The development of novel, Radical Purex, reprocessing technologies, leading to fission product waste streams with high levels of inert constituents, may mean that vitrification is no longer the optimum solution for the immobilisation of highly active wastes at the back end of the nuclear fuel cycle. A ceramic phase assemblage is described which uses the inert constituents of the waste stream as a functional component of the waste form, permitting high waste loadings to be achieved and thus enabling waste minimisation considerations to be satisfied. The initial development of this phase assemblage is presented.

A kinetic model recently developed [1] for the radiolytically induced oxidative dissolution of the spent fuel matrix is presented. This is based on experimental studies on the generation and evolution of radiolytic products in a closed system containing fragments of PWR-fuel [2]. The outcome of this model is currently being integrated in the present PA exercise being prepared by SKB. The calibration of the model against various experimental information and its predictive capabilities for the long term performance of the spent fuel matrix are presented.

Chlorine-36 has been identified as a potential source of radiological risk in the disposal of nuclear fuel waste. The radioisotope 36Cl (t1/2 = 3 × 1O5 a) is produced by neutron activation of Cl impurities in UO2 fuel. The total average Cl impurity level in four unirradiated CANDU UO2 fuel samples was 2.3 ± 1.1 ppm. ORIGEN-S calculations using a 5 ppm Cl impurity in a CANDU fuel resulted in a 36Cl activity comparable to the activity of 129I and 14C produced in the fuel thus requiring 36Cl to be considered in disposal risk assessments. The “instant release” of 36Cl from the gap and grain boundary regions of the fuel to solution was measured by leaching both clad fuel and fuel samples crushed to grain-sized particles. The 36Cl concentration was measured by Accelerator Mass Spectrometry. The 36Cl releases from fuel samples taken from 8 different fuel bundles ranged from 0.5% to 20.4% of the total 3 Cl inventory over a leaching period of 32 days. The 36Cl released was found to correlate with the stable Xe gas release, the fuel burnup and the linear power rating (LPR). For a typical CANDU fuel with an LPR of -42 kW/m, the “instant release” of 36Cl would be about 5% of the total inventory.

A fuel leaching experiment has been in progress since 1977 to study the dissolution behaviour of used CANDU fuel in aerated aqueous solution. The experiment involves exposure of 50-mm clad segments of an outer element of a Pickering fuel bundle (burnup 610 GJ/kg U; linear and peak power ratings 53 and 58 kW/m, respectively), to deionized distilled water (DDH2O, ∼2 mg/L carbonate) and tapwater (∼50 mg/L carbonate). In 1992, it was observed that the fuel in at least one of the leaching solutions showed some signs of deterioration and, therefore, in 1993, parts of the fuel samples were sacrificed for a detailed analysis of the physical state of the fuel, using SEM and optical microscopy. Leaching results to date show that even after >6900 days only 5 to 7.7% of the total calculated inventory of 137Cs has leached out preferentially and that leach rates suggest a development towards congruent dissolution. Total amounts of 137Cs and 90Sr leached are slightly larger in tapwater than in DDH2O. SEM examinations of leached fuel surface fragments indicate that the fuel surface exposed to DDH2O is covered in a needle-like precipitate. The fuel surface exposed to tapwater shows evidence of leaching but no precipitate, likely because uranium is kept in solution by carbonate. Detailed optical and SEM microscopy examinations on fuel cross sections suggest that grain-boundary dissolution in DDH2O is not prevalent, and in tapwater appears to be limited to the outer %0.5 mm (pellet/cladding) region of the fuel. Grain boundary attack seems to be limited to microcracks at or near the surface of the fuel. It thus appears that grain-boundary attack occurs only near the fuel pellet surface and is prevalent only in the presence of carbonate in solution.

The alteration behavior of UO2 pellets following their reaction under unsaturated drip-test conditions, at 90°C, for time periods of up to 10 years has been examined by solid phase and leachate analyses. Sample reactions were characterized by preferential dissolution of grain boundaries between the original press-sintered UO2 granules comprising the samples, development of a polygonal network of open channels along the intergrain boundaries, and spallation of surface granules that had undergone severe grain boundary corrosion. The development of a dense mat of alteration phases after two years of reaction trapped loose granules, resulting in reduced rates of particulate uranium release. The paragenetic sequence of alteration phases that formed on the present samples was similar to that observed in surficial weathering zones of natural uraninite (UO2) deposits, with alkali and alkaline earth uranyl silicates representing the long-term solubility-limiting phases for uranium in both systems.

The release fractions of the five elements in the ε-phase (99Tc, 97Mo, Ru, Rh, and Pd) as well as that of 238U are reported for the reaction of two oxide fuels (ATM-103 and ATM-106) in unsaturated tests under oxidizing conditions. The 99Tc release fractions provide a lower limit for the magnitude of the spent fuel reaction. The 99Tc release fractions indicate that a surface reaction might be the rate controlling mechanism for fuel reaction under unsaturated conditions and the oxidant is possibly H2O2, a product of alpha radiolysis of water.

The dissolution rate of unirradiated UO2 (s) has been studied as a function of hydrogen carbonate concentration at three different temperatures (298.15 K, 313.15 K and 333.15 K) under oxidizing conditions in a continuous flow-through reactor with a thin layer of solid particles (particle size from 100 to 300 μm). From the results of these experiments, two different rate laws have been determined. At high temperature (313.15 K and 333.15 K), we obtained a dissolution rate proportional to hydrogen carbonate concentration while at 298.15 K, the rate almost depends on the square root of the hydrogen carbonate concentration. This indicates a different reaction mechanism depending on temperature which can be related to the oxidation step of the overall process. The apparent activation energy obtained was 41 kJ mo1−1.

Extensive Research is performed in many countries in order to evaluate the spent fuel behaviour under repository conditions. Several aspects as the control of the oxidative spent fuel dissolution by secondary phases formation are not yet clear.

Coprecipitation experiments from SIMFUEL solutions are performed to study if minor elements will influence the formation of secondary phases. Therefore, coprecipitation studies from SIMFUEL solutions aims at identification of stable phases of significant simulated fission products. These experiments provide upper limits for solution concentration and distribution ratios of simulate fission products at several pH values. SIMFUEL pellets, which simulate an irradiated fuel with burnup of 50 GWd/tU were provided by AECL Research Laboratories, Canada. Experiments were carried out by addition of an aliquot of the initial SIMFUEL solution to 5 m NaCI free of carbonates solution. The selected pH was maintained constant during the experiments. The pH range considered was from 5.5 to 9.3. Analyses of the solutions were performed for uranium by Laser fluorescence and for the minor elements by ICP-MS. Solid phases formed at pH 5.5 were dissolved and analysed by ICP-MS. Results of the evolution in solution vs. pH of simulated fission products concentrations are shown in this paper.

The solubility behavior of unirradiated UO2 pellets was studied under oxic, air-saturated conditions in deionized water, in NaHCO3 solutions and in two types of synthetic groundwater (25°C). The Allard groundwater represents natural fresh groundwater conditions at great depths in granite bedrock and bentonite groundwater simulates the effects of bentonite on granitic fresh groundwater. The release of uranium was measured during static batch dissolution experiments of long duration (6 years). A comparison was made with the theoretical solubility data calculated with the geochemical code, EQ3/6, in order to evaluate solubility (steady state) limiting factors. Various hypotheses for redox control (redox potential of the bulk solution or redox potential at the surface) were tested in the modeling calculations.

The measured concentrations for uranium at steady-state in deionized water were equal to the solubility of schoepite (PO2 = 0.2 atm). In NaHCO3 solutions with lower carbonate concentrations (0.98 – 1.96 mmol/1) and in Allard groundwater they were close to the calculated solubilities of U at the U3O7/U3O8 redox potential. In bentonite groundwater, the results suggest the formation of a secondary phase with a lower solubility. Only uranium oxide with a crystal structure of uraninite (UO2 - U3O7) was identified in all waters, when analyzing particulate material in the solutions after contact with UO2 pellets.

The purpose of this work has been to measure and model the intrinsic dissolution rates of uranium oxides under a variety of well-controlled conditions that are relevant to a geologic repository. When exposed to air at elevated temperature, spent fuel may form the stable phase U3O8. Dehydrated schoepite, UO3H2O, has been shown to exist in drip tests on spent fuel.

Equivalent sets of U3O8 and UO3H2 dissolution experiments allowed a systematic examination of the effects of temperature (25–75°C), pH (8–10) and carbonate (2–200×10−4 molar) concentrations at atmospheric oxygen conditions.

Results indicate that UO3H2O has a much higher dissolution rate (at least ten-fold) than U3O8 under the same conditions. The intrinsic dissolution rate of unirradiated U3O8 is about twice that of UO2. Dissolution of both U3O8 and UO3.H2O shows a very high sensitivity to carbonate concentration. Present results show a 25 to 50-fold increase in room-temperature UO3H2O dissolution rates between the highest and lowest carbonate concentrations.

As with the UO2 dissolution data the classical observed chemical kinetic rate law was used to model the U3O8 dissolution rate data. The pH did not have much effect on the models, in agreement with the earlier analysis of the UO2 and spent fuel dissolution data,. However, carbonate concentration, not temperature, had the strongest effect on the U3O8 dissolution rate. The U3O8 dissolution activation energy was about 6000 cal/mol, compared with 7300 and 8000 cal/mol for spent fuel and UO2 respectively.

The thermodynamic and kinetic dissolution properties of a synthetically obtained soddyite have been determined at different bicarbonate concentrations. This uranium-silicate is expected to be a secondary solid phase of the oxidative alteration pathway of uranium dioxide in waters with low phosphate content and, consequently, it is likely to constitute one of the long-term uranium solubility limiting solid phases.

The experimental data obtained at the end of the experiments correspond fairly well to the theoretical model calculated with a log K0S0 of 3.9±0.7.

On the other hand, the general trend of the total uranium in solution measured in the experiments as a function of time has been fitted by using a kinetic equation obtained from the principle of detailed balancing of the dissolution reaction. In addition, the EQ3/6 code has also been used to model the uranium concentrations as a function of time. In both modeling exercises comparable results were obtained. The dissolution rate, normalized to the total surface area used in the experiments as measured with the BET method, gave an average value of 6.8 (±4.4) 10−14 mol cm−2 s−1.

Spent nuclear fuel will, by the radiation, split nearby water into oxidizing and reducing compounds. The reducing compounds are mostly hydrogen that will diffuse away. The remaining oxidizing compounds can oxidize the uranium oxide of the fuel and make it more soluble. The oxidised uranium will dissolve and diffuse away. The nuclides previously incorporated in the spent fuel matrix can then be released and also migrate away from the fuel.

A model is proposed where the produced oxidizing species compete for reaction with the fuel and for escaping out of the system. The chemical reaction rate of oxygen and fuel is taken from literature values based on experiments. The escape rate of oxidants to a receding redox front in the backfill is modelled assuming a redox reaction of oxidizing component and reducing component in the surrounding. The rate of movement of the redox front is determined from the rate of production of oxidants. This is estimated using a previously devised model that has been calibrated to in situ observed radiolysis.

Three cases are modelled. In the first case it is assumed that the reducing compound is insoluble and that the reaction between oxygen and reducing mineral is very fast. In the second case it is assumed that the reducing component has a known solubility and that it can migrate to meet the oxygen and quickly react. In a third case a finite reaction rate is modelled between the oxygen and the reducing species.

The sample calculations show that if the reducing mineral has to be supplied from the backfill a large fraction of the spent fuel could be oxidised. If the corrosion products of a degraded steel canister can supply the reducing species and the redox reaction is fast, very small amounts of the fuel could be oxidised. Literature data indicate that the redox reaction rate may not be so fast that it can be considered instantaneous and then a considerable fraction of the fuel could be oxidised. The model gives a means of exploring which mechanisms and data may be of most importance for radiolytic fuel dissolution, but the realism of the data and the model must be tested further. There is a lack of understanding and data on reaction rates, heterogeneous as well as homogeneous. This is crucial to the results.

Spent nuclear fuel (SNF) is unstable under oxidizing conditions. Although recent studies have determined the paragenetic sequence for uranium phases that result from the corrosion of SNF, there are only limited data on the potential of alteration phases for the incorporation of transuranium elements. The crystal chemical characteristics of transuranic elements (TUE) are to a certain extent similar to uranium; thus TUE incorporation into the sheets of uranyl oxide hydrate structures can be assessed by examination of the structural details of the β-U3O8 sheet type.

The sheets of uranyl polyhedra observed in the crystal structure of β-U3O8 also occur in the mineral billietite (Ba[(UO2)3O2(OH)3]2(H2O)4), where they alternate with α-U3O8 type sheets. Preliminary crystal structure determinations for the minerals ianthinite, ([U24+(HO2)4O6(HO)4(H2O)4](H2O)5), and “wyartite II” (mineral name not approved by IMA committee on mineral names), {CaCo3}[U4+(UO2)2O3(OH)2](H2O)4, indicate that these phases also contain β-U3O8 type sheets. The β-U3O8sheet anion topology contains triangular, rhombic, and pentagonal sites in the proportions 2: 1:2. In all structures containing β-U3O8 type sheets, the triangular sites are vacant. The pentagonal sites are filled with U6+O2 forming pentagonal bipyramids. The rhombic dipyramids filling the rhombic sites contain U6+O2 in billietite, U4+O2 in β-U3O8U4+(H2O)2 in ianthinite, and U4+O3 in “wyartite-II” (in which one apical anion is replaced by two O atoms forming a shared edge with a carbonate triangle of the interlayer). Interlayer species include: H2O (billietite, “wyartite II”, and ianthinite), Ba2+ (billietite) Ca2+ (”wyartite II”), and CO3−2 (”wyartite II”); there is no interlayer in β-U3O8. The similarity of known TUE coordination polyhedra with those of U suggests that the β-U3O8 sheet will accommodate TUE substitution coupled with variations in apical anion configuration and interlayer population providing the required charge balance.

To solve the problem of silts and soils that are contaminated with radioactive and toxic substances, the following method has been developed at SIA RADON. The material is mixed with limestone and other components, including up to 70 % (mass) of dried residue of liquid radioactive waste. The mixture is heated at 800 to 1000 °C, shredded, and used to form cement. This cementation process may be used to treat radioactive or other chemical waste.

The work demonstrated that in the case of silts, for example, the product volume is reduced by a factor of 1.5 to 3 compared to the initial silts volume. A fast hardening, durable product is obtained, the quality of which is not inferior to that obtained by using the traditional binders. For some parameters, e.g., hardening rate and macroelement leaching, the current method is significantly better than the traditional cementation process. It has also been demonstrated that, over a wide range of parameters, when dry residue of liquid radioactive waste is used in the initial mixture, the part of NOxCan be reduced to nitrogen, thereby reducing NOx emissions.

The gamma-ray initiated oxidation of dicyclohexano-18-crown-6 (DCH-6) was studied in two-phase system under agitation: nitric acid, aqueous solution and organic solvent.

The distribution coefficients of radionuclides, hydrodynamic characteristics (phase separation and interface surface tension), and the radiolysis products after irradiation at a dose rate (60Co) of 1.7 Gy/s were determined. The maximum total dose was 3.O×105 Gy (3.0×107R).

The possibilities for using various technologies for liquid radioactive waste (LRW) management were considered. Technologies based on the application of sorbents highly selective to radionuclides were shown to have good prospects. Possible ways of increasing selective sorbent efficiency in cleaning radionuclide-polluted waters of various types were examined. The results of studying new high-selective sorbents produced in the Institute of Chemistry FEDRAS are given. The use of these sorbents enables one to treat LRW of various types in a single stage. The results of testing a sorption filter installation for management of LRW produced during operation, repair and disposal of nuclear reactors are presented. Use of this installation resolves a major portion of LRW disposal problems.

A reprocessing of uranium hexafluoride waste, which was accumulated in the nuclear industry for many years, is a very vital question now. There are a lot of gas tanks filled with UF6 that have been exposed for a long time and are posing a serious ecological danger. A plasma chemical technique is proposed here to transform the uranium hexafluoride depleted with isotope 235U (less than 0.1%) into the safe storage form - powdered uranium oxides. The second product is HF (70–80% aqueous solution) which may be easily reprocessed by means of the rectification column up to pure anhydrous hydrogen fluoride.

The designed technological scheme and the basic technical characteristics of the plasma chemical converter are presented. The proposed technique is efficient, easy controlled and free of any harmful by-products.

A new repository waste package (WP) concept for spent nuclear fuel (SNF) is being investigated. The WP uses depleted uranium (DU) to improve performance and reduce the uncertainties of geological disposal of SNF. The WP would be loaded with SNF. Void spaces would then be filled with DU (∼0.2 wt % 235U) dioxide (UO2) or DU silicate-glass beads.

Fission products and actinides can not escape the SNF UO2 crystals until the UO2 dissolves or is transformed into other chemical species. After WP failure, the DU fill material slows dissolution by three mechanisms: (1) saturation of WP groundwater with DU and suppression of SNF dissolution, (2) maintenance of chemically reducing conditions in the WP that minimize SNF solubility by sacrificial oxidation of DU from the +4 valence state, and (3) evolution of DU to lower-density hydrated uranium silicates. The fill expansion minimizes water flow in the degraded WP. The DU also isotopically exchanges with SNF uranium as the SNF degrades to reduce long-term nuclear-criticality concerns.

We have studied a co-precipitation method to remove actinides from radioactive liquid waste. Lanthanum phosphate was selected as a co-precipitation material because actinides and lanthanides are among the homology series in the periodic table. In this report, the conditions, under which the lanthanum compound is precipitated was looked at, as well as, ways to widen the pH range in which the precipitation occurrs in sodium nitrate solution. The properties of the resulting precipitates were also investigated. Further, we investigated the incorporation ratio of amerícium (decontamination factor), one of the actinides, into the precipitate from sodium nitrate solution. Lanthanum phosphate was found to be more effective compared to ferric compounds as the co-precipitation material to remove amerícium. The incorporation conditions for strontium using lanthanum phosphate were also investigated. The removal of strontium, by this method, was more effective when the pH is above 7.0.

In a simplified (SiO2, B2O3, Na2O, Al2O3, ZrO2) glass, corresponding to the basic matrix for the French nuclear waste containment glass, the authors developed a molecular dynamics simulation of the atom displacement cascades resulting from a disintegration of the actinides. After simulating the secondary cascades resulting from the first atoms displaced by a collision with a recoil nucleus, an actinide (U) was added to the model and the cascade produced by accelerating this type of atom was explicitly investigated, notably to observe the morphological evolution of the cascades and the resulting changes in the glass structure.