTM C1662–07) by which the sample is inserted in a column or a container and subjected to a continuous flow of fresh solution. ‘Integral tests’ or ‘Service conditions tests’ or ‘Tests including environmental materials’ are tests designed to account for the overall service environment expected during the life of the product. Glass alteration may be measured in three different ways:

• • •

Elemental analysis (ICP-MS, ICP-AES, etc.) of the leachate (solution after contacting the sample), in order to obtain information on the leaching kinetics for each element, as a function of time. Weighing of the sample, to determine the overall weight loss. Analysis of the alteration rind on the sample (SEM; SIMS, STEM, etc.) in order to study the thickness of the altered layer and to understand the fate of the various elements released from the glass and retained in the alteration layer.

The quantitative expression describing the rate of release of an element in water is the ‘leach rate,’ usually expressed in g m² d1, and normalized with respect to the mass fraction of the given element in the glass. LR ðiÞ ¼ mi =x i :S:t

where mi is the weight (or activity) of element i released into solution, xi is the mass fraction (or specific activity) of this element in the glass, S is the area of the glass surface exposed to the leachant, and t is time. During a static test the ‘normalized mass loss’ is often used as an indicator: NLðiÞ ¼ mi =x i :S:; expressed in g m2 The evolution of NL(i) ¼ f(t) describes the overall kinetics of the process for the given experimental conditions. This curve is the basis for all the glass behavior studies. If an element (i) is a good alteration tracer, i.e., congruently released with glass dissolution and not trapped back into the alteration products, then the equivalent thickness of altered glass Eth can be calculated by dividing the normalized mass loss by the glass density r: Eth ðiÞ ¼ NLðiÞ=r Boron is most often a good tracer of glass alteration (it). Sodium, Lithium and Molybdenum are good tracers in dilute media, but they may be integrated in alteration products in more concentrated media. More information on glass alteration mechanisms is provided in Section 4. Exhaustive leaching characterizations are performed on inactive but chemically representative samples; some tests are performed on fully active material to check the impact of radioactive environment on leaching behavior.

Waste Glasses

4

15

Long Term Behavior of Nuclear Waste Glasses

The main phenomena that could alter glass containment properties over the long term are heat (for HLW only), radiations damage and alteration by water. Their occurrence is not expected at the same time scale. For instance the risk of crystallization is principally limited to the thermal phase, i.e., in interim storage over the few decades during which the maximum glass temperature will decay from about 400 1C to less than 100 1C For thousands of years the glass matrix is expected to remain dry and the major potential glass alteration mechanism is selfradiation damage. Can it change the glass containment properties when, after a few thousand years, canister and over pack breach, and glass alteration by water start? Many studies on glass self-radiation damage address this question. Eventually, on the very long term, the rate of radio nuclides release into the near field will be controlled by the rate of glass alteration by water. Worldwide, for over 30 years, large research efforts have been conducted to understand all the mechanisms of glass alteration by water and to develop comprehensive models, and to adapt them for the evaluation of repositoryiv performance.

4.1

Glass Crystallization and Long Term Thermal Stability

Thermal stability constitutes one of the prime criteria for glass selection. It underlies the preservation of a homogeneous glass over time. Theoretically, glass could naturally evolve to a crystalline state thermodynamically more stable. But such transformation, thermally induced, gets dramatically slow (or even stops), when the glass is maintained at temperatures lower than the glass transition temperature (Tg). Predicting thermal stability at low temperature and in the long term therefore requires experiments performed in supercooled liquids as well as modeling. For nuclear glasses the main work in this field was performed by X. Orhlac,8 which helped confirming the thermal stability of R7T7 glass on the very long term. Devitrification experiments conducted on this glass,9 made it possible to identify three major crystalline phases (CaMoO4, CeO2, and ZnCr2O4) and two minor phases (albite NaAlSi3O8 and silicophosphate) between 630 and 1200 1C. Yet, their crystallization remains limited (a maximum 4.24 wt%), since these phases consist of glass minor constituents (Figure 10). Even after a heat treatment designed for a maximum crystallization (100 h at 780 1C), no change can be observed in the main properties of the nuclear glass (chemical durability and mechanical properties). Plotting the nucleation and growth curves of these phases highlighted several essential points:

• • •

nucleation sharply emerges during the first hours of the treatment, then stops beyond this period of time. Nucleation is heterogeneous, inducing crystallization on the pre-existing active sites. Moreover, nucleation curves are strongly amplified and shift to lower temperatures in the presence of insoluble noble metal particles; seed crystal growth is very low, and, after a few dozen hours, displays a saturation phenomenon; strong nucleation coupled with slow growth globally leads to a material which can hardly be devitrified (Figure 11).

The stability of high-level R7T7-type nuclear glass at low temperature and in the long term was then investigated by modeling. The mathematical model selected is based on the KJMA theory (KJMA, for Kolmogorov, Johnson, Mehl, and Avrami), and describes the transformation kinetics as a function of time and temperature.10 Atomic diffusion is the main factor which limits crystallization, as demonstrated by measuring the diffusion activation energy. Viscosity is therefore the key parameter which determines the nucleation-growth kinetics in glass: it is this very parameter which conditions diffusive atomic transport in the silicate-based liquid. Consistently, nucleation-growth kinetic processes can be determined by means of independent viscosity measurements in a broad range of temperatures.

Figure 10 The sequence of alteration of a vitrified waste package.

A repository is the final destination of HLW glass, a disposal site selected and engineered to definitively isolate the radioactivity contained within the waste from the biosphere and man.

iv

16

Waste Glasses

Figure 11 Temperature range for the nucleation and growth of the main crystalline phases likely to be formed after a devitrifying thermal treatment of a borosilicate nuclear glass, which confines high-level effluents arising from UOX fuel treatment From CEA DEN Monographs ‘Nuclear waste Conditioning.’ Etienne Vernaz, Éditions du Moniteur: Paris, 2009: ISBN 978–2–281–11380–8.

The model validation was achieved under isothermal conditions on a simplified barium disilicate glass, known for its homogeneous, fast crystallization. Applying this model to the R7T7 glass shows that periods of several millions of years are required for the three main phases to be completely crystallized at any temperature below 600 1C. Clearly, if during the slow cooling of large industrial block there is no, or minute, crystallization in the temperature range 900–600 1C, then there will be neither other crystallization, nor crystal growth on the long term, for kinetics reasons. These results confirm the thermal stability of actual high-level waste confining glasses.

4.2

Glass Resistance to Self-Irradiation

The main source of irradiation in nuclear glasses result from a decays from actinides, b decays from FPs, and g transitions accompanying a and b decays.11 Alpha disintegrations is characterized by the production of an heavy recoil nucleus (RN) and the emission of a light a particle, yielding a helium atom at the end of the path. RN, shedding large amounts of energy over a short distance result in atom displacement cascades, thus breaking a large number of chemical bonds. Alpha decays are thus the main cause of atomic displacement. On the other hand beta disintegrations produced by FPs and gamma transitions lead mainly to electronic interactions (electron excitation, ionization) with the glass network atoms but not to atomic displacements. The effects of self-irradiation have been studied by investigation of glasses doped with short half-life actinides, by atomistic modeling, and by external irradiations.

4.2.1

Investigations of glasses doped with a short half-life actinide

This investigation method is the most representative of nuclear glasses ageing: isotropic irradiation is produced in the whole of the glass volume preserving the electrical neutrality (unlike external irradiations), alpha particle and the RN are produced allowing electronic interactions and atomic displacements, and eventually helium builds up in the glass, exactly as it will be in the real case. The effects of alpha disintegrations were mainly investigated through studying 244Cm-doped-glass glasses that can integrate within a few years doses equivalent to those to be delivered to the nuclear glass for thousands of years under disposal conditions.12 Figure 12 shows the ‘DHA’ Atalante laboratory. It is an example of a shielded line at Marcoule (France) devoted to high-level waste studies, where actinide-doped glass samples are fabricated to carry out ‘accelerated’ studies of the effects of self-irradiation. The inset in the figure shows a 238Pu-doped glass block manufactured in the Vulcain laboratory in 1975. Results produced all over the world, in France,13 UK,14 US,15 or Japan,16 are quite consistent. Due to the effect of alpha decay the glass density decreases slightly (Figure 13) and its mechanical properties appreciably improve, especially fracture toughness that characterize glass resistance to cracking. The variations in these properties reach a saturation level and stabilize beyond 2  1018 /g.

Waste Glasses

17

Figure 12 The Atalante laboratory for high-level waste ‘DHA.’

Figure 13 Evolution of the density of curium-doped glasses with the a-decay dose (from Ref9).

Up to a dose of 1019 a/g, no helium accumulation (He bubble) is observed. No significant change in glass durability is observed. Furthermore there is no dose rate effect, as variations can be reproduced among the various glasses under study which exhibit quite different integration rates, spreading over four orders of magnitude.

4.2.2

Atomistic modeling of glass self-irradiation

The second approach to understand radiation damages focuses on atomistic modeling. In particular, molecular dynamics can provide insight into the ballistic effects induced by the deceleration of RNs emitted at the end of a decay. Numerous studies conducted on simplified glasses representative of the basic matrix nuclear glass (SiO2, B2O3, Al2O3, Na2O, and ZrO2) demonstrated the remarkable capacity of this type of glass to restore its structure following the passage of a RN. The following conclusions could be drawn from the whole of the calculations performed in relation to individual cascades in glasses: Displacement cascades take place in two separate steps (Figure 14):

•

the ballistic stage during which collisions between atoms take place as a whole. This phase coincides with the strong heating of the matrix and a depolymerization of the structure by interatomic bond breaking. In parallel, a decrease in atom density can be observed within the cascades;

18

Waste Glasses

Figure 14 Evolution of displacement cascade from the initial glass to the reconstruction of the glass structure after the ballistic phase (from Ref9).

•

the relaxation stage during which glass structure reconstruction takes place. The glass structure then experiences significant reconstruction to a state close to its initial state, but still with a slight structural depolymerization and a slight swelling on the whole;

In Figure 14, four stages of the evolution of displacement cascade can be viewed. In the top left corner, the initial glass containing the uranium atom (light blue atoms) to be accelerated with a 800 eV energy (t¼ 0 ps). In the top right corner, the start of the ballistic phase induced by the uranium projectile (t¼ 0.013 ps). In the bottom left corner, the final step of the ballistic phase when the maximum number of broken bonds can be observed (t¼ 0.038 ps). In the bottom right corner, reconstruction of the glass structure after the ballistic phase (t ¼ 0.25 ps). A model of cumulative local quenching was built from these data in order to help explain the origin of the small evolutions observed under irradiation, as well as the origin of their stabilization under high doses.17–19 As displacement cascades accumulate, glass structure is fully destabilized by nuclear interactions. Then the material can be quickly reconstructed without any external energy and its structure is close to that a glass frozen in at high temperature, which results in the observed evolutions. When the whole of the glass volume has been damaged once by the displacement cascades, any new alpha disintegration produced will again temporarily destabilize the structure, but the latter will be rebuilt in the same way as after the first damage. So the glass no longer undergoes significant change, which could explain why its properties are stabilized beyond a given dose. It is worthwhile mentioning that the saturation dose experimentally observed in relation to macroscopic property evolutions (as seen in Figure 13) coincides with that required for full glass damaging by displacement cascades, which corroborates the proposed model. As a conclusion, the insignificant evolution of nuclear glasses under alpha self-irradiation with respect to crystallized minerals could be explained by the self-repairing properties of the glass structure.

Waste Glasses 4.2.3

19

External irradiation of glasses

This complementary investigation axis is based upon the use of nonradioactive glasses in which irradiation stress is simulated by external irradiation techniques (neutrons, heavy ions, electrons, g). The major disadvantages of this experimentation type consist of the upsetting effects of injected high dose rates into low irradiated volumes. Today, accurate knowledge of these offsets allows relevant effects to be sorted out from experimental artefacts and the results obtains are quite comparable to those obtained with other investigation methods. This type of study was undertaken as early as the 1980s to evaluate the effects of beta disintegrations. Glasses with chemical compositions representative of the industrially produced nuclear glasses were thus irradiated by electrons during one year up to doses equivalent to those received in about 1000 years of disposal, i.e., 70% of the total bg dose received in 1 million years. These irradiations have not entailed detectable modifications of the macroscopic properties (density, mechanical properties). In addition, glass still displays a homogeneous microstructure after irradiation.

4.3

Nuclear Glass Alteration by Water

The mechanisms which control nuclear glass leaching kinetics have been investigated worldwide for more than three decades. This large accumulated knowledge allows building computational models likely to be used for performance assessment of a geological repository. These models have to be applicable to all the vitrified waste packages industrially produced, taking into account many different environmental conditions.

4.3.1

Basic mechanisms of glass alteration

In contact with water the main alteration mechanisms of borosilicate glass are the following20,21:

• • • • •

Exchange and hydrolysis reactions involving the mobile glass constituents (alkalis, boron, etc.) rapidly occur during the initial stages. Slower hydrolysis, especially of silicon, drives the initial glass dissolution rate. The in situ condensation of many hydrolyzed species (Si, Zr, Al, Ca, RE, etc.) results in the creation of an amorphous gel layer at the glass/solution interface regardless of the alteration conditions. This layer is more or less reorganized by hydrolysis and condensation mechanisms according to the environmental conditions. This amorphous layer can soon constitute a barrier against the transport of water toward the glass and of solvated glass ions into solution. The existence of this transport-inhibiting effect rapidly causes this layer to control glass alteration when water renewal becomes very low. Some glass constituents released from the glass during process can precipitate as crystallized secondary phases. The precipitation of these crystallized phases within or on the top of this amorphous layer or in solution, can sustain glass alteration by consuming the elements that form the protective barrier.

4.3.2

Initial rate of glass dissolution

The initial dissolution rate ro is the hydrolysis rate obtained in pure water when no diffusion barrier slows down the kinetics of alteration. It is an intrinsic property of the material that characterizes its chemical durability in water. This rate depends mainly on the glass composition, the temperature, the pH and to a lower extent to the solution composition.22 It is generally considered as the maximum rate of glass alteration for a given temperature and pH. With an activation energy of about 75 kJ/mole, r0 ranges over seven orders of magnitude between the temperature of 4 and 3001 C. It is the reason why natural glasses exhibit very low alteration at room temperature even after millions of years but high alteration under hydrothermal condition (‘hyaloclastites’). This value explains also the inherent difficulty of measuring r0 at room temperature, (alteration of about 1 nm per day hindered by interdiffusion) while a ‘Soxhlet’ measurement operating at 100 1C will allow a prompt measurement, representative of the sample in its mass.

4.3.3

Alteration rate in saturated conditions and final rate of glass dissolution

In a closed system or under conditions in which the water renewal rate is very slow, such as in the case of a geological repository, apparent silica saturation is observed in the leachatev and a strong ‘rate drop’ is systematically evidenced for borosilicate glass. This rate drop has been related both to affinity effects (a decrease in the hydrolysis rate coupled with an increase in the concentrations in solution), and to the formation of an alteration gel standing as a diffusive barrier between glass and solution.23 Once this ‘saturation’ conditions are established, a steady rate of glass alteration is generally observed. This ‘residual rate’ seems to be related to the phenomenon of gel dissolution and to the rate of secondary phase precipitation. For most nuclear borosilicate glass composition this final rate is very slow (about 5 nm/year at 50 1C for R7T7-type glass). ‘Leachate’ refers to the leaching solution after contact with the glass sample.

v

20

Waste Glasses

The origin of such a small residual rate of glass alteration observed in saturation conditions seems to be related to the persistence of a weak dissolution rate of the gel probably itself resulting from the slow precipitation of secondary phase (mainly phyllosilicates if pH does not exceed 10).24,25 In some specific cases ‘alteration resumption’ can be observed, once a residual rate regime has been established. This so are called ‘renewed alteration’ is observed only for very alkaline conditions (equilibrium pH higher than 10) and specific glass composition (with high Al/Si ratio). This can be related to a high precipitation rate of some specific secondary phases (zeolites).26

4.3.4

Essential role of the ‘Passivating Reactive Interphase’ (PRI)

Historically, this apparent saturation state described above was expressed in equations as if equilibrium with the fresh glass could be achieved. Today it is considered that a saturation state can only be achieved with respect to a hydrated layer. As saturation is approached in solution, the rate of condensation of many gel forming element (Si, Al, Zr, Ca, etc.) increases, allowing the formation of a thin amorphous layer. Frugier et al. proposed the term ‘Passivating Reactive Interphase’ (PRI) to take into account the fact that not all of the gel layer becomes passivating but only a thin inner layer in which a high condensation rate has led to closed porosity.27 Diffusion coefficients in the PRI are consistent with diffusion in solids with values of the order of magnitude of 1020m2 s1. Furthermore Monte Carlo modeling of the gel layer formation by hydrolysis and condensation mechanisms allows describing the conditions for which porosity closure is reached, in good agreement with experimental data.28 Recent work delved into these processes and showed that (1) the closure of the porosity is not a prerequisite for accounting such low diffusion coefficients, (2) the porosity of PRI formed in silica saturated solution remains open but water molecules are trapped in pores o1 nm formed by repolymerization of the silicate network following the release of mobile species, and (3) diffusion profiles within the PRI require a new interdiffusion model taking into account the water dynamic at nanoscale.29 It will be noticed that phosphate glass, that can in some cases,30 display quite low initial alteration rates, are not expected to form any PRI, and then no large rate drop can be expected in saturation condition.

4.3.5

Influence of glass composition

Within a given domain glass dissolution kinetics strongly depends on its composition.31 Some non-linear effects have been evidenced based on semi-empirical statistical methods,32 or, in few cases, fully explained using experimental approaches and Monte Carlo numerical model. A given element usually modifies specifically each kinetic regime. As an example Ca has no effect on the forward rate but strongly favors the rate drop as it is incorporated into the PRI contrarily to Mg that is incorporated in secondary clayey minerals playing as a silicon sink and promoting higher dissolution rate. From operational point of view, rates of actual glass or rates corresponding to the best or the worse glass within a given domain are calculated from empirical equation built from a limited number of tested glasses. Considering the R7T7 domain, forward rate ranges from 1.6 to 4.1 g m2 d1 at 100 1C and residual rate from 2.7  105 to 3.2  104 g m2 d1 at 90 1C in initially pure water.

4.3.6

Influence of groundwater and environmental materials

In a geological disposal, water chemistry initially in equilibrium with the host rock will be disturbed by engineered materials (stainless steel canister, overpack, liner, concrete, etc.) and also by the heat produced by the glass canister itself. As a consequence it is expected that glass dissolves in an open medium in which temperature, water flow and composition evolve with time and space, at least during the first ten of thousand years (this time depends on the host rock and the disposal design). Chemical elements brought by water, like Ca, Mg, organic matter, etc may influence glass dissolution mechanisms for example by promoting the PRI condensation,33 increasing the hydrolysis rate of the silicate network,34 or the allowing the precipitation of secondary crystalline phases.35,36 In fractured rock environment, such as a granitic disposal, the water renewal rate will be the main environmental parameter. In a clay environment, with no flow rate (or too low compare to diffusion), the predominant effects of environmental materials will be silica sorption onto likely oxide and hydroxide minerals and low precipitation of silicate minerals that acts as silica sink.37 In any environment those phenomena can also be expected with iron corrosion product. This kind of reaction will maintain a high glass dissolution rate until the close environment of the glass is saturated. Beyond this transient regime that can be investigated by reactive-transport codes the final rate regime will control long-term glass dissolution. Predicting such final rate is a challenge that requires specific integrated mock-ups, in situ tests, simulation by reactive-transport codes and validation by natural or archeological analogues. Finally, in salt rock, lower rates than in pure water are expected, especially if the water is weakly renewed. Numerous studies have investigated effects of ionic strength and chemical composition of the brine.38–41 Because most of these studies are quite ancient, and also because the prediction of the water availability near the glass and the migration of soluble species in salt is more complicated that in other host rocks, their conclusions, namely in term performance assessment should be revisited as mechanistic knowledge has been improved.

Waste Glasses 4.3.7

21

Influence of glass fracturing

Since glass is fractured after cooling, the reactive surface is greater than the geometrical one. Neither the surface of an actual glass block nor the evolution of the reactive surface in geological disposal condition have been precisely determined nor estimated. Several experimental techniques have been carried out to investigate glass blocks cracking networks and determine their impact on glass life time.42 Considering an inactive R7T7 glass block, largest cracks are estimated to increase the geometric surface by a factor around 5 and smallest ones by a factor around 40. Up to now all countries have calculated glass package lifetimes considering a constant cracking factor. Mechanistic modeling under development will help addressing this issue in the next future. However most studies on natural or archeological analogues shown that inner cracks have generally a minor contribution to the overall glass alteration (cf. Section 4.3.9).

4.3.8

Modeling glass long term behavior

Predicting long-term behavior of glass requires a multiscale approach as space and time scales related to key phenomena are too large to be simulated by a single mechanistic model. As a consequence discrete modeling approaches have been developed from ab initio calculation at atomistic level,43 up to performance assessment model at macroscopic level (also called operational models).44 In between Monte Carlo model allows to bridge the gap between atomistic level and measured dissolution kinetics.45,46 One key point is the glass dissolution rate law. Many mechanistic rate laws have been proposed.47–49 The most advanced one is probably the ‘GRAAL’ model proposed by Frugier et al.21 In this model the glass-related parameters are the solubility limit of the PRI, the water and solvated ions interdiffusion coefficient in this interphase, the PRI dissolution rate. The other model parameters are relative to secondary phases likely to precipitate, depending on the chemical elements supplied by the glass or by the surrounding medium: phase solubility limits and precipitation kinetics. For R7T7 glass a very good agreement is observed between simulation and experimental data for a very large set of experimental conditions.50 The operational model that is proposed to assess the R7T7 glass long term behavior in the proposed setting for the French geological repository, is the so-called ‘r0-rf’ model.44 In this model the rate of glass alteration is supposed to keep its initial r0 value until all conditions are obtained to get the final rate (i.e., full saturation of the media, all silica sorption site saturated). Then the final rate rf is applied. For R7T7 glass the model parameters (alteration rates r0 and rr, glass surface area accessible toward water) were determined as a function of temperature, pH and glass composition throughout the whole R7T7 composition range. Uncertainties on the parameters values were also determined. The model can be used to calculate the glass block lifetime depending on the time/temperature profile, the pH of the medium, the date of water ingress in contact with the glass, and the quantity of accessible silicon sorption sites on the metal canister alteration products. A typical calculated glass lifetime plot is given in Figure 15. Two assumptions concerning the quantity of unsaturated sorption sites during the initial rate phase are proposed and water ingress in contact with the glass after 4000 years,51 is assumed. Should the environment be saturated, then the total package lifetime of a R7T7 glass will exceed 300 000 years, even in a pessimistic scenario where large amount of iron corrosion product act as silica sink.

4.3.9

Natural an archeological analogues

Validating predictive models is one of the major difficulties of investigating the long-term behavior of containment materials because the relevant time scales largely exceed what is accessible to laboratory experimentation. Whenever possible, therefore, 100,0%

_______ _______

60

Masse of Corrosion Products = 2730 kg Masse of Corrosion Products = 100 kg

Numerical uncertainties (1σ)

10,0%

40

30

1,0%

20 Two environmental conditions 10

0,1% 1000

10000

100000

1000000

0 10000000

time (years)

Figure 15 Operational modeling; example of calculation of R7T7 glass lifetime in two environmental conditions.

Temperature (°C)

Total Fraction of Altered Glass (%)

50

22

Waste Glasses

Figure 16 Predicted percentage of alteration for Embiez Glass (curves) and measured alteration of both kind of surfaces (stars). Geochemical modeling has been achieved using the Hytec code including the GRAAL model for glass dissolution, water diffusion within smallest cracks, advection within the largest ones and specific boundary conditions related to 56 m-deep seawater.

natural or archeological analogues are examined for this purpose. They enable to check that no long term mechanism is forgotten. They give us some very long term integrated experiments, against which predictive models can be qualitatively validated. For instance, archeological glass blocks from a shipwreck discovered near the French Mediterranean island of Les Embiez have been examined because of their morphological analogy with nuclear glasses and their known, stable environment. Like nuclear glasses, these blocks were fractured after production; they were then leached for 1800 years in seawater.52 In that specific case a quantitative agreement has been achieved between geochemical simulation and measurements on the archeological artefact showing that bridging the gap between short term laboratory data and long-term natural system is possible via a rigorous methodology (Verney-Carron et al., 2010)53 (Figure 16). The same methodology could be applied to much older basaltic glasses for which the environment can be characterized. These glasses not only exhibit the same alteration mechanisms and kinetics as nuclear glasses in laboratory experiments, but their alteration products also reveal strong similarities, especially between the palagonite on basaltic glasses and the gel on nuclear containment glasses, which can constitute a diffusion barrier. These studies can contribute to a finer definition of the chemical model of nuclear glasses and to the long-term validation of the gel protective properties.54,55

4.4

Conclusions on Glass Long Term Behavior

For more than 30 years a very significant research effort on nuclear glass alteration mechanisms has been carried out worldwide, and a large data base has been produced. Academic researches on long term crystallization, radiation damage and alteration by water, enable nowadays a good mechanistic understanding of the key phenomena that can alter nuclear waste glass properties on the long term. A sound methodology was established to use the best of academic knowledge of alteration mechanisms for performance assessment of glass package in complex environments. This methodology includes:

• • •

Assessing the evolution of the boundary conditions including normal and incidental scenario of evolution. Understanding elementary alteration processes at a mechanistic level. Assessing the couplings between the different mechanisms which simultaneously occur within a given scenario. Such couplings may indeed modify significantly the global evolution.

Finally, the models describing the different processes have to be integrated in a global predictive model which often requires to be simplified by selecting the most significant processes and parameters. Operational models also include a conservative approach to overcome the lack of knowledge and wrap the general trend. The use of this kind of operational model demonstrates that waste glass life time can be over millions of years if the glass composition is optimized and disposal conditions appropriate. Furthermore, through this long-term research on waste glass a new ‘science of long term behavior’ has been developed. This science and methodology is now applied to numerous other matrixes (cement, bitumen, spent fuel, etc.).

Waste Glasses

23

Figure 17 The Piver process in France (1969–1980).

5

Vitrification Processes

As nuclear energy is a very concentrated one, the overall volume of nuclear waste is small. This is especially true for HLW that will be concentrated into a small volume of glass.vi Consequently the scale of radioactive waste vitrification facilities is usually much smaller than that found in traditional glassmaking. In addition, and especially when processing high-level waste, the very high levels of radiation preclude direct contact with the equipment. Any waste resulting from exchange of failed equipment, for instance, becomes radioactive waste and must be managed as such. Consequently, high-level waste vitrification facilities must be designed to be remotely operated, and to minimize maintenance as well as secondary waste generation. Off-gas treatment systems must be very efficient to remove any volatilized or entrained radionuclide. Since most of waste streams are nitrate-rich, NOx fumes are produced and must be abated. The whole vitrification process must be contained efficiently in order to prevent release of radionuclides to the environment. Another significant difference between traditional glass-making and waste vitrification is that, most often, the waste is in a liquid form while, in glass making, the batch materials are dry solids. For waste vitrification it is then necessary to evaporate the liquid and calcine the salts prior to reacting them with the glass formers. This operation requires large amounts of additional energy provided directly in the melter or in a specific pre-treatment step. In the end, the glass product must be disposed of, usually in metallic canisters. For that purpose, most of the time, the glass product must be poured into these canisters. This requires, first, that glass viscosity be around 100 poise or lower at the time of pouring and, second, that the vitrification equipment be designed with a pouring function. According to the nature of the waste to be vitrified, and to the context, a number of processes have been studied, among which several have been deployed industrially. The first attempts,56 were extrapolations of the crucible work performed in the laboratories. The process was performed batchwise in a single crucible, where all the operations of evaporation, calcination, vitrification, and evacuation of the product were performed successively, in a sequence. The melting crucible could be the canister itself (the process was then a ‘lost-crucible’ process) or a melter from which the glass product was poured into the canister. The first French industrial facility, PIVER (Figure 17), for instance, was of this type. The metallic melter was heated from the outside by a stack of inductors. The facility was used to process actual high-level radioactive waste into 100 kg glass blocks. Similar facilities, operated with lostcrucibles or not, were designed or built in various other countries (UK, Italy, etc.). Very soon, however, it was concluded that batch processes did not allow throughputs compatible with commercial operation. The PIVER throughput, for instance, was around 5 kg/h of glass. Most countries, then, decided to abandon batch processes and design continuous vitrification processes, with two major options for feeding the waste:

• •

one-step processes, where liquid waste is fed directly to the melter and all the steps of evaporation, calcination and vitrification are performed in it. This is the case for instance of the Defense Waste Processing Facility at Savannah River, USA two-step processes where liquid waste is first fed to a calciner before entering the melter. This is the case for instance in France, at the AVM facility at Marcoule or at the R7 and T7 facilities at La Hague.

vi For example the amount of HLW glass produced each year in France, related to the reprocessing of the spent fuel of about 50 reactors, is in the range of 100 m3.

24

Waste Glasses

Figure 18 The French two-step continuous vitrification process.

In the following sections we will describe the major existing facilities and the emerging new processes which are being designed to further improve the capabilities and efficiency of these processes.

5.1 5.1.1

Existing Processes for Radioactive Waste Vitrification The French two-step continuous vitrification process

Following the PIVER experience, the French CEA started to develop a two-step process, in order to separate the functions of evaporation-calcination and vitrification, in Figure 18. This allows keeping a melter of relatively small size, since most of the energy is provided at the level of the calciner. Another major decision was to select a vitrification method by which power is supplied to the glass from the outside, without direct contact of the glass with the power source. A metallic melter heated by induction provided by an external stack of inductors was selected, following the good results obtained with PIVER. This disposition allows protecting the power source from contamination. On the other hand, the size of the melter is limited by the ability to transmit heat to the core of the molten bath. The first industrial facility for vitrifying high-level waste in France was the Atelier de Vitrification de Marcoule (AVM) which was the first industrial vitrification facility in the world, commissioned in 1978. This facility has vitrified the high-level waste solutions from the UP1 reprocessing plant and has then been used to vitrify the effluents resulting from the decommissioning and decontamination of the same UP1 plant. This mission has now been completed, and the AVM facility is now being decommissioned after more than 30 years of successful operation. The experience gained from the operation of AVM has been later incorporated into the design of the larger facilities R7 and T7 at La Hague, with 3 vitrification lines each, which started operation in 1989 and 1992, respectively. The same technology has been selected for the WVP (Waste Vitrification Plant) at Sellafield in the UK (Figure 18). In the French continuous process used at La Hague, the concentrated high-level waste solutions are received and stored in cooled and stirred tanks. After sampling and analysis, they are fed at a metered rate to the calciner. In the calciner they are heated progressively up to about 400 1C to evaporate the liquid and transform them into a finely divided powder called the calcine. The calcine falls into the melter, together with glass formers which are fed under the form of a pre-fabricated glass called frit. The mixture is heated at the surface of the molten glass bath and undergoes the final vitrification reactions (from temperatures of about 700 1C) and finally become digested into the homogeneous molten glass at about 1100 1C. The molten glass is then poured batchwise into metallic canisters which are then weld-sealed and evacuated. The calciner is a tilted rotating tube inserted into a furnace heated by resistors. In the calciner, the solution is evaporated and most of the nitrate salts (with the exception of alkalis) are converted to oxides by decomposition of the nitrates. A calcination additive, which decomposes under the action of temperature and reacts with the nitrates, is added to ease the fragmentation of the calcine and to limit the volatility of some radionuclides. The calcine then falls into the melter together with the glass frit. The melter is a metallic crucible made of nickel-base alloy heated by induction. In order to promote heat transfer and enhance melt homogeneity, the melter is equipped with stirring and gas sparging devices.

Waste Glasses

25

Figure 19 The French La Hague HLW glass storage facility.

The melter fills progressively during continuous feeding. When the higher operating level is reached in the melter, a batch of 200 kg of molten glass is poured into a stainless steel canister through a pouring nozzle situated below the melter. Pouring is activated by heating the nozzle with a specific inductor. The melter then continues to process the next batch. Each glass canister holds two batches of 200 kg of glass. After filling and cooling, the glass canisters are closed tightly by welding a cover on top of their mouth. The sealed canisters are decontaminated by shot-blasting and checked for absence of residual contamination. They are then transferred to a storage facility (Figure 19) where they are stacked in pits cooled by a forced flow of air to evacuate the residual heat produced by radioactive decay of the FPs. At the time of production, the heating power of each individual canister can be higher than 2 kW. After several years of cooling in a forced ventilation storage facility, the residual power has decreased enough to allow transferring the canisters to a facility cooled by natural convection. The off-gas from the calciner and the melter is composed of water vapor, nitrogen compounds and entrained material. It is extracted at the top of the calciner and goes through a dust-scrubber, to remove most of the large particulate material and aerosols for recycling to the calciner, a condenser, washing columns and filters to decontaminate the gas prior to release to the stack in compliance with radioactive and chemical release standards. The liquid effluent from off-gas treatment is collected and treated in specific effluent treatment facilities to concentrate the activity and re-cycle most of it to vitrification. This technology is now used in the French industrial facilities of R7 and T7 at La Hague and has proven its efficiency and operability. By the end of 2014 about 22 200 glass canisters have been produced (of which about 3500 at the AVM facility and about 18 700 at La Hague). This amounts to more than 8500 t of glass and around 3  108 TBq of activity safely immobilized. The small size and modular design of the technology makes it easily operable and maintainable. The major limitations of this process are:

•

• •

The life expectancy of the metallic crucible which is in direct contact with hot (B1100 1C) and corrosive molten glass. Through continuous developments it is now possible to replace the metallic melters about once a year in the current La Hague facility. On the other hand, this operation is made easy by the small size and design of the facility; melter exchange is performed within a week, and the resulting waste can be size-reduced and processed together with the other metallic waste of the reprocessing facilities. Capacity: since the glass is heated from the outside, the size of the melter, and thus the capacity, is limited. In the current facilities, the capacity of a metallic melter is about 25 kg/h of glass, and several lines are needed if a higher throughput is required. On the other hand, this small size is also an advantage for maintenance and waste generation, as seen above. Limitation in melt temperature: in order to preserve the integrity of the metallic crucible, operation temperature is limited to about 1100–1150 1C. This, in turn, is a limit for throughput (since throughput theoretically varies like T4). Moreover, this temperature limit also reduces the range of glass compositions that can be processed in such a facility to those who have melting temperature below or around 1150 1C.

5.1.2

Liquid-fed ceramic melters

Not all the liquid waste is amenable to separate calcination: when the waste holds large amounts of alkalis, the corresponding nitrate salts tend to form molten phases in the calciner and prevent adequate calcination, generating numerous sticking and caking problems. In such situations, direct liquid feeding of the melter has been implemented. This is the case in the USA (Defense Waste Processing facility – DWPF, West-Valley Demonstration Plant – WVDP, Hanford WTP), where the acidic high-level liquid waste has been neutralized by caustic prior to storage. Direct liquid feeding can also be selected with the intent of keeping only one processing step such as in Belgium (PAMELA), Germany (VEK), Russia, or Japan (Tokai Vitrification Plant – TVF and K-Plant in Rokkasho Mura). In these situations, the liquid waste and glass frit (or separate glass formers) are fed continuously to the top of the melter, and evaporation, calcination, and vitrification reactions are performed in the ‘cold cap,’ a colder layer that sits on top of the molten

26

Waste Glasses

bath and progressively dissolves into the melt. In such a configuration, it is necessary to supply all the heat to perform those transformations through the surface of the glass bath and the cold cap. The power requirements are such that, in order to obtain an adequate throughput, the surface area of the melt must be extended (throughput theoretically varies proportionally to the melt surface area). Even if boosting can be provided by implementing radiative heaters in the melter plenum above the cold cap, this results in much larger melters, which cannot be heated from the outside any more. A technology directly inspired from traditional glass-making melters has been selected. The melters are lined with layers of refractory bricks in order to protect the cooled metallic walls, and heating is performed by directly applying current to the conductive melt through metallic electrodes, usually made of Inconel 690 and sometimes cooled by an internal circulation of air. The current heats the melt by Joule effect. The melt, in its turn, transfers the heat to the cold cap. Since, owing to the larger surface area, these melters hold large volumes of glass, pouring is usually performed by overflow, air-lift or vacuum siphon and can be continuous. Batch bottom pouring is nevertheless implemented for some specific applications described below. Several facilities have been operated or are still in operation worldwide. The first industrial facility to have been operated with a ceramic melter has been the PAMELA facility in Belgium, commissioned in 1985, which has processed 490 metric tonne (MT) of glass in about 6 years. The melter had a flat bottom and pouring was performed essentially by overflow. Two melters were used, with a melter life of around 3 years. The melter was approximately 2.6  2  2 m in size, weighed about 20 MTs and held about 300 L (750 kg) of glass. This facility experienced difficulties owing to the settling of glass insoluble noble metals from the waste at the bottom. This conductive settled layer tended to disturb current distribution in the bulk of the melt and led to loss of capacity and, ultimately melter failure. In order to prevent such occurrences when dealing with noble metal rich feeds, the bottom of the cavity can be designed with a ‘dead zone’ below the level of the electrodes, to collect the sludge in a manner that should prevent any interactions with power distribution to the melt; This solution was implemented at West Valley (WVDP) in the US for the vitrification of a backlog of HLW solutions from a pilot commercial reprocessing plant. This melter started operation in 1996 and produced 275 canisters of glass before being stopped and emptied in 2002. The melter, with a design throughput of about 45 kg/h of glass, was of large dimensions (3  3.2  3.3 m), with a glass hold-up of about 860 L (1150 kg) and a melt surface area of 2.2 m². The melter was equipped with two pouring chambers and a bottom electrode to promote evacuation of the noble metals. Despite these dispositions, at the end of its life, the melter showed signs of declining capacity and power distribution upsets, probably attributable to the slow accumulation of conductive noble metals at the bottom. Another solution to deal with this issue is to provide sloped walls above a bottom pouring device, to promote the evacuation of the settled noble metals. This solution has been implemented in the Tokai Vitrification Facility (TVF) (Figure 20) and K-plant in Japan, but did not completely suppress the issue, and in VEK in Germany, which has been recently decommissioned. TVF has been commissioned in 1995 and the first melter has processed 130 canisters before its replacement in 2002. Melter #2, with an improved bottom configuration, is now implemented and operating. The melter section is 0.8  0.83 m, with a glass holdup of 350 L, and a design throughput of 9 kg/h. The sloped walls at the bottom make an angle of 451 with respect to the vertical

Figure 20 The TVF ceramic melter (Japan).

Waste Glasses

27

direction. Pouring is activated by heating the pouring nozzle with an inductor. Pouring is stopped by blowing cold air to freeze the glass inside the nozzle. The electrodes are cooled with air. In Rokkasho, a much larger plant, with a design throughput of 80 kg/h and a similar conception, is in the process of active start-up. VEK in Germany was designed to process a backlog of HLW produced by a pilot reprocessing plant. The melter had a conical bottom, with slopes of around 601 with respect to the vertical direction. The melter was designed to process 10 L/h of feed, or produce 7 kg/h of glass. It was a cylindrical melter with an outside diameter of 1.5 m and a height of 1.7 m, with a glass hold-up of 150 L (375 kg), a melt surface area of 0.44 m², and an overall weight of 8 MT. The electrodes were cooled with air. Very large capacity ceramic melters have also been commissioned or are in the process of commissioning for defense high-level waste in the US. These waste compositions are not as radioactive as the waste from commercial reprocessing, and they do not contain significant amounts of noble metals. For these applications, throughput is the major concern, since the volumes of waste to be vitrified are quite impressive. The first vitrification facility commissioned in the US was the DWPF, at Savannah River, in the USA. The liquid-fed ceramic melter has been designed to process thick slurry retrieved from the SRS tank farms at a design rate of up to 100 kg/h of glass (or 200 L/h of feed). The frit is introduced as a powder and mixed with the feed suspension. This facility has been commissioned in 1996 and has produced about 7000 MTs of glass. The melter holds a melt volume of 2500 L (6.5 MT), with a melt surface area of 2.6 m². Pouring is continuous except at the time of canister change-out, and performed via a siphon. The overall weight of the melter (including the glass and cooling water) is around 80 MT. Melter # 1 was decommissioned in October 2002 and Melter # 2 started operation 5 months later, in March 2003 (Figure 21). More recently, this melter has been equipped with sparging devices to improve heat transfer and, thus, increase throughput. After completing its useful life, melter #1, together with its supporting rack, were inserted into a box, evacuated on a trailer, and entombed for long term storage in a specific underground cavity on site. For Hanford WTP, two vitrification facilities are being built. For high-level waste, the facility will host two ceramic melters with surface areas of 3.75 m² for a design throughput of 125 kg/h per melter. For low activity waste, it is intended to implement two elongated melters with melt surface areas of 10 m², for a throughput of 625 kg/h per melter. For these melters, capacity is critical. In order to improve capacity, and enhance heat transfer from the melt to the cold cap, extensive air sparging is being implemented. Liquid-fed ceramic melters are also used in Russia to process high-level waste into a phosphate based matrix. The selection of the phosphate based matrix allowed reducing melting temperature to below 1000 1C, but this matrix is still very corrosive and detrimental to the melter lifetime. The interest of liquid feeding is thus the fact that one single piece of equipment is needed to perform all the necessary reactions. Ceramic melters are also quite stable in operation. The major limitations of these melters are the following:

• • •

They are large pieces of equipment which, once used, become large waste. Melter exchange can require several months. With some exceptions, once started, they must be maintained hot until their decommissioning, in order to avoid deteriorating the ceramics. Their tolerance to glass insoluble elements, such as noble metals, or crystal forming elements, is limited.

Since the molten glass is in direct contact with the refractories and electrodes, operating temperature is limited, usually around 1150 1C, sometimes a little higher, in order to preserve the integrity of the refractories and electrode material. This is a limiting factor for capacity and for the type of glass that can be processed, as for the hot metallic melters described above.

5.2 5.2.1

Emerging Processes for Radioactive Waste Vitrification Cold crucible induction melters

As it has been seen above, the two major current processes for radioactive waste vitrification, although they have proven their sturdiness and adaptation to very demanding environments, have reached their limits, in terms of throughput, tolerance to some elements in the glass or glass composition. Other solutions have then been sought in the last two decades, with the emergence of the cold crucible induction melter (CCIM) technology both in France and Russia. For this technology, the crucible itself is composed of a water-cooled metallic structure. The heating mode is direct induction in the molten bath, a technique that allows transmitting power directly at the heart of the glass, using a water-cooled inductor (Figure 22). In order to allow the penetration of the magnetic field provided by the inductor past the metallic crucible (to avoid the Faraday cage effect), the crucible is made of several sectors separated by a thin layer of isolating material. The power (typically around 600 kW) is directly induced in the glass by the high-frequency magnetic field (typically 300 kHz in France). Since cold glass is not conductive, melting is started by introducing a conductive material (for instance graphite) on the cold glass in order to generate the first molten zone. This molten zone then progressively extends to fill the whole volume of the melter. A thin layer of molten material freezes at the zone of contact with the cold metallic wall. This frozen layer, several mm thick, forms a ‘cold crucible’ which prevents further contact between the molten mass and the metallic wall.

28

Waste Glasses

Figure 21 The DWPF melter (USA).

This technology offers a number of advantages over the previous processes:

• • • • • •

Since the molten glass does not contact the melter wall, temperature is not limited any more, offering a whole new range of possibilities for glass or even ceramics formulations. The cold layer also protects the equipment from corrosion, and thus allows processing effluents that were deemed too corrosive for the previous techniques (such as effluents containing sulfur, chlorine, molten salts, etc.). The molten glass is not polluted by any material from the crucible. Significant increase in capacity associated with the possibility of raising the operating temperature. Since the crucible is protected from both heat and corrosion, its lifetime is extended. The cold crucible layer can be easily removed at the end of operation, thus allowing easier decontamination of the crucible.

Waste Glasses

29

Figure 22 The cold crucible induction melter (CCIM).

Figure 23 CCIM in active cell at La Hague.

The use of CCIM in industrial facilities has started. The cold crucible technology is used in Russia to vitrify low activity effluents from nuclear power plants. In S. Korea, an incineration-vitrification facility based on the French CCIM technology has been commissioned a few years ago to process dry active waste and resins from nuclear power plants. And in France, for High-Level radioactive Waste, a first CCIM has been retrofitted by remote operation in one of the lines of the R7 facility to process some specific effluents,57 difficult to vitrify with a hot crucible. The industrial commissioning of a CCIM in an existing vitrification line in the R7 facility (Figure 23) on April 2010 was the successful outcome of an innovative project, conducted in close collaboration between the CEA and AREVA. This deployment is the apex of more than 20 years of R&D work and has shifted the status of this technology to that of fully mature technology.58 In parallel with the implementation of the CCIM at la Hague it was necessary to develop and qualify glass matrices adapted to the melter and tailored for the various waste considered for processing. For instance:

• •

Rinsing effluents from the decommissioning of some older facilities. The first stream to be treated industrially was decommissioning effluents from the UP2–400 facility. At the end of 2012, 190 canisters of this type had been produced with the CCIM. Legacy waste rich in molybdenum. This molybdenum-rich waste could not be processed in metallic melters owing to its corrosiveness and to the low solubility of molybdenum in traditional borosilicate glass. For this waste, a glass-ceramic matrix was designed, which is molten at temperatures that cannot be reached in a metallic melter. During 2013, 28 CSD-U, containers filled with UMo glass, were produced in a test campaign.

These first industrial campaigns allowed to validate the good behavior of the process in active conditions, to demonstrate throughputs in accordance with inactive tests, and to demonstrate product compliance with the expected characteristics.

5.2.2

Incineration/vitrification processes

Alternative processes of incineration-vitrification coupling CCM and plasma torches are developed with the aim of drastically reducing ILW volume while confining them in a more durable glassy matrix. The proposed technology relies on the large skill developed by the CEA’s vitrification teams on the cold crucible melter but also on the oxygen transferred mode plasma torches that have been tested for more than 10 years in Marcoule.

30

Waste Glasses

Waste + Glass precursor Oxygen Oxygen

Cathode

Anode

Burned gases exhaust Metallic Cooled walls Plasma Molten glass HF Current

Inductor

Figure 24 The SHIVA process principle.

The SHIVAvii process (Figure 24) has been developed to study the feasibility of these technologies for the future.59 It allows performing in the same vessel the incineration of the burnable wastes, the vitrification of their mineral charge and the combustion of off-gasses. Significant advantages can be obtained by supplying the waste directly into the oxygen plasma arc located above a glass bath heated by direct induction in a cold crucible. The temperature is very high and so is the efficiency of the combustion in the excited oxygen-rich atmosphere that also promotes a good oxidation of the glass. The treatment of various waste types has been investigated. Burnable wastes such as ionic exchange resins, bituminous wastes, sulfate slurries, graphitic sludge, have been successfully incinerated with a good incorporation of their mineral charge into the glass.60 In addition, mineral wastes such as sludge issuing from nuclear treatment have been treated too. The current studies focus on the incineration-vitrification of chlorinated organic wastes. In this case, the main difficulty is to manage the volatile metallic chlorides in the process.

6

Conclusions and Outlook on Waste Glasses

Today vitrification is the world reference solution to the containment of HLW. Since the first trials of vitrification of FPs solutions in Saclay (France) in the 1950s, until the last development of a new generation of cold crucible melter in the 10’, vitrification is a success story which allowed having waste glasses that meet all industrial requirements while providing an excellent long-term behavior. Thus, in all reprocessing country, vitrification plays a vital role in our ability to safely manage the high-level waste from nuclear energy. Given the importance of this matrix for high-level waste management a huge number of researches devoted to nuclear glasses have been supported over the last thirty years either by nuclear industry or by governmental organization. These studies have focused on both optimizing complex glass properties (solubility of components, structure, durability, etc.) understanding the containment properties and the long term behavior, and on the continuous improvement of vitrification processes operating in hostile environments. Through these studies, general knowledge of complex glasses has considerably progressed, particularly in the area of understanding the mechanisms of glass alteration by water. To date there are probably many more publications on the alteration of the R7T7 glass than on the alteration of window glass. Many progresses were done on radiation damage too. If thirty years ago there were fears that the glasses could fall into ruin under the effect of self-irradiation, today it is known that glass is a self-repairing materials whose macroscopic properties are not affected by long term auto-irradiation. Great progress has also been made in atomistic modeling of complex glasses that allow checking that our models are based on an atomistic understanding of basic phenomena. This success of vitrification is expected to further increase in the next thirty years, for at least two reasons. First, in a context of global nuclear renaissance with the need to conserve resources, more and more countries will choose nuclear fuel recycling and vitrification will prevail for waste treatment. On the other hand, with the increasing desire to protect the environment, a large number of matrices used today to confine intermediate nuclear waste and even hazardous waste, will be probably replaced by glass, as incineration/vitrification allows both a drastic waste volume reduction and a final containment with improved performance. vii

French acronym for Systeme Hybride d’Incineration Vitrification Avancé.

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With the proliferation of waste types treated, we should expect a proliferation of glass compositions too. One should have to be very careful about the quality of glass products to never degrade the image of this excellent containment matrix. In fact the glass is a wonderful material that can pass almost continuously from a soluble borax glass to almost eternal obsidian. Care should be taken that the great glass flexibility is not used at the expense of its quality. The development of good containment glass requires mastering glassy materials, process aspects, all together with the science of their long-term behavior.

References 1. Bonniaud, R., Cohen, Y., Sombret, C., 1958. Essais d'incorporation des solutions concentrées de produits de fission dans les verres et les micas. Deuxième Conférence Internationale des Nations Unies sur l'utilisation de l'Energie Atomique à des fins pacifiques. Genève. 1−9−1958. 2. Mendel, J.L., Ross, W.A., Roberts, F.P., et al., 1977. Annual Report on the Characteristics of High-Level Waste Glass BNWL-2252. Richland, WA: Pacific Northwest Laboratory. 99352. 3. Zarzycki, J., 1982. Les verres et l'état vitreux: Masson. Paris. 4. Jo, P., 2001. Matériaux non Cristallins et Science du Désordre. Presses polytechniques et universitaires Romandes. 5. Réaumur, M., 1739. Mém Acad Sci. 370. 6. MacDowell, John F., 1966. Nucleation in glasses. Industrial and Engineering Chemistry 58 (3), 38–45. 7. Vernaz, E., 1990. Physico-chemical properties and log-term behaviour of French R7T7 nuclear waste glass. Xth National Scientif Technical Conference with international participation on Glass and fine Ceramics, Varna, Bulgaria, October 18−20. 8. Orhlac, X., 2000. Étude de la stabilité thermique du verre nucléaire. Modélisation de son évolution à long terme. PhD Thesis, Montpellier II University, CEA Report CEA-R5895. 9. CEA-DEN Monographs, 2009. Nuclear Waste Conditioning. Éditions du Moniteur, Paris, ISBN 978−2−281−11380−8 (Topic Editor: Etienne Vernaz). 10. Orlhac, X., Fillet, C., Sempere, R., Phalippou, J., 2001. Journal of Non Crystalline Solids 291, 1–13. 11. Vernaz E., N. Jacquet Francillon, R. Bonniaud, 1982. Echos du CEA 38−43. 12. Matzke, H.J., Vernaz, E., 1993. Journal of Nuclear Materials 201, 295–309. 13. Peuget, S., Cachia, J.N., Jégou, C., et al., 2006. Journal of Nuclear Materials 354, 1. 14. Marples, J.A.C., 1988. Nuclear Instruments and Methods in Physics Research B 32, 480. 15. Weber, W.J., Ewing, R.C., Angell, C.A., et al., 1997. Journal of Materials Research 12, 1946. 16. Inagali, Y., Furuya, H., Idemitsu, K., et al., 1992. Materials Research Society Symposium Proceedings 257, 199. 17. Peuget, S., Delaye, J.-M., Jégou, C., 2014. Specific outcomes of the research on the radiation stability of the French nuclear glass towards alpha decay accumulation. Journal of Nuclear Materials 444 (1−3), 79–91. 18. Peuget, S., Maugeri, E.A., Charpentier, T., et al., 2013. Comparison of radiation and quenching rate effects on the structure of a sodium borosilicate glass. Journal of Non-Crystalline Solids 10/2013 (378), 201–212. 19. Delaye, J.-M., Ghaleb, D., 2005. Physical Review B 71, 224204. 20. Vernaz, E., Gin, S., Jegou, C., Ribet, I., 2001. Present understanding of R7T7 glass alteration kinetics and their impact on long-term behavior modeling. Journal of Nuclear Materials 298 (1,2), 27–36. 21. Frugier, P., et al., 2008. SON68 nuclear glass dissolution kinetics: Current state of knowledge and basis of the new GRAAL model. Journal of Nuclear Materials 380 (1−3), 8–21. 22. Jollivet, P., et al., 2012. Chemical Geology 330, 207–217. 23. Vernaz, E., Dussossoy, J.L., 1992. Current state of knowledge of nuclear waste glass corrosion mechanisms: The case of R7T7 glass. Applied Geochemistry 7 (1), 13–22. 24. Gin, S., Guittonneau, C., Godon, N., et al., 2011. The Journal of Physical Chemistry C 115, 18696–18706. 25. Gin, S., Frugier, P., Jollivet, P., et al., 2013. International Journal of Applied Glass Science 4, 371–382. 26. Ribet, S., Gin, S., 2004. Role of neoformed phases on the mechanisms controlling the resumption of SON68 glass alteration in alkaline media. Journal of Nuclear Materials 324, 152. 27. Jollivet, P., Angeli, F., Cailleteau, C., et al., 2008. Investigation of gel porosity clogging during glass leaching. Journal of Non-Crystalline Solids 354 (45−46), 4952–4958. 28. Cailleteau, C., Angeli, F., Devreux, F., et al., 2008. Insight into silicate glass aqueous alteration mechanisms. Nature Materials 7, 978–983. 29. Gin, S., Jollivet, P., Fournier, M., et al., 2015. Nature Communications. 7360. 30. Sales, B.C., Boatner, L.A., 1984. Lead phosphate glass as a stable medium for the immobilization and disposal of high-level nuclear waste. Materials Letters 2 (4), 301–304. 31. Frugier, P., Martin, C., Ribet, I., Advocat, T., Gin, S., 2005. Journal of Nuclear Materials 346, 194–207. 32. Ribet, S., Muller, I., Pegg, I., Gin, S., Frugier, P., 2004. Materials Research Society Symposium Proceedings 824, 309–314. 33. Chave, T., Frugier, P., Gin, S., Ayral A., 2011. Geochimica et Cosmochimica Acta 75 (15), 4125–4139. 34. Gin, S., Godon, N., Mestre, J.P., Vernaz, E., Beaufort, D., 1994. Applied Geochemistry 9, 255–269. 35. Michelin, A., et al., 2013. Environmental Science & Technology 47, 750–756. 36. Jollivet, P., et al., 2012. Journal of Nuclear Materials 420, 508–518. 37. Pozo, C., Bildstein, O., Raynal, J., Jullien, M., Valke, E., 2007. Journal of Nuclear Materials 35, 258–267. 38. Pederson, L.R., et al., 1984. Scientific basis for nuclear waste management. Materials Research Society Symposium Proceedings 26, 417–426. 39. Grambow, B., Muller, R., 1990. Scientific basis for nuclear waste management. Materials Research Society Symposium Proceedings 176, 229–240. 40. Lutze, W., Muller, R., Montserrat, W., 1987. Scientific basis for nuclear waste management. Materials Research Society Symposium Proceedings 112, 572–584. 41. Feng, X., Pegg, I., 1994. Physics and Chemistry of Glasses 35, 98–103. 42. Minet, Y., Godon, N., 2003. In environmental issues and waste management technologies in the ceramic and indutries VIII. Book Series. Ceramic Transactions 143, 275–282. 43. Geneste, G., Bouyer, F., Gin, S., 2006. Journal of Non-Crystalline Solids 352, 3147–3152. 44. Ribet, I., Betremieux, S., Gin, S., Angeli, F., Jégou, C., 2009. In Global Conference Proceedings. Paris, September 6−11. 45. Devreux, F., Barboux, P., Filoche, M., Sapoval, B., 2001. Journal of Materials Science 36, 1331–1341. 46. Ledieu, A., Devreux, F., Barboux, P., Minet, Y., 2006. Nuclear Science and Engineering 153, 285–300. 47. Grambow, B., Muller, R., 2001. First-order dissolution rate law and the role of surface layers in glass performance assessment. Journal of Nuclear Materials 298, 112–124. 48. Bourcier, W.L., Pfeiffer, D.W., Knauss, K.G., McKeegan, K.D., Smith, D.K., 1990. A kinetic model for borosilicate glass dissolution affinity of a surface alteration layer, scientific basis for nuclear waste management. In: Materials Research Society Symposium Proceedings 176, 209–216. 49. Oelkers, E.H., 2001. General kinetic description of multioxide silicate mineral and glass dissolution. Geochimica Cosmochimica Acta 65, 3703–3719.

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Waste Glasses

50. Frugier, P., Chave, T., Gin, S., Lartigue, J.E., 2009. Application of the GRAAL model to leaching experiments with SON68 Nuclear glass in initially pure water. Journal of Nuclear Materials 392, 552–567. 51. Vernaz, E., Poinssot, C., 2008. Scientific Basis for Nuclear Waste Management XXXI. In: MRS Symposium Proceeding Vol 1107, Overview of the CEA French research program on nuclear waste. 52. Verney-Carron, A., Gin, S., Libourel, G., 2008. A fractured Roman glass block altered 1800 years in seawater: Analogy with nuclear waste glass in deep geological repository. Geochimica et Cosmochimica Acta 72, 5372–5385. 53. Verney-Carron, A., Gin, S., Frugier, P., Libourel, G., 2010. Geochimica and Cosmochimica Acta 74, 2291–2315. 54. Techer, I., Lancelot, J., Clauer, N., Michel Liotard, J., 2001. Alteration of a basaltic glass in an argillaceous medium: The Salagou dike of the Lodève Permian Basin (France). Analogy with an underground nuclear waste repository. Thierry Advocat Geochimica et Cosmochimica Acta 65 (7), 1071–1086. 55. Parruzot, B., Jollivet, P., Rébiscoul, D., Gin, S., 2015. Geochimica et Cosmochimica Acta 154, 28–48 56. Vernaz, Étienne, Bruezière, Jérôme, 2014. History of nuclear waste glass in France. Procedia Materials Science 7, 1–294. In: 2nd International Summer School on Nuclear Glass Wasteform: Structure, Properties and Long-Term Behavior, SumGLASS 2013. 57. Naline, S., Girold, C., et al., 2010. Vitrification: A challenging French vitrification project to retrofit a cold crucible inductive melter at the la Hague plant. In: Waste Management Symposium (WM - 2010), 07/03/2010−11/03/2010; Phoenix, AZ. 58. Ladirat, C., Hollebecque, J., Chauvin, E., 2015. The Industrial Experience Feedback and R&D Ensure the Longevity of French Vitrification Facilities, WM2015 Conference, March 15−19, Phoenix, AZ. 59. Lemont, F., et al., 2010. An innovative hybrid process involving plasma in a cold crucible melter devoted to the future intermediate level waste treatment: The SHIVA technology - 12th International Conference on Modern Materials and Technologies (CIMTEC - 2010), 06/06/2010−11/06/2010, Montecatini Terme, Italie. 60. Girold C., Lemort F., Pinet O., 2006. The vitrification as pathway for longlife organic waste treatment. Waste Management Conference, 26/02/2006−28/02/2006, Tucson, Etats-unis (Waste Management Symposium proceedings).