A method is suggested to assess the tolerable salt content of the evaporator bottoms from the data on solubility in salt systems taken as simplified models of liquid radioactive waste (LRW) arising from nuclear power plants (NPP) with boiling reactors. It has been demonstrated that the degree of evaporation may be substantially increased by implementing the process in nitric acid. Equations have been derived that allow the calculation of the minimum needed acidity of the solution to allow maximum evaporation.

A spreadsheet model, PRETREAT, was developed for mass-balance calculations of pretreatment processes proposed for the immobilization of Hanford nuclear wastes. In PRETREAT, a graphical user interface is implemented for convenient navigation through the large number of worksheets used in the model, for data input, modification of process parameters, and viewing model results. PRETREAT can be used to estimate radionuclide and chemical species concentrations in the waste streams at each stage of the pretreatment flowsheet, and to help in assessments of potential radiological, chemical, and criticality hazards and in identification of mitigation strategies. The model can also be used to evaluate the sensitivity of model results to uncertainties in model parameters such as decontamination or separation factors.

A thermochemical approach was suggested for treating and conditioning specific streams of radioactive wastes for example spent ion exchange resins (IER), mixed, organic or chlorine-containing radioactive waste (PVC, etc.). Conventional thermal treatment of such waste encounters serious problems concerning complete destruction of organic molecules and possible emissions of radionuclides, heavy metals and chemically hazardous species. The thermochemical treatment uses powdered metal fuels (PMF) that are specifically formulated for the waste composition and react chemically with the waste components. The composition of the PMF is designed in such a way as to minimize the release of hazardous components and radionuclides in the off gas and to confine the contaminants in the ash residue. The thermochemical procedures allow decomposition of organic matter and capturing hazardous radionuclides and chemical species simultaneously. Previous preliminary work demonstrated the feasibility of applying the thermochemical approach to treatment of spent IER using PMF. Herein, the results are presented of theoretical and experimental studies to define the optimal PMF composition as well as the PMF/waste ratio.

At the present time the primary problem in a closed nuclear fuel cycle is the management of high level liquid waste (HLLW) generated by the recovery of uranium and plutonium from spent nuclear fuel. Long-term storage of the HLLW, even in special storage facilities, poses a real threat of ecological accidents. This problem can be solved by incorporating the radioactive waste into solid fixed forms that minimize the potential for biosphere pollution by long-lived radionuclides and ensure ecologically acceptable safe storage, transportation, and disposal. In the present report, the advantages of a two-stage HLLW solidification process using a “cold” crucible induction melter (CCIM) are considered in comparison with a one-stage vitrification process in a ceramic melter.

This paper describes the features of a process and equipment for a two-stage HLLW solidification technology using a “cold” crucible induction melter (CCIM) and identifies the advantages compared to a one-stage ceramic melter. A two-stage pilot facility and the technical characteristics of the equipment are described using a once-through evaporator and cold-crucible induction melter currently operational at the IA.Mayak. facility in Ozersk, Russia. The results of pilot-plant tests with simulated HLLW to produce a phosphate glass are described. Features of the new mineral-like waste form matrices synthesized by the CCIM method are also described. Subject to further development, the CCIM technology is planned to be used to solidify all accumulated HLLW at Mayak – first to produce borosilicate glass waste forms and then mineral-like waste forms.

This paper describes the results of a joint research program of Russian institutes at St. Petersburg, Krasnoyarsk and Zheleznogorsk and the Idaho National Engineering and Environmental Laboratory to develop a new material for stabilizing radioactive solutions and other uses, such as high-temperature catalysis. An open-cell glass crystalline porous material, Gubka, or “sponge” in Russian, having an open-cell porosity of up to 90 %, was produced from hollow glass crystalline microspheres (cenospheres) formed in fly ash from coal combustion. The cenospheres were separated into fractions based on grain size, density, magnetic properties, and whether or not they were perforated. Selected fractions were molded and agglomerated by sintering with or without a binder at high temperatures. Depending on the cenosphere fractions selected, sintering conditions and additional treatments, Gubka was formed with an open-cell porosity ranging from 40-90 %. The porous material has a bulk density of 0.3-0.6 g/cm3, and two types of porous openings: 0.1-30 micrometer flow-through pores in the cenosphere walls and 20-100 micrometer interglobular pores between the cenospheres. Examples of Gubka application described in this paper include stabilization of different surrogate radioactive waste solutions containing 0.0001 to 0.7 M nitric/hydrochloric acid and 0 to 1.2 M sodium nitrate. Waste solid loadings of 46-55 wt.% nitrate salts, or 26-37 wt.% oxides after calcination, were achieved in those tests.

Leaching of the iodide ion from BiPbO2I (BPI), BPI encapsulated in cement (BPIC) and AgI was studied in a low salinity solution and in a high salinity solution under reducing conditions. Although BPI released a limited amount of iodide ions (less than 1%) into the low salinity solution, it released more than 30% of iodide ions into the high salinity solution within 80 days. AgI released more than 30% of iodide ions into both low and high salinity solutions within 80 days. It was proved that BPI is more stable than AgI in the low salinity solution under reducing conditions. BPIC released a limited number of iodide ions (less than 5%) into both low and high salinity solutions. BPIC showed the best leach resistance in the high salinity solution.

BiPbO2NO3 (BPN) was developed to remove iodide ions in a solution and fix them in BPI by the ion exchange reaction. Ion exchange properties under reducing conditions were studied. An anion exchange capacity of 1.0 mEq/g and a distribution coefficient of larger than 0.1 m3/kg were obtained in a solution at a pH of between 9 and 13. The advantages of the process using BPN for removing and immobilizing Iodine-129 were discussed from the standpoint of process simplification.

Leachabilities and solubilities of the synthesized iodide sodalite and natural (chloride) sodalite were measured by leach test. The iodide sodalite was synthesized in nitrogen gas flow at 800°C for 2 hours. The crystalline structure of the product was certified by XRD analysis. The natural sodalite containing chlorine was obtained at Bancroft, Canada. The several types of solution were used to evaluate the influence of the solubility of sodalite that included the varied pH and the chemical compositions. The solubilities of chloride sodalite were calculated by the thermodynamics data. The solubility measured for the synthesized iodide sodalite was compared with that calculated. The solubility of the synthesized iodide sodalite was approximately 2 x 10-4 mol/L, which shows a good agreement with the calculated one of the natural sodalite.

The operation and, particularly, the decommissioning of NPPs and radiochemical plants result in substantial arisings of radioactive metal waste (RAMW) having different activity levels (from 5×10−4 to ≍ 40 Ci/kg).

This paper reviews the specific features of the technology and equipment used to melt RAMW in electric arc and induction furnaces with ceramic or cold crucibles. The experimentally determined and calculated data are given on the level to which RAMW is decontaminated from the main radionuclides as well as on the distribution of the latter in the products of melting (ingot, slag, gaseous phase).

Special attention is focused on the process and the facility for the induction-slag melting of RAMW in furnaces equipped with cold crucibles. The work described is under way at SSC RF VNIINM to master the technology of melting simulated high activity level Zr-alloy and stainless steel waste.

A thermochemical approach to conditioning ash residue that involves the use of exothermic metallic compositions (EMC) has been previously reported. EMC intermixed with ash residue at the appropriate ratios can sustain a glass forming reaction wave that produces monolith glass-like blocks. Herein, the thermochemical conditioning process is extended for conditioning contaminated clay soils. The results are reported for a study of the conditioning process and product materials to determine the optimal processing parameters and to characterize the product materials.

After various thermal treatments (up to 450°C), residual thermo-hydro-mechanical (T-H-M) properties of two OPC high performance concretes (HPC) were analyzed in the context of surface long-term storage. Materials were prepared with silico-calcareous aggregates (standard HPC) and hematite aggregates (heavy HPC). The initial microstructural (porosity ≈10%) and transport (gas permeability ≈10-19 m2) properties are similar for both concretes. As far as the mechanical aspect is concerned, heavy HPC shows a higher compressive strength and elastic modulus than standard HPC (78 and 63 MPa, 81 and 49 GPa, respectively). Heavy HPC is also characterized by a higher thermal conductivity (7.3 W m-1 K-1 compared to 2.7 W m-1 K-1 for standard concrete). Results analysis show that thermo-hydro-mechanical damages are smaller for heavy HPC. Between 60 and 250°C, the elastic modulus and the compressive strength of standard HPC decrease by 40% and 16%, respectively. For heavy HPC, these parameters respectively decrease by 10% and 4%. A similar trend was observed for thermal conductivity evolution. Gas permeability and porosity data confirm the good behavior of heavy HPC. As a conclusion, hematite HPC seems to provide more interesting T-H-M residual properties than standard HPC. Limited thermal expansion and thermal gradients induced by hematite are probably responsible of this behavior.

Long-term behavior of cemented waste packages requires the knowledge of water diffusion coefficients (water is supposed to be the main degradation agent, by solubilization of some chemical compounds). A simple modeling, based on the use of hydration code CEMHYD3D is proposed and validated for pure OPC pastes (sound or degraded) at mass ratio water / cement lying between 0.25 and 0.65. The nature of the porosity available for diffusion and the role of residual anhydrous compounds are discussed.

Solid-state 133Cs magic angle spinning nuclear magnetic resonance (MAS NMR) measurements have been used to investigate Cs speciation in cements designed for immobilizing low-level nuclear wastes. Cs in ordinary Portland cement cured for only 10 days at room temperature appears to only inhabit aqueous solutions, as evidenced by a sharp signal with no sidebands, whether or not substantial quantities of blast furnace slag or silica fume are present. No significant 133Cs spectral differences were observed when slag or silica fume were present for curing periods of 10 days, but differences due to Cs partly inhabiting crystalline sites instead of water solutions were observed after 12 months in the slag-containing cement. Pre- incorporation of Cs in various zeolites, followed by mixing the zeolites into the cement, leads to enhanced Cs retention in the cement when it is subject to water leaching at 25°C.

Low-level radioactive waste (LLRW) is disposed of by shallow land burial. Reinforced concrete is used as the radioactive waste repository. However, the concrete structure is in contact with water and will gradually degrade over an extended period of time due to leaching.

It is important to investigate the interaction between radionuclides and degraded concrete in the safety assessment of nuclear waste disposal. The authors measured the distribution coefficients (Kd) of various radionuclides for calcium-leached mortars. The calcium-leached mortars were prepared by an accelerated leaching test for mortar based on the electrical potential gradient. These degraded conditions are similar to that of degraded concrete in contact with water for a long period of time. The degradation degree of calcium-leached mortar is evaluated by the CaO/SiO2 molarratio (Ca/Si ratio) of calcium silicate hydrate (C-S-H).

As a result, the relationship between Kd and the Ca/Si ratio in C-S-H can be roughly grouped into the following three types:

1.137Cs and 85Sr – Kd decreases with an increase in the Ca/Si ratio.

2. 95mTc and 110mAg – There is no correlation between Kd and the Ca/Si ratio.

3. 14C, 241Am and 125I – Kd increases with an increase in the Ca/Si ratio.

As Kd from radwaste elements in concrete systems show a wide range of values, a modelling of cesium and lead immobilization in Calcium Silicate Hydrate (CSH, xCaO.SiO2.H2O, with 0.7 (aged cements)<Ca/Si<1.8 (fresh cements)) matrix has been developed. This model accounts for radwaste surface binding mechanisms on silanol sites (>SiOH) and precipitation equilibria. Values of Kd from cesium and lead in CSH matrix can thus be calculated with Ca/Si evolution and ionic strength effect. Predictive calculations have been carried out with success with different Ca/Si ratios, ionic strengths and liquid/solid ratios, and results are well superimposed with experimental isotherm data. If the CSH really allow accounting for the radwaste behavior in hydrated cement matrices, this model can be used in safety assessment calculations, with varying pH and [Ca2+] as cement degradation state parameters.

A categorization of cement components was attempted from the viewpoints of their chemical constituents and sorption behaviors. In this study, we assumed that the cement components are categorized into four groups, calcium silicate compounds, hydroxides, calcium aluminate compounds, and others. The sorption behaviors were compared among the four groups by batch sorption experiments carried out for 12 kinds of cement components using the radionuclides cesium, iodine, inorganic carbon, and nickel as different chemical species. The characteristic sorption trends were obtained in each group. We found that the sorption mechanism is closely related to the chemical constituents and surface species. Categorization of these cement of components will contribute to the understanding of sorption mechanisms.

The extent and the consequences of interactions between cementitious materials used in radioactive waste management and clay host rock are described. In-situ tests were performed on seven cement formulations representing materials applied in repository construction, for backfilling or for solidification of radioactive waste. Samples were exposed to realistic repository conditions of the Boom Clay Formation in the HADES underground laboratory. Chemical, physical and mineralogical changes across the cement-clay interface were identified by combined observations from Electron Probe Microanalysis, Infrared microscopy and X-Ray powder diffraction. Significant interactions in both the cement and the clay part were found in a zone extending up to several hundreds of microns. The most prominent features are (1) leaching of cement with loss of calcium and/or silicon; (2) development of a calcium-rich zone in Boom Clay close to or at contact; (3) the formation of a contact zone marked by the precipitation of a (hydrated) magnesium aluminate phase; (4) reduction in apparent porosity of initially porous/permeable materials and (5) precipitation of calcite within the cement. This elemental exchange tends to diminish pH and reduce the buffering capacity of the cement. Although hydroxide will diffuse into the clay, the development of an extensive alkaline halo in the surrounding clay is unlikely owing to the buffering capacity of the Boom Clay pore water.

During interim storage, the main evolution factor of the bituminized wastes is radiolysis under auto-irradiation, while in deep disposal conditions, after radioactive decrease, site resaturation with water, and container corrosion, the main factor of evolution is leaching by water.

This paper presents two models, designed to describe respectively radiolytic gas evacuation, and bituminized waste alteration under leaching.

The first model is based on the integration of simple descriptions of gas production by radiolysis, gas diffusive transport, bubble nucleation and ripening, and bubble transport under buoyancy and viscosity.

The second model is based on simple descriptions of water adsorption and solubilization at the surface, water diffusion into the matrix, water uptake by the most soluble salts, salt grain swelling and alteration of matrix waterproofing, diffusive transport of salts out of the waste.

The potential effect of water-soluble organic ligands, originating from the radiolytic degradation of Eurobitum bituminized waste, on the Pu and Am solubility and sorption in Boom Clay, was assessed. Granulated inactive Eurobitum simulate in contact with representative solutions was gamma-irradiated, and the irradiated solutions were thoroughly analyzed (GC-MS, IC, ICP-AES,…). Pu and Am solubility and sorption tests were then performed in pure Real Boom Clay Water (RCW) and in RCW containing 50 mass% of the solution containing the radiolytic degradation products. The relevance of the applied methodology is discussed.

The thermally stimulated current (TSC) and dc conductivity for iron phosphate glasses containing up to 28 mol% Cs2O have been measured in a temperature range from 120 to 400 K. The dc conductivity and activation energy were constant and independent of Cs2O content. With increasing cesium concentration in cesium iron phosphate glasses the slowly moving cesium ions are more tightly bound to the non-bridging oxygen ions and make no measurable contribution to dc conductivity. The dc conduction in these glasses is totally electronic, controlled by electron hopping between iron ions. The ionic conduction is immeasurably small because of the low mobility of the cesium ions. This agreement is reinforced by the excellent chemical durability of the glasses, where the dissolution rate at 90oC changes little with increasing Cs2O content. Raman spectroscopy indicated that the structure of these glasses was composed of predominantly pyrophosphate (P2O7) groups, but the metaphosphate chains (PO3) also existed.

The fission products resulting from reprocessing of commercial spent fuel are currently vitrified industrially by COGEMA at La Hague. The properties of 21 non-radioactive borosilicate glass samples containing between 4 and 6% of the platinum-group metals (PGM: Pd and Ru) compared with 1.6% in the industrial glass were investigated for chemical composition variations covering the full specification range. After a brief morphological description of the undissolved PGM in the glass, the viscosity variations at temperatures ranging from 1100 to 1200°C are discussed with emphasis on the effects of the particle inclusions on the rheological properties of the glass. Variations in the chemical durability of quenched glass specimens are then discussed. The initial leach rate V0 at 100°C remained near the values obtained for the reference glass. The same tests were repeated on glass heat-treated to obtain maximum crystallization, and the results confirmed that the chemical durability of the glass is practically unaffected by the crystallization observed in this type of glass.