High-level Liquid Waste Research Articles

MINERAL RESOURCES OF THE NUCLEAR INDUSTRY OF RUSSIA AND ISOLATION OF RADIOACTIVE WASTE

The tasks of ensuring the sustainable development of the nuclear energy industry in Russia are considered, which are aimed at solving the problems of the initial and final stages of the nuclear fuel cycle from the reproduction of the mineral resource base of uranium to the management of high-level liquid waste from the reprocessing of spent nuclear fuel. An analysis was made of the trends in the development of the mineral resource base of uranium in connection with the predicted growth in electricity generation at nuclear power plants. Based on the example of the territory of southeastern Transbaikalia, within the framework of topical mineralogical studies, new approaches to the allocation and justification of prospecting areas are proposed based on models of mineral-forming systems of uranium deposits using Earth remote sensing data, GIS modeling and neural network technologies. A detailed analysis of trends in solving the problems of the final stages of the nuclear fuel cycle has been carried out. For isolation and subsequent disposal of the REE-actinide fraction of high-level waste, it is proposed to use phosphate glass ceramics with monazite

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.

Retrieval and processing of radioactive muck settled in the bottom of radioactive liquid waste storage tanks

Alkaline radioactive liquid waste storage tanks contain huge quantity of chemical muck settled at the bottom of the tank. This muck contains fission products, corrosion products, actinides and other chemicals. As the precipitate remain settled for decades, the elements get converted from hydroxides to oxy-hydroxides and oxides which are difficult to dissolve in mineral acids. Treatment feasibility of this legacy muck was studied by simulating muck using acidic radioactive liquid waste followed by its dissolution and conditioning in borosilicate glass matrix. Dissolution was better in a mixture of nitric acid and oxalic acid medium. Immobilization studies of the muck dissolved solution in borosilicate glass showed that 25% waste oxide could be immobilized in glass matrix whereas when mixed with High Level liquid Waste (HLW), 34% waste oxide could be immobilized in glass matrix. Chemical durability and thermal stability of the glass samples were evaluated and are within the appropriate limit.

A functionalized ionic compound for the extraction of ReO4−/TcO4− from highly acidic environment

Selective extraction of TcO4− from high-level liquid waste (HLLW) is essential for spent fuel reprocessing but most of the available extractants can't effectively extract TcO4− from highly acidic environment. In this manuscript, a functionalized ionic compound (I-TPyam) that contains the amide group was synthesized, which can efficiently extract ReO4− from highly acidic environment. 0.3 M I-TPyam solution can extract 97.3 % and 88.0 % of ReO4− from ReO4− solutions that contains 1 M and 3 M nitric acid. Even in more realistic simulated HLLW that contains 22 metal ions and 2.5 M HNO3, it can still extract 87.0 % of the Re. This performance can be further improved by two-stage extraction or denitration treatment, with extraction efficiency increased to 94.9 % and 96.4 %, respectively. Besides, this extractant can also effectively extract Re from neutral ReO4− solution and simulated LAW, with distribution ratio reached 10,468 and 246, respectively. A small portion of Mo will be co-extracted but can be selectively stripped in the stripping process. Our XPS experiments, along with DFT simulations highlighted the predominate role of pyridinium and amide groups in Re (Tc) extraction, revealed the driving forces and explained why I-TPyam prefer to extract ReO4− instead of NO3−.

Vitrification of simulated high-level liquid waste by laser

Vitrification of simulated high-level liquid waste by laser

Flexible waste management system for the future application of MA P&T technology to the current high-level liquid waste

ABSTRACT The flexible waste management (FWM) system has been proposed in order to apply the partitioning and transmutation (P&T) technology for minor actinides (MA) to the current high-level liquid waste (HLLW). This system consists of HLLW calcination to produce easy-to-dissolve granule; its temporary storage is required until the P&T technology deployment and re-dissolution of the granule to remove MA from HLLW after the future P&T technology deployment. The fundamental experiments were conducted to clarify the applicability of FWM system to the current HLLW, and the results indicate the technological validity of the system. The evaluations were carried out for the reduction of storage amount, potential radiotoxicity, and repository area of high-level waste.

Fast removal of strontium from nitric acid solution using an acid-resistant silica-based adsorbent

To efficiently separate strontium (Sr) from simulated high level liquid waste (HLLW), a novel acid-resistant inorganic-organic hybrid adsorbent, named as HEMAP/SiO2-P, was prepared by loading 2-hydroxyethyl-2-methyl-2-propenoate phosphate (HEMAP) into styrene-divinylbenzene modified silica (SiO2-P) with vacuum impregnation method. The experimental results showed that the adsorbent exhibited good adsorption selectivity with the SFSr/metal ions over 64.2 in 3 M HNO3 simulated HLLW. In addition, the adsorption of Sr by the adsorbent was in accordance with the pseudo-second-order kinetic model and the Langmuir isotherm model, and the adsorption equilibrium was obtained within 1 min with the saturation adsorption capacity of 61.2 mg/g. The adsorption of Sr was not markedly changed with increasing temperature. Furthermore, the results of adsorption experiments, characterization analysis, extraction experiments and density functional theory (DFT) calculations revealed that two organic ligand molecules were coordinated with one Sr molecule and simultaneously involved two nitrate molecules to maintain the charge balance during the adsorption process.

Cyclohexyl substituted diglycolamide ligands for highly efficient separation of strontium: Synthesis, extraction and crystallography studies

A series of novel diglycolamide (DGA) ligands containing cyclohexyl substituent on the amidic N atom were investigated for their solvent extraction and complexation ability toward Sr2+ in HNO3 solution. Compared with the most commonly used N,N,N′,N′-tetraoctyldiglycolamide (TODGA), the introduction of cyclohexyl substituent into the DGA skeleton can significantly enhance the extractability toward Sr2+ in the case of toluene as the diluent. And the more cyclohexyl substituents there are, the more significant the effect is. About 8−80 fold increase of the Sr2+ distribution value for LI−LVI ligands can be achieved than that for TODGA at 2.0 mol/L HNO3. The extractability of these cyclohexyl substituted DGA ligands follows the order of LI ≈ LIII > LII > LIV > LV > LVI > > TODGA. Meanwhile, for the DGA ligands with the same number of cyclohexyl substituents, LIII with moderate alkyl chain length (n-butyl) has higher extractability than LII with short chain (n-ethyl) and LIV with long chain (n-octyl). Slope analysis reveals the formation of 1:3 metal-ligand extracted species during the extraction process, which is also confirmed by the results obtained from ESI-MS, FT-IR and single-crystal X-ray diffraction analyses. Furthermore, the single crystal of DGA ligand with Sr(NO3)2 was obtained for the first time, and the analysis of its structure indicates that each DGA molecule participates in the complexation with the 9-fold coordinated Sr2+ through two amide carbonyl O atoms and one ether O atom, accompanied by two free NO3− in the outer coordination sphere. The extraction of Sr2+ by the novel DGA ligand is a spontaneous and exothermic process of increasing entropy, and follows neutral complexing extraction model. In addition, an extraction test of Sr2+ from simulated high level liquid waste was also carried out and demonstrates a highly efficient and selective separation of Sr2+ over the other co-existing ions.

Efficient separation of strontium in different environments with novel acid-resistant silica-based ion exchanger

Radioactive strontium (90Sr) is widespreadly concerned due to its high biochemical toxicity, relatively long half-life, high fission yield and potential special applications in heat source and medical field. Herein, a novel acid-resistant silica-based adsorbent Sb2O5/SiO2 was prepared by a two-step method, i.e. vacuum impregnation followed by oxidation. The adsorbent was characterized by XRD, SEM, TEM, EDS, BET, FT-IR and XPS techniques, which revealed its characterization of small particle size (75–150 μm), abundant functional groups and well-crystallized structure. The experimental results revealed Sb2O5/SiO2 exhibited good adsorption selectivity for Sr in pH 9–3 M HNO3 solution, the adsorption equilibriums were obtained within 5 min in pH 6 solution and 30 min in 1 M HNO3 solution with the adsorption capacities of 160.6 mg/g and 51.8 mg/g, respectively. Moreover, it also demonstrated good application potential in medical isotope preparation area as efficient strontium-yttrium (Sr-Y) separation was achieved with the separation factor over 7769 in pH 6 solution. Furthermore, the column experiments confirmed that Sr could be effectively removed both in seawater and simulated high level liquid waste (HLLW), and the purity of Y could reach to 99.7 % in Sr-Y separation. Finally, density functional theory (DFT) calculations confirmed that the adsorption mechanism was ion exchange between Sr(II) and H+ (in Sb-OH), and the adsorption was accompanied by the transfer of charge and the decrease of adsorption energy.

Efficient separation of palladium from nitric acid solution by a novel silica-based ion exchanger with ultrahigh adsorption selectivity

Palladium (Pd) in high level liquid waste (HLLW) is a valuable alternative resource besides its natural reserves. On the other hand, the presence of Pd has extremely negative impacts on the vitrification process of HLLW. To separate Pd from HLLW, a novel silica-based adsorbent named HEMAP/SiO2-P was designed and prepared by using 2-hydroxyethyl-2-methyl-2-propenoate phosphate as the functional group donor, dichloromethane as the diluent and styrene–divinylbenzene modified silica particle as the carrier. The experimental results reveal that HEMAP/SiO2-P possesses excellent adsorption selectivity (SFPd/other metals > 183.2) and large adsorption capacity (393.4 mg/g) towards Pd in 1 M HNO3 solution. Moreover, column experiments show that the adsorbent can effectively separate Pd from other coexisting metal ions in simulated HLLW with the recovery yield of 99.4%. The experimental and characterization analyses as well as DFT calculations confirm that the adsorption mechanism involves ion exchange between Pd(II) and H+ (in P-OH) accompanied by the transfer of charge.

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.

Complexation and Separation of Palladium with Tridentate 2,6-Bis-triazolyl-pyridine Ligands: Synthesis, Solvent Extraction, Spectroscopy, Crystallography, and DFT Calculations.

Selective extraction of palladium from high-level liquid waste (HLLW) is desirable for the sustainable development of nuclear energy and resource recovery. In this work, three tridentate 2,6-bis-triazolyl-pyridine ligands (L-I, L-II, and L-III) bearing different alkyl side chains were synthesized and systematically studied for the complexation and extraction of palladium. Altering the alkyl side chains of the ligands led to pronounced differences in extraction performance. Among the three ligands, L-II decorated with two n-octyl groups exhibited the highest Pd(II) extraction efficiency at acidity levels of 1-5 M HNO3 and outstanding selectivity over 13 coexisting competing metal ions. Results from UV-vis titration experiments and theoretical calculations suggested that the differentiated extraction abilities of the ligands could be because of their different hydrophilicity rather than electron-donating effects. Slope analyses and electrospray ionization-high resolution mass spectrometry (ESI-HRMS) experiments revealed the formation of both L/Pd 1:1 and 2:1 species during the extraction process. These stoichiometries were further confirmed by job plots and NMR titration experiments. The ligands were found to aggregate slightly, especially at higher concentrations, which could result from multiple intermolecular hydrogen bonds as illustrated by X-ray crystallography. The configurations of PdL and PdL2 were further elucidated by analysis of single crystal structure and density-functional theory (DFT) calculations, respectively, where the first coordination sphere of Pd(II) was surrounded by four nitrogen or oxygen atoms in a quadrangular manner. This study provides an alternative method to separate palladium from HLLW and brings a new understanding of the coordination and complexation behaviors of Pd(II) with tridentate nitrogen ligands.

Efficient separation of americium by a mixed solvent of two extractants, a diamideamine and a nitrilotriacetamide.

The Japan Atomic Energy Agency (JAEA) has proposed the Solvent Extraction from Liquid waste using Extractants of CHON-type for Transmutation (SELECT) process by solvent extraction as a new separation technology to recover minor actinides (MA) from high-level liquid waste (HLLW) produced by spent fuel reprocessing. The MA separation in the SELECT process comprises the recovery of MA and rare earths (RE) from HLLW, MA/RE separation, and Am/Cm separation. Three highly practical extractants are used in the MA separation. Furthermore, this flow configuration facilitates the preparation of nitric acid concentrations in the aqueous phase. However, the separation factor between Cm and Nd in the MA/RE separation is small (SFCm/Nd = 2.5), requiring many extraction stages for continuous extraction in a mixer settler. Therefore, this study investigated the separation of only Am from an aqueous nitric acid solution containing MA (Am and Cm) and RE using an organic phase mixed with two extractants alkyl diamideamine with 2-ethylhexyl alkyl chains (ADAAM(EH)) and hexa-n-octylnitrilotriacetamide (HONTA) used in the SELECT process. Under high-concentration nitric acid conditions, Am and La, Ce, Pr, Nd (light lanthanides) were extracted in the ADAAM(EH) + HONTA mixed solvent, whereas Cm, medium, and heavy lanthanides, and Y were partitioned in the aqueous phase. Subsequently, only light lanthanides could be back extracted from the ADAAM(EH) + HONTA mixture solvent containing Am and light lanthanides in low nitric acid concentrations. Furthermore, Am could be easily stripped with 0.2M or 5M nitric acid. This method does not require the mutual separation of Cm and Nd, which have low separation factors. Am can be efficiently separated by one extraction and two back extractions, reducing the number of steps in the SELECT process.

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.

Efficient separation of palladium from high-level liquid waste with novel adsorbents prepared by sulfhydryl organic ligands containing imidazole, thiazole and oxazole composited with XAD7HP

Separation of palladium from high level liquid waste (HLLW) not only manages the dispose of HLLW, but could also aid in overcoming the shortage of palladium resources. Herein, three novel adsorbents, namely, 2-mercaptobenzimidazole (MBI)/XAD7HP, 2-mercaptobenzothiazole (MBT)/XAD7HP, 2-mercaptobenzoxazole (MBO)/XAD7HP were prepared by vacuum impregnation method using Amberlite XAD7HP resin (XAD7HP) as the carrier for the selective separation of Pd(II) from simulated HLLW. SEM-EDS (Scanning Electron Microscopy), BET (Brunauer-Emmett-Teller) and PXRD (Powder X-ray diffraction) analysis showed that the ligands were successfully impregnated into the carrier. The saturated adsorption capacities of MBI/XAD7HP, MBT/XAD7HP, MBO/XAD7HP in 0.5 M HNO3 were 126.2 mg Pd/g, 137.6 mg Pd/g, 98.4 mg Pd/g, respectively, and excellent adsorption and separation performance towards Pd(II) (SFPd/other metals > 500 and Kd > 104 mL/g from 0.1 to 4 M HNO3) were also achieved. The adsorption kinetics and adsorption isotherms for Pd(II) matched well with pseudo-second-order kinetic model and Langmuir model, indicating that the adsorption for Pd(II) by three adsorbents were monolayer chemical adsorption. The adsorption mechanism was analyzed by FT-IR, XPS combined with DFT calculation. The extremely high selectivity of three adsorbents towards Pd(II) among 15 metal ions is due to the soft N and S donor atoms at ligands which have good affinity for Pd(II) in HNO3 solution. As MBT/XAD7HP exhibited the best adsorption and separation properties, it is supposed to have more application potential for the separation of Pd(II) from HLLW.

Radiation Resistance and Adsorption Behavior of Aluminum Hexacyanoferrate for Pd.

Irradiation resistance is important for adsorbents used in radioactive environments such as high-level liquid waste. In this work, a silica-based composite adsorbent (KAlFe(CN)6/SiO2) was synthesized and γ-irradiated from 10 to 1000 kGy. The angles of the main X-ray diffraction peaks slightly decreased with the increase in irradiation dose, and a minor decomposition of CN- occurred after irradiation to 1000 kGy, indicating that the KAlFe(CN)6/SiO2 adsorbent could preserve structural integrity with a dose below 100 kGy. In 1 to 7 M HNO3, the adsorption ability of the irradiated KAlFe(CN)6/SiO2 remained performant, with a higher Kd than 1625 cm3 g-1. The adsorption equilibrium of Pd(II) in 3 M HNO3 was attained within 45 min before and after irradiation. The maximal adsorption capacity Qe of the irradiated KAlFe(CN)6/SiO2 on Pd(II) ranged from 45.1 to 48.1 mg g-1. A 1.2% relative drop in Qe was observed after 100 kGy irradiation, showing that γ-irradiation lower than 100 kGy insignificantly affected the adsorption capacity of KAlFe(CN)6/SiO2. Calculating and comparing the structures and free energies of different adsorption products via the density functional theory (DFT) method showed that KAlFe(CN)6/SiO2 was more inclined to completely adsorb Pd(II) and spontaneously generate Pd[AlFe(CN)6]2.

Improvement in the elution performance of an N,N,N’,N-tetraoctyl diglycolamide impregnated extraction chromatography adsorbent using neodymium via micro-particle-induced X-ray emission analysis

An adsorbent used for the recovery of trivalent minor actinides (MA(III); Am and Cm) from high level liquid waste generated from reprocessing of spent nuclear fuel was subjected to micro-particle induced X-ray emission (micro-PIXE) analysis to improve its elution performance. The experimental adsorbent comprised SiO2 particles, a polymer coating, and an N,N,N′,N′-tetraoctyl diglycolamide (TODGA). The particles to be analyzed were subjected to Nd adsorption and an elution operation, but Nd in the adsorbent was found to be uniformly distributed. This might have been caused by individual differences in the amount of impregnated TODGA. The remaining Nd species were not localized to a specific part of the adsorbent after the adsorption operation. Some Nd elements were retained in the adsorbent after elution, probably because of the poor diffusion of the mobile phase inside the adsorbent. An adsorbent having a different microstructure from the first was then evaluated, and rapid elution was observed on new adsorbent along micro-PIXE analysis.

Influence of Radioactive Sludge Content on Vitrification of High-Level Liquid Waste

The radioactive sludges formed at the bottom of high-level liquid waste (HLW) storage tanks pose challenges when the HLWs are vitrified. This study aims to determine the influence of the sludge content (enriched in Na2O, Al2O3, NiO, Fe2O3, and BaSO4) on the structure and properties of waste glasses in order to find the optimal ratio of sludges to HLW during vitrification. In the experiments, the simulated sludge and simulated HLW were mixed at different ratios from 0:8 to 4:4, with an overall waste content of 16 wt %, in a borosilicate glass wasteform. It is found that the glass density, molar volume, sulfur retention, and glass transition temperature changed little when increasing the sludge content of the glasses, while the viscosity, chemical durability, and crystallization features of the glasses varied notably. The crystals formed in the glasses during the thermal treatment were exclusively Fe-substituted diopside (Ca, Mg, Fe)2Si2O6. An increase in the Al2O3 and NiO content of the glasses may have been responsible for the increased crystallinity at high temperatures. The leaching rate of Si, B, Na, and Cs from the glasses declined with the increasing addition of sludge to the glasses. Although all the glasses fulfilled the requirements for vitrification processing and glass-product performance, it is recommended that the sludge content of the whole waste should not exceed 25 wt %. This study guides further research on the immobilization of high-level sludges.

Potential methods to reduce the thermal impact of vitrified waste from spent mixed oxide fuel on geological repositories

Compared to the vitrified, high-level radioactive waste (HLW) from spent UO2 fuel (UO2-HLW glass), that of mixed oxide (MOX) fuel (MOX-HLW glass) exhibits high heat generation over a long period of time. Thus, to reduce the associated waste volume and repository footprint, it is essential to prevent bentonite buffer alteration by mitigating heat generation. In this study, we evaluated five potential methods of decreasing the bentonite buffer temperature in a geological repository for MOX-HLW glass. Of these five methods, the separation of minor actinides (MA) from high-level liquid waste (HLLW) was the most effective. This is because the main nuclide involved in heat generation in MOX-HLW glass is Am-241. An MA separation rate of 81% was required to maintain the bentonite buffer temperature below the Japanese upper disposal limit of 100 °C. Thus, it is important to measure the amount of Am-241 when disposing of MOX-HLW glass. Assuming that only Am-241 was disposed of, the maximum Am-241 waste loading under a bentonite buffer temperature of 100 °C was 0.33 wt%, equating to the disposal of only 1.32 kg of Am-241 per 400 kg of MOX-HLW glass. Thus, there is a need for new methods of treating and disposing of Am-241.Graphical abstractRelationship between bentonite buffer temperature and (a) minor actinide (MA) separation ratio from high-level liquid waste (HLLW), (b) mixed oxide (MOX) high-level radioactive waste (HLW) glass and UO2 HLLW mixing ratio, (c) waste loading in MOX-HLW glass, (d) MOX-HLW glass storage time, and (e) canister footprint, representing five methods of mitigating heat generation from disposed MOX-HLW glass.