Simulated High-level Waste Research Articles

Partitioning of the high-level radioactive waste from the purex process by 40% TBF in isoparaffin using magnesium nitrate as a salting-out agent

Centrifugal contactor rig trials of the partitioning flowsheet were carried out using simulated and real high level waste solutions obtained by reprocessing of WWER-100 spent nuclear fuel with burn-up of 47 MW day/t. 40% TBP in Isopar M was used as a solvent with magnesium nitrate as a salting-out agent. Complete removal of Cs, Sr, Zr and Mo to raffinate was achieved. Rare earth elements, Am and Cm were totally concentrated in TPE and REE product. Comparison of the results of this work with previous trials with iron (III) nitrate as a salting-out agent [1] demonstrated that using magnesium nitrate as a salting-out agent resulted in higher decontamination factors of TPE and REE fraction from Cs (~10000) than using iron (III) nitrate (7500).

Research on calcination thermal decomposition process of simulated high-level liquid waste based on two-step vitrification

The two-step vitrification process involves converting high-level waste from a liquid to a solid state through calcination, followed by melting the calcines with glass matrix material to form glassy wasteforms. The calcination process is a critical stage in the vitrification process. This study investigated the thermal decomposition performance and phase evolution of power reactor simulated high-level waste during the calcination process. As the calcination temperature increased from 100 to 900 °C, the resultant waste calcines showed increasingly darker color and smaller general particle sizes. Below 600 °C, the waste experienced thermal decomposition of nitrate and formate compounds, leading to an approximate 40 wt% weight reduction. At 600 °C, the waste's weight remained constant, and crystalline phases such as ZrxM1-xO2 (M = Ce, Pr, Nd), BaMoO4, CeO2, and GdxCe1-xO2 were observed.

Efficient and Selective Removal of Palladium from Simulated High-Level Liquid Waste Using a Silica-Based Adsorbent NTAamide(C8)/SiO2-P.

In order to realize the effective separation of palladium from high-level liquid waste (HLLW), a ligand-supported adsorbent (NTAamide(C8)/SiO2-P) was prepared by the impregnation method in a vacuum. The SiO2-P carrier was synthesized by in situ polymerization of divinylbenzene and styrene monomers on a macroporous silica skeleton. The NTAamide(C8)/SiO2-P adsorbent was fabricated by impregnating an NTAamide(C8) ligand into the pore of a SiO2-P carrier under a vacuum condition. The adsorption performance of NTAamide(C8)/SiO2-P in nitric acid medium has been systematically studied. In a solution of 0.2 M HNO3, the distribution coefficient of Pd on NTAamide(C8)/SiO2-P was 1848 mL/g with an adsorption percentage of 90.24%. With the concentration of nitric acid increasing, the adsorption capacity of NTAamide(C8)/SiO2-P decreases. Compared to the other 10 potential interfering ions in fission products, NTAamide(C8)/SiO2-P exhibited excellent adsorption selectivity for Pd(II). The separation factor (SFPd/other metals > 77.8) is significantly higher than that of similar materials. The interference of NaNO3 had a negligible effect on the adsorption performance of NTAamide(C8)/SiO2-P, which maintained above 90%. The adsorption kinetics of Pd(II) adsorption on NTAamide(C8)/SiO2-P fits well with the pseudo-second order model. The Sips model is more suitable than the Langmuir and Freundlich model for describing the adsorption behavior. Thermodynamic analysis showed that the adsorption of Pd(II) on NTAamide(C8)/SiO2-P was a spontaneous, endothermic, and rapid process. NTAamide(C8)/SiO2-P also demonstrated good reusability and economic feasibility.

A hydrolytically stable complexant for minor An separation from Ln in process relevant diluents.

The evaluation of bis-1,2,4-triazine complexants containing eight-carbon, alkoxy-functionalized phenyl substituents at the 3,3'-positions for selective minor actinide extraction in simulated high level waste is reported. The complexant retained chemoselective extraction efficiency of actinides over lanthanides while breaking through the nonpolar diluent solubility and hydrolytic stability barriers that limit many BTP complexants. Thus, we report a BTP complexant that is readily soluble in both kerosene and isooctanol without forming precipitates, emulsions, or third phase after contact with nitric acid and extracts 241Am3+ from 154Eu3+. Separations and spectroscopic data are reported herein.

Multi-physical field simulation calculation and analysis of simulated high-level waste liquid spray calcination.

Aiming at the independent research and development of a simulated high-level waste liquid spray calcination transformation treatment test device, a three-dimensional multi-physical field model of spray calcination was established by means of finite element analysis method. In this paper, the simulated high-level waste liquid is a mixed solution of nitrate solution and sucrose. The main chemical components of nitrate dissolution are HNO3 and NaNO3. The process of evaporation and calcination of high-level waste liquid to form oxides is also called the pretreatment of high-level waste liquid or the conversion of high-level waste liquid. In this experiment, the atomized droplets sprayed at high speed are evaporated, dried and calcined in turn in the calciner to obtain the calcined product. The distribution law of temperature flow field and chemical reaction state and results inside the test device were revealed by simulation calculation. The results show that under the condition of multi-physical field coupling, the chemical reaction temperature has an effect on the yield of the product. The temperature is positively correlated with the product concentration, and the effect of temperature on the yield of NO2 is greater than that of Na2O. At the same time, in this chemical reaction, the concentration of reactants (NaNO3 and HNO3) had a positive correlation with the concentration of main products (NO2 and Na2O). However, the rate of increase in the concentration of the main products (NO2 and Na2O) decreased with the increase of the concentration of the reactants (NaNO3 and HNO3).

Selective separation of rare earth, Sr, Mo, and Zr from simulated raffinate of uranium/plutonium co-purification process

Spent fuel contains many elements including those that are radioactive. Separation and recovery of these substances have been challenging. This paper presents an efficient strategy for stepwise separation of the rare earth, strontium, molybdenum, and zirconium from the simulated raffinate of a uranium/plutonium co-purification process. This strategy employed three extractants with different metal ion selectivity, and cyclohexanediamine tetraacetic acid (CDTA) was introduced as a complexing agent to inhibit the extraction of Zr(IV) and precious metals. The recovery efficiency of all the rare earth elements was more than 97 % through the three-stage extraction, with N,N,N′,N′-tetranoctyl-diglycolamide (TODGA) as the extractant and N,N-dibutyloctanamide (DBOA) as the phase modifiers. Sr(II) was selectively recovered using N,N,N′,N′-tetracyclohexyl-diglycolamide (TCHDGA). A bisphosphonic acid extractant, N,N-hexadecylamine bis (methylene phenylphosphonic acid) (HADMPPA), was designed and synthesized for the selective separation of Zr(IV) and Mo(VI). The separation conditions of the recovered elements from other elements in each step were systematically examined. Moreover, the mechanism of the extraction reaction was investigated using FTIR spectroscopy, the slope method, and the equimolar series method. To our knowledge, this is the first time that four groups of major elements have been sequencially separated from the simulated high-level waste liquid in one process. This work could contribute to nuclear fuel reprocessing as well as the sustainable development of nuclear energy.

Selective Separation of Zr(IV) from Simulated High-Level Liquid Waste by Mesoporous Silica.

The efficient separation of Zr(IV) ions from strong acidic and radioactive solutions is a significant challenge, especially in the context of the aqueous reprocessing of nuclear fuels. The complexity of such solutions, which are often characterized by high acidity and the presence of radioactive elements, poses formidable challenges for separation processes. Herein, several mesoporous silicas (HMS, MCM-41, KIT-6, and SiO2-70 Å) with excellent acid and radiation resistance properties were employed as sorbents to remove Zr(IV) ions from simulated high-level liquid waste. The batch experiments were designed to investigate the influence of adsorption time, HNO3 concentration, initial Zr(IV) concentration, adsorbent dosage, and temperature on the adsorption behavior of Zr(IV). The results indicate that the adsorption equilibrium time of mesoporous silica materials was approximately 8 h, and all the adsorption processes followed the pseudo-second-order kinetics equation. The isotherms of Zr(IV) adsorption by KIT-6 exhibited good agreement with the Langmuir model, while the Freundlich model could be utilized to fit the adsorption on HMS, MCM-41, and SiO2-70 Å. The adsorption capacity of MCM-41 for Zr(IV) in 3 mol/L HNO3 was 54.91 mg/g, which is three times the adsorption capacity reported for commercial silica gel (17.91 mg/g). The thermodynamic parameters indicate that the adsorption processes for zirconium are endothermic reactions. Furthermore, the mesoporous silicas exhibited a pronounced selectivity in the adsorption of Zr(IV) within a simulated high-level liquid waste containing 10 co-existing cations (3 mol/L HNO3). This suggests that mesoporous silicas have great potential for Zr(IV) removal in actual radioactive liquids with high acidity during spent fuel reprocessing.

Study on Dynamic Column Behavior and Complexation Mechanism of DBS-Modified Crown Ether-Based Silica to 90Sr.

A crown ether-loaded hybrid adsorbent suitable for the separation and enrichment of strontium from high-level liquid waste was synthesized. 4',4'(5″)-di(tert-butylcyclohexano)-18-crown-6 (DtBuCH18C6) and its modifiers dodecyl benzenesulfonic acid (DBS) and 1-dodecanol were impregnated into silica-based polymer support. The hybrid adsorbent exhibited excellent Sr(II) selectivity ability, and effective chromatographic separation and recovery of Sr(II) from simulated high-level liquid waste could be achieved with a (DtBuCH18C6 + DBS + dodec)/SiO2-P packed column. The recovery rate of Sr(II) calculated based on the mass balance was approximately 99% and over 80% for the other coexisting metal ions. An appropriate increase in the concentration of Na-DTPA eluent was favorable to improve the efficiency of the elution process because of the increased complexation capacity of [DTPA]5- to Sr(II). The developed theoretical model can simulate the dynamic breakthrough curves of the material on the basis of short column data, thereby predicting the scale-up column of the practical operation. Density functional theory calculation was used to explore the action mechanism of DBS modifiers on the Sr(II) complexation process of crown ether groups. Two Sr(II) complexation isomeric models of DtBuCH18C6 were established, and the calculation results revealed a similar complexation ability. DtBuCH18C6 could form a stable Sr(II) complexation structure with DBS coordination, which further indicated that DBS could be a ligand to promote the Sr(II) adsorption ability of crown ether materials.

Precipitation behaviour of simulated high-level liquid waste during the denitration process based on two-step vitrification

In the two-step vitrification process, it is necessary to first reduce the acidity of the waste before calcination. In this study, the influence of formic acid as a denitration agent on the precipitation behavior of a simulated high-level liquid waste during denitration was investigated. The denitration process was conducted at a temperature of 90℃, and the denitration ratio ranged from 1.0 to 2.0. The denitration products were analyzed using various techniques including ICP-OES, XRF, SEM-EDS, Raman, and XRD. The results revealed a significant decrease in nitrate content after denitration when the denitration ratio was between 1.0 and 2.0. As the denitration ratio increased, the nitrate content progressively decreased, while the formic acid content increased. A denitration precipitate containing elements such as Zr, Mo, La, Ce, Nd, Fe, Te, Pr, Cs, Sm, Sr, Y, Co, and Ni was observed. The content of each component and the morphology of the precipitate were found to be correlated with the denitration ratio. Crystallization of MoO2 and Ln2Zr3(MoO4)9 (Ln = La, Ce, Nd, Pr, Sm) in the precipitation occurred when the denitration ratio was between 1.0 and 1.2. On the other hand, predominantly rod-shaped particles (Ln(HCOO)3, Ln = La, Ce, Pr, Nd, Sm) were formed when the denitration ratio was 1.4 or higher.

Statistical glass structure gene modeling on liquidus temperature of high level waste glasses

Statistical glass structure gene modeling (GSgM) on liquidus temperature (TL) of simulated high level waste (HLW) glass is detailed in this article. TL values of a set of HLW borosilicate glasses were experimentally determined by x-ray diffraction (XRD) method with 24 h heat-treated glass in gradient temperature furnace, and the primary crystalline phase was CaMoO4. Integral peak areas obtained from the FTIR peak fitting were used as the structural data, and the statistical TL-structure model was first established. The S-P (TL) model was shown to have a high accuracy, i.e., R2 0.978, P value 0.0009 and RMSE 7.6 °C. Model validation showed the predicted and measured TL values were 960.4 °C and 965 °C, respectively. Model iteration further reduced P value to 0.0001 and RMSE to 7 °C. Relationship between composition (C) and the glass network structure (S) was then simulated, and composition-FTIR structure-TL analysis indicated that SiO2, ZnO, rB3+, MoO3, ZrO2, B2O3 and rare earth affect TL by influence the structural units Si–O–Si rocking vibrations, νs (Si–O–Si) in [SiO4] and νs ([BO3] or BO2O−). Results showed that GSgM offers an alternative approach to simulate TL of complex glass with limited experimental data, providing a plausible linkage between composition, structure and property.

Adsorption behavior of platinum-group metals and Co-existing metal ions from simulated high-level liquid waste using HONTA and Crea impregnated adsorbent

The volume and toxicity of radioactive waste can be decreased by separating the components of high-level liquid waste according to their properties. An impregnated silica-based adsorbent was prepared in this study by combining N,N,N′,N′,N″,N″-hexa-n-octylnitrilotriacetamide (HONTA) extractant, N′,N′-di-n-hexyl-thiodiglycolamide (Crea) extractant, and macroporous silica polymer composite particles (SiO2–P). The performance of platinum-group metals adsorption and separation on prepared (HONTA + Crea)/SiO2–P adsorbent was then assessed together with that of co-existing metal ions by batch-adsorption and chromatographic separation studies. From the batch-adsorption experiment results, (HONTA + Crea)/SiO2–P adsorbent showed high adsorption performance of Pd(II) owing to an affinity between Pd(II) and Crea extractant based on the Hard and Soft Acids and Bases theory. Additionally, significant adsorption performance was observed toward Zr(IV) and Mo(VI). Compared with studies using the Crea extractant, the high adsorption performance of Zr(IV) and Mo(VI) is attributed to the HONTA extractant. As revealed from the chromatographic experiment results, most of Pd(II) was recovered from the feed solution using 0.2 M thiourea in 0.1 M HNO3. Additionally, the possibility of recovery of Zr(IV), Mo(VI), and Re(VII) was observed using the (HONTA + Crea)/SiO2–P adsorbent.

Volatilization of sodium and boron from nuclear waste glass and associated effects on glass structure and thermal stability

During nuclear waste vitrification, loss of sodium (Na) and boron (B) occurs, as these elements are highly volatile at high temperatures, which affects the properties of the glass products. Moreover, the evaporation of volatile nuclides (Cs and Tc) occurs together with that of Na and B, which may cause environmental hazards. Thus far, only a few studies have focused on the volatilization of Na and B from nuclear waste glass and loss-related effects on glass properties. In this study, we investigated the volatilization behaviors of Na and B from a simulated high-level waste glass as functions of heating temperature and dwelling duration, and evaluated the effects of Na and B loss on the glass structure and thermal stability. The results showed that volatilization of Na occurs predominantly through the diffusion-controlled process at a temperature range of 950 °C to 1100 °C, and owing to its higher frequency factor, Na diffuses and volatilizes at a faster rate than that of B. Volatilization of B occurs by both a diffusion-controlled process from bulk to the surface and a chemical reaction process on the surface. Based on the data obtained regarding the composition of Na and B and structure of the glass, a hypothetical model was proposed to explain the volatilization behaviors of Na and B from a structural viewpoint. As the loss of Na and B increases, the amount of BO4− reduces, whereas the amount of BO3 increases; finally, the Si-O network becomes more polymerized and has less thermal stability, resulting in the crystallization of the glass at elevated temperatures. These results would serve as the basis to estimate the amount of Na and B loss and the properties of the glass melt during nuclear waste vitrification, especially in melter idling.

Silica-based covalent organic framework composite for efficient separation and enrichment of palladium and its heterogeneous catalysis application

Considering the necessity and challenge in palladium recovery from high-level liquid waste and hydrogen energy development, the first attempt was made to adsorb Pd(II) from simulated high-level liquid waste firstly, then to achieve in situ reduction by a simple method and followed by using it directly for efficient hydrogen production through the catalytic decomposition of formic acid. In this work, a composite of covalent organic framework and SiO2 (Tp-Azo-COF/SiO2) was prepared by wet impregnation method using SiO2 as a carrier. The characteristics of the composite were examined by SEM, EDS, BET, TG, XRD, etc. The results showed that Tp-Azo-COF/SiO2 exhibited regular silicon-based spherical shapes in microscopic form and good heat resistance. In batch experiments, Tp-Azo-COF/SiO2 was demonstrated excellent adsorption and selectivity for Pd(II) at a wide range of HNO3 acidity (1–5 M), and the separation factor was as high as 2653 at 3 M HNO3. Tp-Azo-COF/SiO2 was able to reach a maximum adsorption capacity of 85.40 mg/g at 120 min at 298 K in 3 M HNO3. FT-IR and XPS analysis show that the adsorption mechanism of Pd(II) by the composite occurs mainly on the C-NH of Tp-Azo-COF, and NO3– is also involved in the coordination. In dynamic column experiments, Tp-Azo-COF/SiO2 was demonstrated a nearly complete separation and recovery of Pd(II) with good reusability, and the enrichment coefficients all maintained at about 30. Finally, an efficient and environmentally friendly process was proposed to achieve the in situ reduction of Pd(II)-loaded Tp-Azo-COF/SiO2. Pd-loaded Tp-Azo-COF/SiO2 was demonstrated better decomposition performance than commercial Pd-based catalysts hydrogen production by the catalytic decomposition formic acid.

Vitrification of simulated high-level liquid waste by laser

Vitrification of simulated high-level liquid waste by laser

Neutron resonance absorption imaging of simulated high-level radioactive waste in borosilicate glass

We performed a preliminary study of neutron resonance absorption imaging to investigate the spatial distribution of constituent elements in borosilicate glasses containing simulated high-level radioactive waste, in which elemental inhomogeneities affect the physical and chemical stabilities of the glass. Dips generated by the resonance absorptions of Rh, Pd, Na, Gd, Cs, and Sm were observed in the neutron transmission spectra of the glass samples. The spatial distributions of these elements were obtained from the neutron transmission images at the resonance energies. The distributions of Rh and Pd visualized the sedimentation of these platinum group elements. In contrast, the lanthanides (Gd and Sm) and Cs were uniformly dispersed. These results show that neutron resonance absorption imaging is a promising tool for characterizing borosilicate glasses and investigating the vitrification mechanism of high-level radioactive waste.

Cuprous Oxide-Based Cationic Hydrogel by the Integration of Enrichment and Immobilization of Radioiodine (I-, IO3-) in Aqueous Solution.

The efficient capture and immobilization of radioiodine (I-, IO3-) is of great importance in radioactive waste management. Here, a Cu2O-loaded three-dimensional bulk cationic hydrogel composite (Cu2O@CH) was successfully prepared by simple redox reactions and UV photopolymerization, which realized the rapid enrichment and efficient immobilization of I- and IO3-. The adsorption experiments showed that the maximum adsorption capacity of Cu2O@CH for I- in the solution at pH = 3 reached 416.5 mg/g, and the adsorption capacity of IO3- in the solution at pH = 6 could reach 313.4 mg/g. It exhibited extremely fast adsorption kinetics for I- and IO3-. In addition, Cu2O@CH also exhibited efficient I- and IO3- removal from simulated high-level liquid waste. The rapid capture and effective immobilization of radioiodine (I-, IO3-) were realized by the electrostatic interaction of -N+(CH3)3 groups in Cu2O@CH with I- and IO3-, as well as the chemical reactions between Cu2O and I-. The bulk cationic hydrogel composite explored the multifunctional role toward fast, high adsorption capability and easy handling, highlighting its superiority compared to the powder adsorbent, which renders it a potential adsorbent for the removal of radioactive iodine (I-, IO3-) in nuclear wastewater treatment.

Corrosion testing of domestically produced baddeleyitecorundum and chromium-containing materials in borosilicate glass melts

Practical importance of corrosion and erosion resistance of refractories was justified for the design of furnaces with direct electric heating that are used for vitrification of high-level waste (HLW). The main promising materials are listed, and the algorithm of assessment of their resistance to the effects of borosilicate melts is given for the operating conditions of electric furnaces. Corrosion testing of refractories was carried out under static and dynamic conditions in a low-melting borosilicate glass melt and in a borosilicate glass melt containing simulated liquid HLW. Refractory materials were selected that have the highest resistance in a glass melt under conditions of HLW immobilization. The testing results will be used for choosing lining material in the design of removable and small-scale melters. Ill. 5. Ref. 11. Tab. 3.

Elimination of the Interfacial Crud in the Extraction of Simulated High-Level Liquid Waste After Denitration in the TRPO Process

ABSTRACT Denitration is a proven technique that can reduce the nitric acid concentration of high-level liquid waste (HLLW) to meet the aqueous acidity requirement of the TRPO (TriAlkyl Phosphine Oxide) process. However, interfacial crud is generated in the TRPO extraction of the HLLW after denitration. In this work, the interfacial crud was characterized by Fourier-transform infrared spectroscopy and dynamic light scattering. The extraction distribution ratio of zirconium by 30% TRPO in kerosene decreases dramatically after the simulated HLLW is denitrated, and we proposed that the presence of zirconium in the simulated HLLW after denitration is responsible for the generation of interfacial crud. Using ammonium molybdate as a precipitant, the concentration of zirconium in the simulated HLLW was successfully reduced without essentially changing the concentration of other metal ions. The effects of reaction time, temperature, and amount of ammonium molybdate added on the zirconium removal process were investigated and the precipitates formed in the removal of zirconium were characterized by scanning electron microscope, energy dispersive X-ray spectroscopy, and X-ray photoelectron spectroscopy. Finally, after reducing the concentration of zirconium in the simulated HLLW to below 10 mg/L by using ammonium molybdate, the extraction experiment was performed again and no interfacial crud was observed.

Influence of Rock Structure on Migration of Radioactive Colloids from an Underground Repository of High-Level Radioactive Waste

Studies of leaching of vitrified simulated high-level radioactive waste (HLW) evidence that most of actinides or their simulators enter leaching water in a colloidal form. In this paper, we consider a mechanism of colloid-facilitated migration of radionuclides from an underground repository of HLW located at a depth of a few hundreds of meters in fractured crystalline rocks. The comparison between data of field and laboratory measurements showed that the bulk permeability of the rock massif in field tests is much greater than the permeability of rock samples in laboratory experiments due to an influence of a network of fractures in the rock massif. Our theoretical analysis presents evidence that this difference can take place even in a case when the network is not continuous, and the fractures are isolated with each other through a porous low-permeable matrix of the rock. Results of modelling revealed a possibility of mechanical retention of radionuclide-bearing colloid particles in the frame of rock during their underground migration.

Crystallization of molybdenum oxide phase from simulated high-level waste glass under slow cooling

Crystallization of molybdenum oxide phase from simulated high-level waste glass under slow cooling