Vitrification Of High-level Radioactive Waste Research Articles

Interaction behaviour of alloy 690 upon exposure to P2O5 containing borosilicate glass at simulated vitrification conditions

This study investigates the interaction between alloy 690 and two types of glasses: pristine borosilicate and P2O5-containing borosilicate glass, at typical glass pouring temperatures encountered during the vitrification of high-level radioactive waste in the back-end of Nuclear Fuel Cycle (NFC). Partial crystallization of both glasses was observed at the alloy 690/glass interface, with certain, though not entirely identical, crystalline phases forming at the interface. The density of these crystalline phases is significantly higher than that of the surrounding glass, which raises concerns about these reaction products settling at the bottom of the furnace. Such sedimentation could potentially obstruct the freeze valve, thereby halting the vitrification process. Additionally, intergranular grooves on the alloy surface exposed to P2O5-containing borosilicate glass were found to disappear with prolonged exposure. This phenomenon is attributed to the strong corrosive action of the highly basic and oxidizing P2O5-bearing glass, leading to the peeling away of the entire exposed surface of alloy 690.

Melting behaviour of simulated radioactive waste as functions of different redox iron-bearing raw materials

Improved understanding of the mechanisms by which foaming occurs during vitrification of high-level radioactive waste feeds prior to operation of the Waste Treatment and Immobilization Plant at the Hanford Site, USA, will help to obviate operational issues and reduce the duration of the clean-up project by enhancing the feed-to-glass conversion. The HLW-NG-Fe2 high-iron simulated waste feed has been shown to exhibit excessive foaming, and the most recent predictive models overestimate the feed melting rate. The influence of delivering iron as a Fe2+-bearing raw material (FeC2O4·2H2O), rather than a Fe3+ (Fe(OH)3) material, was evaluated in terms of the effects on foaming during melting, to improve understanding of the mechanisms of foam production. A decrease of 50.0 ± 10.8% maximum generated foam volume is observed using FeC2O4·2H2O as the iron source, compared with Fe(OH)3. This is determined to be due to a large release of CO2 before the foam onset temperature (the temperature above which the liquid phases forming are sufficiently viscous to trap the gases) and suppression of O2 evolution during foam collapse. Structural analyses of simulated waste feeds after different stages of melting show that the remaining Fe2+ in the modified feed is oxidised to Fe3+ at temperatures between 600 and 800 °C. This feed was tested in a Laboratory Scale Melter with no excessive foaming or feeding issues. Analysis of the final glass products indicates that the glasses produced using the original HLW-NG-Fe2 feed using Fe(OH)3 and the feed made with FeC2O4·2H2O are structurally similar but not identical: the difference in the structure converges when the glass is melted for 24 h, suggesting a transient structure slightly different to that of the baseline in the glass produced using the reduced raw material.

Dissolution behavior of MgO-P2O5 glass system for vitrification of high-level radioactive waste

Dissolution behavior of MgO-P2O5 glass system for vitrification of high-level radioactive waste

Numerical study of noble metals sedimentation during vitrification of high-level radioactive waste in a cold crucible induction melter

Numerical study of noble metals sedimentation during vitrification of high-level radioactive waste in a cold crucible induction melter

Numerical analysis of safety of a borehole repository for vitrified high-level nuclear waste

We developed a model and carried out transient 3-D computer simulation of radionuclides migration from an underground repository for vitrified high-level radioactive waste (HLW) with long-lived actinides. The repository represents an array of deep boreholes. Canisters with vitrified HLW are disposed one above another in lower parts of the boreholes. The upper parts of the boreholes are sealed. Radionuclides leached from the disposed waste are carried away by thermal convection of groundwater caused by heat generation in HLW. The computer simulation was carried out with use of a refined technique taking into account significant difference between radionuclides concentration in the groundwater in the loaded parts of the boreholes and at the earth surface, dependence of the groundwater properties on pT-parameters, composition of HLW. Values of leaching rate of the waste forms by groundwater depending on time and temperature, as well as role of colloid facilitated transport of radionuclides were determined from previous experiments with Na–Al–P-glass which was used in Russian Federation for HLW vitrification. The results of the calculations allowed us to conclude that distance between containers with HLW and the ground surface of 240 m and more is sufficient for safe disposal of the considered HLW.

Formation and prevention of yellow phase in high-level radioactive waste vitrification

経済産業省資源エネルギー庁委託事業「放射性廃棄物の減容化に向けたガラス固化技術の基盤研究」では，将来の発生が見込まれる軽水炉MOX燃料再処理廃棄物のガラス固化技術やガラス固化体への廃棄物の高充填化技術に関する研究開発を行っている。この中で当所は，廃棄物高充填時に課題となる，モリブデン系の析出相(イエローフェーズ)の形成について，そのメカニズムおよび抑制対策を研究している。本稿では，同事業を通して明らかとなったイエローフェーズ形成へのナトリウムの影響について紹介する。

Molecular Dynamics Simulation of Amorphous SiO2, B2O3, Na2O-SiO2, Na2O-B2O3, and Na2O-B2O3-SiO2 Glasses with Variable Compositions and with Cs2O and SrO Dopants.

Selection of suitable glass composition for vitrification of high-level radioactive wastes (HLWs) is one of the major challenges in nuclear waste reprocessing. Atomic and molecular level understanding of various structural, thermodynamical, and dynamical properties of a glass matrix can help in preliminary screening and thus reduce the dependency to some extent on tedious experimental procedures. In that context, extensive molecular dynamics (MD) simulations have been performed to calculate various microscopic properties of the glass matrix. The present article demonstrates that the "Buckingham potential-included long-ranged Coulomb interaction" can be utilized to simulate the glasses of varied compositions. The proposed simulation model has been validated for a wide range of glass compositions: pure glass matrix-SiO2 and B2O3; binary glass mixtures-SiO2-B2O3, Na2O-SiO2, and Na2O-B2O3; ternary glass-Na2O-SiO2-B2O3; and also the Cs2O- and SrO-doped matrix of sodium borosilicate. Most importantly, the MD results have been validated with those of in-house synthesized glasses. The effect of alkali addition on the density and network connectivity of the glass matrix has been explored. The results capture well the boron anomalies for varied concentrations of network formers and network modifiers. The intermediate structural ordering in glasses has been explored by calculating the partial and total structure factors. Further, the characteristic vibration density of states of constituent atoms in the glass matrix is determined. In addition, the glass structures with the addition of dopant oxides Cs2O and SrO have been examined as they are known to be prime heat-generating agents in HLWs. The results establish the structure and dynamics of the doped glass matrix to be a complex nature of the dopant's mass, concentration, charge, and ionic radius. The present MD results might be of great academic and technological significance for further studies in the field of vitrification and prediction of effects associated with the dopant's nature and concentration.

Effects of energy deposition on mechanical properties of sodium borosilicate glass irradiated by three heavy ions: P, Kr, and Xe

Sodium borosilicate glasses are candidate materials for high-level radioactive waste vitrification; therefore, understanding the irradiation effects in model borosilicate glass is crucial. Effects of electronic energy deposition and nuclear energy deposition induced by the impact of heavy ions on the hardness and Young’s modulus of sodium borosilicate glass were investigated. The work concentrates on sodium borosilicate glasses, henceforth termed NBS1 (60.0% SiO2, 15.0% B2O3, and 25.0% Na2O in mol%). The NBS1 glasses were irradiated by P, Kr, and Xe ions with 0.3 MeV, 4 MeV, and 5 MeV, respectively. The hardness and Young’s modulus of ion-irradiated NBS1 glasses were measured by nanoindentation tests. The relationships between the evolution of the hardness, the change in the Young’s modulus of the NBS1 glasses, and the energy deposition were investigated. With the increase in the nuclear energy deposition, both the hardness and Young’s modulus of NBS1 glasses dropped exponentially and then saturated. Regardless of the ion species, the nuclear energy depositions required for the saturation of hardness and Young’s modulus were apparent at approximately 1.2 × 1020 keV/cm3 and 1.8 × 1020 keV/cm3, respectively. The dose dependency of the hardness and Young’s modulus of NBS1 glasses was consistent with previous studies by Peuget et al. Moreover, the electronic energy loss is less than 4 keV/nm, and the electronic energy deposition is less than 3.0 × 1022 keV/cm3 in this work. Therefore, the evolution of hardness and Young’s modulus could have been primarily induced by nuclear energy deposition.

Comparison of hardness variation of ion irradiated borosilicate glasses with different projected ranges

Abstract Borosilicate glass has potential application for vitrification of high-level radioactive waste, which attracts extensive interest in studying its radiation durability. In this study, sodium borosilicate glass samples were irradiated with 4 MeV Kr 17+ ion, 5 MeV Xe 26+ ion and 0.3 MeV P + ion, respectively. The hardness of irradiated borosilicate glass samples was measured with nanoindentation in continuous stiffness mode and quasi continuous stiffness mode, separately. Extrapolation method, mean value method, squared extrapolation method and selected point method are used to obtain hardness of irradiated glass and a comparison among these four methods is conducted. The extrapolation method is suggested to analyze the hardness of ion irradiated glass. With increasing irradiation dose, the values of hardness for samples irradiated with Kr, Xe and P ions dropped and then saturated at 0.02 dpa. Besides, both the maximum variations and decay constants for three kinds of ions with different energies are similar indicates the similarity behind the hardness variation in glasses after irradiation. Furthermore, the hardness variation of low energy P ion irradiated samples whose range is much smaller than those of high energy Kr and Xe ions, has the same trend as that of Kr and Xe ions. It suggested that electronic energy loss did not play a significant role in hardness decrease for irradiation of low energy ions.

Oxidation state and local environment of iron in multicomponent aluminoborosilicate glasses

For safe longterm storage, highlevel radioactive wastes (HLW), generated during the processing of spent nuclear fuel or execution of defense programs, are subjected to hightemperature processing (vitrifi� cation) to obtain aluminophosphatebased (in Russia) or borosilicatebased (abroad) glass (1). Some HLW batches remaining after such processing in Russia and the United States and currently stored in stainless steel tanks contain high concentrations of iron, as well as a number of other transition metals (chromium, man� ganese, nickel) and aluminum. Vitrification of such HLW faces problems since these components contrib� ute to glass crystallization with a change in the rheo� logical properties of glass and in the characteristics of the final products intended for longterm storage. Consequently, it is necessary to develop new glass compositions with a high content of irongroup ele� ments. In the United States, borosilicate glasses are suggested for this purpose (2). In Russia, the only HLW vitrification plant to produce aluminophos� phatebased glass operates at the "Mayak" production association (Chelyabinsk oblast). This glass is also sug� gested for highiron waste; however, the possibility of using borosilicateor aluminoborosilicatebased glass is considered as well (3). In this context, it is necessary to study the influence of iron and other transition met� als on the structure and properties of the glasses, in particular, it is necessary to determine the effect of the oxidation state and speciation of iron on its phase composition and interphase distribution in vit� rified HLW.

A simplified mathematical model of glass melt convection in a cold crucible induction melter

Cold crucible induction melting is emerging as a promising technology for immobilizing nuclear waste in glass matrices. Since the transport properties such as viscosity and electrical conductivity of molten glass exhibit strong temperature dependencies, performance of the cold crucible induction melter is highly sensitive to the thermal field prevailing in the molten glass pool. A simplified mathematical model was developed to numerically investigate the impact of molten glass properties such as viscosity, thermal conductivity and electrical conductivity on the performance of a cold crucible induction melter meant for high level radioactive waste vitrification. The present investigation of the thermal convection using the simplified model confirms the findings of previous studies. Numerical simulation of thermal convection shows that low electrical conductivity and high viscosity existing in the cooler parts of the molten glass bath can lead to poor electromagnetic induction and localized heating in the present melter-inductor configuration. The stable thermal stratification due to bottom cooling leads to a relatively stagnant fluid layer in the lower part of the glass melt. These features of the thermal convection can limit the heat transfer and mixing in the glass melt which in turn can affect both the melting capacity and product homogeneity adversely. Mechanical stirring using a water-cooled stirrer can overcome these limitations. The present study confirms that mechanical stirring of the glass melt can enhance the electromagnetic induction through thermal homogenization. Mechanical mixing eliminates the relatively stagnant fluid layer observed in thermal convection. Distribution of induced power, temperature and velocity predicted by the simplified model exhibit matching characteristics of the results obtained by other investigators. The simplified model reduces the computational load substantially as it eliminates complex electromagnetic computations.

Possibility of Various Types SNF Reprocessing at the PA Mayak exampled with AMB SNF

For the purpose of reprocessing of irradiated nuclear fuel from the water-cooled graphite-moderated pressure-tube reactor named AMB from decomissioned Russian “Atom Peaceful Big”, modernization of the process flow-sheet of the RT-1 plant is being carried out at PA Mayak with participation of FSUE KRI and VNIINM. A particular AMB SNF feature is extremely broad range of fuel compounds with the main ones being the uranium-molybdenum metal, uranium oxide and uranium carbide compositions usually dispersed in magnesium or calcium. Wide range of fuel compositions required to amend SNF dissolution, extraction processing, evaporation of high-level radioactive wastes and vitrification of high-level radioactive wastes. The above set of laboratory research was completed with dynamic tests using samples of AMB from the water-cooled graphite-moderated pressure-tube reactor. Tests have shown the possibility of processing the entire range of AMB SNF at the radiochemical plant RT-1 plant of the PA Mayak. Thus, the ability of the RT-1 plant to process different fuel compositions, including the long-term research reactor fuel have been proved experimentally.

Glass matrices for vitrification of radioactive waste – an Update on R & D Efforts

Radioactive waste gets generated at different stages of nuclear fuel cycle like mining/milling, fuel fabrication, reactor operation, reprocessing of spent fuel and the production & application of radioisotopes for various industrial, medical and research purposes. High Level radioactive Waste (HLW) is generated during reprocessing of spent nuclear fuel and it contains most of the radioactivity present in entire fuel cycle. Vitrification of HLW in borosilicate matrix is being practiced using induction heated metallic melters at industrial scale plants at Tarapur and Trombay [1]. The nature of HLW largely depends on off – reactor cooling of spent nuclear fuel, its type and burn – up, and reprocessing flow sheet. In view of varying characteristics, processing of HLW at Tarapur and Trombay has offered a wide spectrum of challenges in terms of development of matrices and characterization to accommodate compositional changes in waste. The present paper summarizes details of extensive R and D efforts made in the Department of Atomic Energy towards development and characterization of glass formulations for immobilization of HLW.

Vitrification of high-level radioactive waste considering the behavior of platinum group metals

To demonstrate a method using liquid metal for the removal of PGMs during the vitrification process of high-level radioactive waste, removal of Pd was performed using Cu from molten glass containing fission products such as Nd, Sr, Zr, Mo, Te and Ni. Almost all the Pd, 93%, was extracted into liquid Cu and removed as a separable Cu–Pd metal button from the glass. Tellurium and Ni were also extracted 42 and 5.6%, respectively. Nearly 100% of the other elements, especially the heat generating elements such as Sr and Cs, for which Na was used as a substitute of Cs remained in the glass.

06/01639 Commissioning and operation of high level radioactive waste vitrification and storage facilities: the Indian experience: Raj, K. International Journal of Nuclear Energy Science and Technology, 2005, 1, (2–3), 148–163

A simulated high level waste (HLW) containing 4 mass% chrome oxide, whose overall composition is representative of the high chrome oxide wastes at Hanford WA USA, was easily vitrified in a phosphate glass at temperatures ranging from 1150 °C, for waste loadings of 55 mass%, to 1250 °C for waste loadings of 75 mass%. Even at these high waste loadings, these wasteforms had an excellent chemical durability. The best chemical durability was achieved when the O/(Si + P) atomic ratio was between 3.5 and 3.8. These wasteforms were also resistant to crystallization although trace amounts of crystalline Cr2O3 were present in wasteforms containing more than 70 mass% HLW. It is concluded that up to 45 mass% of the total HLW at Hanford, especially that containing as high as 4.5 mass% chrome oxide, could be directly vitrified into an iron phosphate glass, that meets all of the current chemical durability requirements by simply adding 25–35 mass% P2O5 to the waste and melting the mixture at 1150–1250 °C for a few (<6) hours.

Commissioning and operation of high level radioactive waste vitrification and storage facilities: the Indian experience

In India, R&D work for management of high-level radioactive liquid waste (HLW) was started along with inception of the Indian Atomic Energy Programme. This culminated in the setting up of the first vitrification facility at the Waste Immobilisation Plant (WIP) at Tarapur. The second vitrification facility has been commissioned at WIP at Trombay and the third such facility is being set up at Kalpakkam. Vitrified waste product (VWP) canisters generated from both Tarapur and Trombay are destined for storage at solid storage surveillance facility (SSSF) at Tarapur, which is the first Indian facility for interim storage vitrified waste product. This paper presents the details of commissioning and operation of these facilities along with work being done on long-term characterisation of vitrified waste product and future vitrification programme.

Research and development of a high-efficiency one-stage melting converter-burial-bunker method for vitrification of high-level radioactive wastes

A new high-efficiency one-stage melting converter-burial-bunker method for vitrification of high-level radioactive wastes has been developed and investigated. The method includes evaporation (concentration), calcination, and vitrification of high-level radioactive wastes in a one-stage process inside a melting converter for non-metallic minerals, followed by burial inside a bunker-storage facility located directly underneath a melting chamber. Specific to the melting process is the direct combustion of a gas-oxygen-air mixture inside a melt. The experimental data for different aspects of the proposed method are presented, including converter/bunker dimensions, burner types and sizes, data for used materials, contents of saturated salty solution and final glass product, and entrainment analysis. The effective flue-gases cleaning systems and the design of the burial-bunker storage facility are also discussed.

Vitrification of High-Level Radioactive Waste by Sintering Under Pressure

ABSTRACTSintered silicate glass waste forms with a continuous glass phase and embedded but not completely dissolved waste particles were prepared by powder technology. We used a typical defense waste compositions, containing 38 wt.% of ZrO2 and prepared sinter glass samples with 30, 35, and 40 wt.% waste loading at 830°C and 27.6 MPa, by hot isostatic pressing. The glass former was amorphous silica powder. Simulated waste was added as calcine. Samples were characterized using scanning and transmission electron microscopy. The results show that extensive sintering took place, particularly with the 40 wt.% samples. Waste components, such as Na2O, CaO, MnO2, La203, Fe2O3, NiO, Cr2O3, and P2O5, dissolved completely in the continuous glass phase. ZrO2 was partly dissolved and occurred as aggregates of tiny baddeleyite crystals. The porosity of the samples was 2-3%, indicating that the optimum sintering conditions have not been met to reach porosities &lt;1%.Chemical durability tests were carried out with sintered glass with 35 wt.% waste loading in deionized water at 90°C under static conditions, showing that the glass is at least as chemically durable as a borosilicate glass with 31 wt.% melted at 1350°C.

High-Level Radioactive Waste Vitrification Technology and its Applicability to Industrial Waste Sludges

High-Level Radioactive Waste Vitrification Technology and its Applicability to Industrial Waste Sludges

Influence of iron ions on the structural properties of some inorganic glasses

The effects of iron on the structural properties of Zn-borosilicate glass and Pb-metaphosphate glass were studied using X-ray diffraction,57Fe Mossbauer spectroscopy and IR spectroscopy. Zn-borosilicate glass was prepared with varying amounts of Fe2O3 (up to 30% wt.). It was found that the chemical form of added iron (α-FeOOH, α-Fe2O3 or Fe3O4) affects the Fe3+/Fe2+ ratio, as well as the distribution of iron ions at different coordination sites. At high concentration of iron the crystallization of zinc ferrite in the glass matrix takes place. X-ray diffraction and57Fe Mossbauer spectroscopy showed that the amount of zinc ferrite in Zn-borosilicate glass decreases with the following order of addition: α-FeOOH→α-Fe2O3→Fe3O4. In Pb-metaphosphate glass doped with high concentration of α-Fe2O3, the crystallization of Fe3(PO4)2 is pronounced. The assignments of IR band positions and the corresponding interpretation are given. The importance of this study for the technology of vitrification of high-level radioactive wastes is emphasized.