High-level Radioactive Liquid Waste Research Articles

Study of Atmospheric Plasma-Based Mass Separation System for High-Level Radioactive Waste Treatment

Study of Atmospheric Plasma-Based Mass Separation System for High-Level Radioactive Waste Treatment

MoO3 Solubility and Chemical Durability of V2O5-Bearing Borosilicate Glass

In the vitrification of high-level radioactive liquid waste (HLW), the separation of sodium-molybdate melts is a problem because it reduces the chemical durability of the vitrified waste. A glass with both high MoO3 solubility and chemical durability is required for the safe disposal of radioactive waste. In this study, we investigate the effects of vanadium oxide on the phase separation of the molybdenum-rich phase and the water resistance of the resulting glass by phase equilibrium experiments and chemical durability test. Phase equilibrium experiments were performed on SiO2-B2O3-Al2O3-ZnO-CaO-Na2O-LiO2-MoO3 system glasses and on glasses with V2O5 added. The results showed that MoO3 solubility increased when V2O5 was added. The increase in MoO3 solubility in borosilicate melts may be associated with the viscosity-lowering effect of V2O5. Chemical durability tests were performed on borosilicate glass compositions obtained from phase equilibrium experiments. The normalized leaching rates of V2O5-bearing glasses were higher than those of other glasses. This is due to the higher network modifier/network former ratio of the glass tested. The normalized elemental mass loss of glass containing waste components increases with increasing leaching duration. This suggests that the waste component prevents the formation of a gel layer at the reaction front.

Selective Separation of Zr(IV) from Simulated High-Level Liquid Wastes by Zeolites

ABSTRACT Effectively separating Zr(IV) from strong acidic and radioactive solutions is crucial for spent fuel reprocessing plants, but it remains a challenging task. This study investigated the adsorption of Zr(IV) in HNO3 solutions using zeolites as adsorbents. The zeolites were characterized by X-ray diffraction (XRD)，scanning electron microscopy (SEM), transmission electron microscopy (TEM) and X-ray photoelectron spectroscopy (XPS). The effects of adsorption time, HNO3 concentration, the initial concentration of Zr(IV), zeolites dosage, and temperature on the Zr(IV) adsorption behaviors were investigated. The results demonstrated that HY zeolite had a higher adsorption capacity for Zr(IV) than H-SSZ-13 zeolite. The adsorption capacity of HY zeolite was influenced by its Si/Al ratio, which determined the density of adsorption sites. The optimal HY zeolite (HY-25) exhibited a maximum adsorption capacity of 30.438 mg/g in 3 M HNO3 solution. Furthermore, the adsorption isotherms and kinetics of Zr(IV) adsorption were investigated. The adsorption of Zr(IV) on zeolites was endothermic and spontaneous, in accordance with the Freundlich’s isotherm model and pseudo-second-order kinetic model. In a simulated strong acidic solution of Zr(IV) and 10 co-existing cations (Ag+, Ba2+, Cs+, Ce3+, Eu3+, Fe3+, La3+, Nd3+, Sm3+, Sr2+), HY-25 exhibited good selective adsorption of Zr(IV), indicating its potential application in the treatment of high-level radioactive liquid waste.

Chromatography separation of minor actinides from Ln3+ rich high-level radioactive liquid waste (HLW) using bis-triazinyl-pyridine (BTP) resin

ABSTRACT Separation of Minor Actinide (MA) from High-level Radioactive Liquid Waste (HLW), especially from Ln3+, is greatly needed in the nuclear fuel cycle to reduce the long-term radiotoxicity associated with the waste and for transmutation use. Devising a simple process and designing an extractant is challenging due to the chemical similarity of An3+ and Ln3+ in aqueous solutions. Hence, a tailor-made extractant was prepared, and the partitioning of MA (An3+: 241Am and 244Cm) from Ln3+ rich HLW has been demonstrated using indigenously developed Camphor-based Bis-Triazinyl-Pyridine (CA-BTP) impregnated XAD−7 (BTP Resin). The BTP resin exhibited good adsorption selectivity toward An3+ over Ln3+, acceptable adsorption kinetic, and adequate stability against HNO3 (<2.5 mol/L) and γ-irradiation. The % decontamination factor (DF) achieved was > 99.99 for 137Cs and 154Eu, and 99.7 for Nd. Separated 241Am was precipitated with a saturated solution of oxalic acid. The performance of BTP resin reveals that it has an excellent potential to separate 241Am from Ln3+-rich HLW solution in pure form.

Recovery of platinum group metal resources from high-level radioactive liquid wastes by non-contact photoreduction

Recovery of platinum group metals (PGMs) including palladium (Pd), rhodium (Rh), and ruthenium (Ru) from high-level radioactive liquid waste (HLLW) possesses enormous environmental and economic benefits. A non-contact photoreduction method was herein developed to selectively recover each PGM from HLLW. Soluble Pd(II), Rh(III), and Ru(III) ions were reduced to insoluble zero-valent metals and separated from simulated HLLW containing neodymium (Nd) as a representative for lanthanides, another main component in HLLW. Detailed investigation on the photoreduction of different PGMs revealed that Pd(II) could be reduced under 254- or 300-nm UV exposure using either ethanol or isopropanol as reductants. Only 300-nm UV light enabled the reduction of Rh(III) in the presence of ethanol or isopropanol. Ru(III) was the most difficult to reduce, which was only realized by 300-nm UV illumination in isopropanol solution. The effects of pH was also studied, suggesting that lower pH favored the separation of Rh(III) but hindered the reduction of Pd(II) and Ru(III). A delicate three-step process was accordingly designed to achieve the selective recovery of each PGM from simulated HLLW. Pd(II) was reduced by 254-nm UV light with the help of ethanol in the first step. Then Rh(III) was reduced by 300-UV light in the second step after the pH was adjusted to 0.5 to suppress the Ru(III) reduction. In the third step, Ru(III) was reduced by 300-nm UV light after isopropanol was added and the pH was adjusted to 3.2. The separation ratios of Pd, Rh, and Ru exceeded 99.8%, 99.9%, and 90.0%, respectively. Meanwhile, all Nd(III) still remained in the simulated HLLW. The separation coefficients between Pd/Rh and Rh/Ru exceeded 56,000 and 75,000, respectively. This work may provide an alternative method to recover PGMs from HLLW, which minimize the secondary radioactive wastes compared with other approaches.

Removal of NOx using ozone injection and subsequent absorption in water

In nuclear reprocessing plant reduction of the nitric acid from the high level radioactive liquid waste is accomplished by the addition of formaldehyde. A significant amount NOx is generated during this conventional acid killing process. In this case NO oxidation followed by NO2 absorption is considered as a suitable option for NOx eradication, which also produces reusable nitric acid. Hence, investigations to explore the NO oxidation in the presence of O3 have been carried out for a relatively large system where the total flow rate of gas stream was 800 SLPM (52.4 m3/ h at 25 °C) and the concentration of NO was 480 ppm. It was observed that the concentration of NO decreases with the increase in the concentration of O3 added. A mathematical model has also been developed to predict the concentration of the gaseous components at the exit of the ozonization chamber. Model calculation shows that the reaction between NO and O3 is completed within fraction of a second. Further investigation on the absorption of NO2 in water was carried out by passing the exhaust gas through a scrubber. The absorption of NO2 in water leads to the formation of NO. An attempt has been made to enhance the performance of the water scrubber by injecting ozone into the gas phase during NO2 absorption. The model prediction reveals that N2O5 is the ultimate fate of the NO2 in the presence of O3. The evolution of N2O5 can be considered as two-step process. Reaction between NO2 and O3 is slow and the rate controlling step. On the contrary NO3 reacts quite rapidly with remaining NO2 and produces N2O5.

Irradiation-Assisted Microstructure Evolution and Mechanical Properties Loss of 310S Welded Joints

In order to reveal the effect of irradiation damage caused by high-level liquid radioactive wastes on the welded joint of the container, the irradiation-induced microstructure evolution and mechanical properties degradation of the 310S stainless steel welded joints were investigated in this study. For this purpose, the 1.3 MeV 60Co and 2 MeV accelerators were used to simulate irradiation experiments on 310S welded joints. The uniaxial tensile tests characterized the specimens' mechanical properties and fracture morphology. The results revealed that elongation was reduced by about 5% of irradiation damage by 60Co, and the fracture morphology shows a large number of secondary cracks. In contrast, the elongation was recovered irradiated by the accelerator, and the fracture morphology showed a large number of dimples. Following the interrupted creep deformation, creep fracture tests were conducted for irradiation specimens. The 60Co irradiation damage significantly decreases the creep resistance, leading to deformation of creep, which is increased to 1.5 times that of those unirradiated specimens. At the same time, the ductility is seriously degraded for the irradiated creep fracture specimens. As a result, the creep fracture strain of 60Co specimens is reduced to 70% of that of unirradiated specimens. Further, ductility reduction was related to the irradiated hardening by 60Co, while Nano-indenter hardness was 5.9 GPa, higher by 44% than the unirradiated specimens. The results are shown in an enrichment of Cr, C and P elements at phase boundaries for 60Co irradiation specimens, while the magnitude of element segregation increased by the accelerator combination irradiation. Finally, the creep cracking analysis results show intergranular cracking was observed on the surfaces of the irradiated specimens, while the M23C6 has a primary relationship with the intergranular cracks. The synergic effect of irradiation promoted damage, and element segregation was the primary cause of the intergranular cracking of the 310S welded joints.

Effects of rare earth variation on feed-to-glass conversion during nuclear waste vitrification

Vitrification has been considered the primary option for immobilization of the high-level radioactive liquid waste (HLLW); however, the compositional variation of HLLW makes the reactions between wastes and glass frits rather complex. Rare earth elements are one of the most enriched elements in HLLW; thus, effects of rare earth (Nd2O3 as a surrogate) variation on feed-to-glass conversion were investigated by means of differential scanning calorimetry, (in situ) X-ray diffraction, 29Si nuclear magnetic resonance spectroscopy, and Raman spectroscopy in this work. Differential scanning calorimetry analysis indicates that phase evolution occurs between ∼600 °C and ∼1000 °C and can hardly be affected by Nd2O3 addition. (In situ) X-ray diffraction demonstrates that the HLLW calcines and the glass frits start reacting above 600 °C and intermediate phases appear between 700 °C and 900 °C. The addition of Nd2O3 facilitates the feed-to-glass conversion as all crystalline phases turn to glass in advance with additional Nd2O3. Besides, Si−O units tend to depolymerize with increasing Nd2O3 addition as verified by both 29Si nuclear magnetic resonance spectroscopy and Raman spectroscopy, which indicates that Nd2O3 depolymerizes glass networks and hence promotes the feed reactions.

Effects of Thermal Shock on the Microstructures and Mechanical Properties Evolution of 310S Welded Joints at 1100 °C

In order to reveal the effects of the glass solidification bottling process of high-level liquid radioactive wastes on the welded joints of containers, the microstructure evolution and mechanical properties of 310S stainless steel welded joints were investigated. For this purpose, samples were heat-treated in a resistance furnace at 1100 °C, with two groups of samples being thermally shocked and heat-treated in the furnace. The results indicated that the grain-size distribution changed from unimodal to bimodal for the thermally shocked samples, which was caused by abnormal growth due to the grain growth driving force during recrystallization. Spinel oxide ((Fe, Cr, Ni)3O4) and Cr2O3 were the main oxides at 1100 °C. The dislocations almost disappeared and needle-like structures that were rich in N and Cr formed in the welded joints after being thermally shocked. The tensile properties of the thermally shocked welded joints showed decreases in yield strength and plasticity. The fracture morphologies of the samples heated in the furnace and the as-welded samples presented with dimples. However, the morphologies of the fracture surfaces of the thermally shocked samples presented large numbers of secondary cracks and smooth characteristics.

Extraction behavior of a novel functionalized ionic liquid for separation of platinum group metals from aqueous nitric acid solution.

A novel ionic liquid (IL) functionalized with thiodiglycol amic acid containing a soft S donor was synthesized for the effective and efficient extraction of platinum group metals (Ru, Rh, and Pd) from aqueous nitric acid solutions, such as high-level radioactive liquid waste (HLLW). The IL enabled rapid extraction of Pd(II) with an extraction ratio of approximately 100%. The extractions of Ru(III) and Rh(III) by the IL were slower than that of Pd(II), but the rates were accelerated by temperature elevation. The extractions of Ru(III) and Rh(III) at 50°C reached equilibrium within 4 and 8h, respectively, with the extraction ratios of over 90% without assisting agents or other methods for the extraction system. Furthermore, the IL could extract more than 90% of Ru(III), Rh(III), and Pd(II) from the simulated HLLW within 2h at 50°C.

Nanoscopic structure of borosilicate glass with additives for nuclear waste vitrification

We investigated the nanoscopic structure of borosilicate glasses as a host for high-level radioactive liquid waste (HLLW) in the presence of the additives Na2O and CaO/ZnO with and without Li2O. These additives have been used to lower the glass melting point, suppress the macroscopic phase separation, and increase the chemical durability of glasses. Small-angle neutron scattering was used to elucidate the effect of the additives on the nanoscopic structure along with macro- and atomic-scale observations obtained using neutron imaging and X-ray diffraction, respectively. The borosilicate glass made without Na2O formed a microphase-separated structure of SiO2- and B2O3-rich domains approximately 100 nm in size, and the material was homogeneous at length scales up to millimeters. With the addition of increasing amounts of Na2O, the domain spacing decreased. Introducing CaO/ZnO additives induced inhomogeneities in the glass, such as void structures several nanometers in size, although the inhomogeneity was suppressed by the coaddition of Li2O. These results provide insights into the performance of host glasses for HLLW vitrification. Incorporating HLLW into glasses is likely to cause changes in the nanoscopic structure of host glasses similar to those revealed here.

High acidity- and radiation-resistant triazine-based POPs for recovery of Pd(II) from nuclear fission products

The recovery of platinum group metals (PGMS) from high-level radioactive liquid waste (HLLW) not only has a complementary role in alleviating the shortage of rare precious metals but also is helpful for the final disposal of radioactive waste. In this work, a new cationic covalent triazine framework (CTF) named as cationic pyridyl triazine polymeric network (CPTPN) is designed and synthesized by ZnCl2-catalyzed ionothermal method for recovering of Pd(II) from HLLW. The excremental results shows that palladium can be efficiently adsorbed by CPTPN from 3 M HNO3 solution via anion exchange with fast adsorption kinetics with adsorption equilibrium reached in 10 min. The adsorption isotherm for Pd(II) follows the Langmuir model, which reveals the maximum adsorption capacity (qm) of 339.8 mg g−1 and 389.7 mg g−1 at 3 M HNO3 for CPTPN-Cl and CPTPN-NO3 , respectively. Further studies show that the qm of CPTPN-NO3 in 6 M HNO3 can reach a higher level at 428.6 mg g−1, which is the highest record for Pd(II) recovery in 6 M HNO3. In addition to high thermal stability, CPTPN exhibits a good irradiation stability with excellent adsorption performance maintained after 1000 kGy of γ-irradiation. Furthermore, CPTPN has an excellent adsorption selectivity for Pd(II) in simulated HLLW. Although the selectivity of CPTPN-Cl for Pd(II) is found to be unsatisfactory due to the significant co-adsorption of Ag(I), this problem can be well solved by changing Cl− ions in CPTPN-Cl to NO3−. Both CPTPN-Cl and CPTPN-NO3 show excellent reusability in 5 adsorption–desorption cycles. Thus, CPTPN can be considered as a candidate adsorbent that can tolerate high irradiation and high acidity for selective recovery of Pd(II) from HLLW.

Selective Pd Separation from Simulated Radioactive Liquid Waste by Precipitation Using Xenon Lamp Irradiation for a Simplified Procedure.

In this study, we developed a simple and one-step Pd separation technique based on photoreduction with Xe lamp irradiation for the determination of 107Pd in highly radioactive samples. A simulated high-level radioactive liquid wastes (HLLW) solution, consisting of 14 major elements in a 3 mol L-1 HNO3 solution, was used to evaluate the separation performance. The Pd precipitate was formed by Xe lamp irradiation, and recovered by centrifugation. The Pd recovery from the simulated HLLW solution reached up to 50%, while 99.5% of the other 13 elements was separated. These results indicate that the applicability of the proposed separation technique to HLLW samples.

HAZOP Analysis and Research of Temporary Acid Adding System for High-Discharge Waste Liquid

To prevent the formation of salt from fission products of high-level radioactive liquid wastes (HLWs), a certain amount of acid is added to maintain the acidity of liquid waste. This study analyzes the accidents associated with the addition of acids in a factory by using the hazard and operability analysis (HAZOP), while elucidating the corresponding defects and risks of this approach. By improving the design of the system, the possibility of an accident is significantly reduced. This study can provide guidance for adding acids to treat other high-level waste liquids.

Chemical forms of rhodium ion in pure water and nitric acid solution studied using ultraviolet-visible spectroscopy and first-principles calculations

We have investigated chemical forms of rhodium (Rh) ion in nitric acid solution by using spectroscopic and first-principles methods, in order to develop the disposal of high-level radioactive nuclear liquid wastes (HLLW). The electronic structure of Rh complexes in both pure water and nitric acid solutions was investigated using ultraviolet-visible (UV-vis) absorption spectroscopy. Comparison with many-electron calculations indicates that Rh ion forms an octahedral bidentate complex coordinated with three nitrate ions, [Rh(NO3)3], in both pure water and nitric acid solutions. This is consistent with thermodynamic prediction. Since the overall feature of UV-vis spectra of the Rh complex remained unchanged in a range of 0-3 M nitric acid solutions, the structure of the Rh complex was found to be stable in the range. In addition, since no significant changes in UV-vis spectra of Rh complex were observed under γ-ray radiation, the Rh complex, [Rh(NO3)3], can be expected to be stable in actual HLLW.

The extraction of SR2+ with dicyclohexano-18-crown-6 in conventional organic solvent and ionic liquid diluents

90Sr (t1/2 = 28.8 a), one of the most significant fission products in high-level radioactive liquid waste (HLLW), contributes to a large part of the heat load and radiation. Removal of 90Sr from the HLLW is beneficial for the final treatment of nuclear waste. In this paper, the extraction of Sr2+ was carried out using dicyclohexano-18-crown-6 (DCH18C6) in a variety of diluents including conventional organic solvents and novel ionic liquid solvents. The effect of several factors, such as nitric acid concentration, crown ether concentration and initial strontium concentration on the extraction of Sr2+ have been studied comprehensively. The higher distribution ratio and the stripping efficiency of Sr2+ were obtained with the binary diluents consisted of n-octanol and acetylene tetrachloride, which were compared with that using pure n-octanol as diluent. As for the CnmimNTf2 (n = 2, 4, 6) ionic liquid solvent systems, the distribution ratio of Sr2+ was much higher in the nitric acid medium with low concentration than in the traditional solvent systems. The results showed that DC2mimNTf2 &gt; DC4mimNTf2 &gt; DC6mimNTf2, which indicated that shorter carbon chain benefits the extraction of Sr2+.

Complexation and bonding studies on [Ru(NO)(H2O)5]3+ with nitrate ions by using density functional theory calculation.

Complexation reactions of ruthenium–nitrosyl complexes in HNO3 solution were investigated by density functional theory (DFT) calculations in order to predict the stability of Ru species in high-level radioactive liquid waste (HLLW) solution. The equilibrium structure of [Ru(NO)(NO3)3(H2O)2] obtained by DFT calculations reproduced the experimental Ru–ligand bond lengths and IR frequencies reported previously. Comparison of the Gibbs energies among the geometrical isomers for [Ru(NO)(NO3)x(H2O)5−x](3−x)+/− revealed that the complexation reactions of the ruthenium–nitrosyl complexes with NO3− proceed via the NO3− coordination to the equatorial plane toward the Ru–NO axis. We also estimated Gibbs energy differences on the stepwise complexation reactions to succeed in reproducing the fraction of Ru–NO species in 6 M HNO3 solution, such as in HLLW, by considering the association energy between the Ru–NO species and the substituting ligands. Electron density analyses of the complexes indicated that the strength of the Ru–ligand coordination bonds depends on the stability of the Ru species and the Ru complex without NO3− at the axial position is more stable than that with NO3−, which might be attributed to the difference in the trans influence between H2O and NO3−. Finally, we demonstrated the complexation kinetics in the reactions x = 1 → x = 2. The present study is expected to enable us to model the precise complexation reactions of platinum-group metals in HNO3 solution.

137CS concentrations in surface waters of the seas of Eurasia: results of expeditionary research in 2017

According to results of the field work conducted in 42nd cruise of the research vessel Akademik Boris Petrov (January-March, 2017) and 68th cruise of RV Akademik Mstislav Keldysh (June-August, 2017), the data on concentration of the man-made radionuclide caesium-137 in surface waters of the East China Sea, the South China Sea, Equatorial part of the Indian Ocean, the Red Sea and Mediterranean, Biscay Gulf, the English Channel, the Baltic Sea, North Atlantic, Norwegian Sea and the Barents Sea have been obtained. It was found that concentration of caesium-137 in the surface water of studied Eurasian seas is at a low level. The exception is the Baltic Sea, where the level of caesium-137 is much higher than the values observed before the Chernobyl accident. A higher concentration of caesium-137 in the surface waters of the East China Sea, compared with previous estimates, was found, which may be due to the significant discharge of a high-level liquid radioactive waste after the Fukushima-1 accident in 2011.

Thermo-physical behavior of borosilicate glasses in the presence of high-level radioactive liquid waste constituents

Elemental composition of high-level radioactive liquid waste (HLW) from samples of various batches was assessed by using side-on-view inductively coupled plasma-atomic emission spectroscopy (ICP-AES). Under robust condition (Mg ratio, 10.2 ± 0.4 at RF power 1.3 kW), t he plasma excitation temperature (Texe, 9925 ± 1950 K) was calculated using the Boltzmann plot method. Thermal behavior of major constituents (> 1 g L−1) of HLW, such as nitrates of Na+, Fe3+, Cs+, Nd3+, and UO22+ , were studied using simultaneous thermal analyzer. Subsequently, the thermal stability of five and seven components borosilicate glasses was studied in detail with the above nitrates individually added as additives. Vapor pressure of sodium metaborate (NaBO2, one of the major vapor species from the glass melt in the off-gas system during vitrification of HLW) was estimated using transpiration thermogravimetry. Based on mass loss, the equilibrium vapor pressures (2.51–26.37 Pa) over a temperature span of 1233.15–1343.15 K were estimated. The molar enthalpy of vaporization (ΔvapH°) for sodium metaborate was calculated to be 287.84 ± 0.94 kJ mol−1 using the experimental data.

Cesium-137 Concentration in the Surface Waters of Eurasian Seas: Evidence from the Expedition Research of 2017

New data were obtained during Cruise 42 of the R/V Akademik Boris Petrov (January–March 2017) and Cruise 68 of the R/V Akademik Mstislav Keldysh (June–August 2017) on the concentration of the anthropogenic radionuclide cesium-137 in the surface waters of the East China Sea, South China Sea, equatorial Indian Ocean, Red Sea, Mediterranean Sea, Bay of Biscay, English Channel, Baltic Sea, North Atlantic, Norway Sea, and Barents Sea. Low cesium-137 concentrations were observed in the surface waters of the Eurasian seas. An exception is the Baltic Sea, where the cesium-137 content is much higher than the values obtained before the Chernobyl accident. Compared with previous estimates, higher cesium-137 concentration was observed in the surface water of the East China Sea, which could be related to the extensive discharge of high-level liquid radioactive waste after the Fukushima Daiichi nuclear accident of 2011.