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The effective capture and recovery of radioiodine species associated with nuclear fuel reprocessing is of significant importance in nuclear power plants. Porous materials have been proven to be one of the most effective adsorbents for the capture of radioiodine. In this work, we design and synthesize a series of conjugated microporous polymers (CMPs), namely, TPDA–TFPB CMP, TPDA–TATBA CMP, and TPDA–TECHO CMP, which are constructed based on a planar rectangular 4-connected organic monomer and three triangular 3-connected organic monomers, respectively. The resultant CMPs are characterized using various characterization techniques and used as effective adsorbents for iodine capture. Our experiments indicated that the CMPs exhibit excellent iodine adsorption capacities as high as 6.48, 6.25, and 6.37 g g−1 at 348 K and ambient pressure. The adsorption mechanism was further investigated and the strong chemical adsorption between the iodine and the imine/tertiary ammonia of the CMPs, 3D network structure with accessible hierarchical pores, uniform micromorphology, wide π-conjugated structure, and high-density Lewis-base sites synergistically contribute to their excellent iodine adsorption performance. Moreover, the CMPs demonstrated good recyclability. This work provides guidance for the construction of novel iodine adsorbent materials with high efficiency in the nuclear power field. [ABSTRACT FROM AUTHOR]

The transfer of nuclear spent fuel from the reactor storage pool to dry storage or for reprocessing or final disposition requires information about its isotopic composition, decay heat, and other thermomechanical properties. The spent nuclear fuel assembly of a typical advanced pressurized water reactor, AP-1000, was characterized using the Monte Carlo MCNPX code and SCALE/ORIGEN code. The simulation of operational history started from the operation of the first fresh core for an average fuel assembly with certain physical isotopic parameters until 25 GWd/tonne U discharge burnup. The analysis considered the calculations of the radionuclide inventories, activity, neutron emission spectrum, gamma-ray emission spectrum, and decay power after 700 effective full power days and for post different time ranges until a 1 million–year cooling period. The comparison of some results of the two codes showed small differences due to the consideration of the continuous-energy variation for neutrons in the MCNPX code and the discrete energy assumption in the SCALE/ORIGEN code. [ABSTRACT FROM AUTHOR]

Significant improvements in aqueous reprocessing are a key need in closing the future nuclear fuel cycle and addressing concerns with past industrial reprocessing technologies. Despite much-dedicated efforts over the past several decades, the separation of the large variety of elements and isotopes, such as minor actinides, found in the fuel is unanimously challenging. This review summarizes and classifies various aqueous chelators that have already been implemented or could potentially be used in the aqueous reprocessing of spent nuclear fuel. A variety of processes are briefly introduced, and the application of aqueous chelators is analyzed, with a focus on the elements targeted for separations, as well as aqueous/organic phase composition and acidity, and inherent physico-chemical properties. Comparisons among these agents are made to provide a clear outline of their unique characteristics. Different aqueous chelators function in different ways and thus are applied with different purposes. However, there exist limitations, even for the most widely applied polyaminocarboxylate ligands, and functional groups have been relentlessly modified over the years to achieve more efficient, sustainable, and cost-effective separations. [ABSTRACT FROM AUTHOR]

Spent nuclear fuel contains heavy element fission products that must be separated for effective reprocessing for a safe and sustainable nuclear fuel cycle. 93Zr and 99Tc are high‐yield fission products that co‐transport in liquid‐liquid extraction processes. Here we seek atomic‐level information of this co‐extraction process, as well as fundamental knowledge about ZrIV (and HfIV) aqueous speciation in the presence of topology‐directing ligands such as pertechnetate (TcO4−) and non‐radioactive surrogate perrhenate (ReO4−). In this context, we show that the flat tetrameric oxyhydroxyl‐cluster [MIV4(OH)8(H2O)16]8+ (and related polymers) is dissociated by perrhenate/pertechnetate to yield isostructural dimers, M2(OH)2(XO4−)6(H2O)6 ⋅ 3H2O (M=Zr/HfIV; X=Re/TcVII), elucidated by single‐crystal X‐ray diffraction. We used these model compounds to understand the pervasive 93Zr‐99Tc coextraction with further speciation studies in water, nitric acid, and tetrabutylphosphate (TBP) ‐kerosene; where the latter two media are relevant to nuclear fuel reprocessing. SAXS (small angle X‐ray scattering), compositional evaluation, and where experimentally feasible, ESI‐MS (electrospray ionization mass spectrometry) showed that perrhenate/pertechnetate influence Zr/HfIV‐speciation in water. In Zr‐XO4 solvent extraction studies to simulate fuel reprocessing, we provide evidence that TcO4− enhances extraction of ZrIV, and compositional analysis of the extracted metal‐complexes (Zr‐ReO4 study) is consistent with the crystallized ZrIV2(OH)2(ReVIIO4−)6(H2O)6⋅dimer. [ABSTRACT FROM AUTHOR]

Results of centrifugal contactor rig trials of the flowsheet of the first solvent extraction cycle in protective glove boxes using simulated solutions of spent nuclear fuel and 30 and 45% TBP in isoparaffin as a solvent are compared. It was shown that an increase in the TBP concentration to 45% resulted in deterioration of the quality of the products obtained by the same flowsheets with 30% TBP in Isopar L as a solvent. Comparison of experimental and computer simulated profiles of component distribution across the stages of extraction units demonstrated satisfactory agreement in nitric acid and uranium concentrations. At the same time, it was found that simulation model of zirconium, technetium, and neptunium extraction requires improvement considering interaction of these elements with complexing agents. [ABSTRACT FROM AUTHOR]

Nuclear energy holds the chief portion of the global primary energy mix that comes with the major issue of releasing volatile nuclear wastes viz. radioiodine (129I and 131I) into air and water bodies during nuclear fuel reprocessing. The efficient capture of volatile radioiodine has attracted a major attention worldwide due to the severe health and environment effects. Among various porous materials, covalent organic frameworks (COFs) are the intriguing class of porous organic materials with crystallinity, synthetically pre-designable functionalities to achieve tunable properties. Herein, a new benzothiazole based COF (COF-3) obtained by the condensation of 2,6-dimainobenzothiazole and 2,4,6-triformylpholoroglucinol is reported with tetragonal pore topology (pore volume = 0.305 cc g−1; BET surface area = 57.9 m2 g−1). The electron rich heteroatoms in the frameworks play a crucial role in adsorbing electron deficient iodine molecules and therefore, the benzothiazole-derived COF is envisioned to be potent material for efficient iodine adsorption. The synthesized COF showed the iodine adsorption capacity of 1.07 g g−1 in vapor phase and 109.0 mg g−1 from n-hexane solution. A reference COF (COF-4) with no thiazole group derived from 1,4-phenylenediamine and 2,4,6-triformylpholoroglucinol with hexagonal pore topology (pore volume = 0.937 cc g−1) and BET surface area 133.9 m2 g−1 showed adsorption of iodine from n-hexane solution with capacity 149.5 mg g−1. The study revealed that although the presence of heteroatoms in the framework facilitates the iodine adsorption by converting the molecular iodine into the polyiodides; the factors such as BET surface area, pore topology and pore volume also play a major role in the adsorption of iodine molecules. [ABSTRACT FROM AUTHOR]

This study offers a new simplified approach of diglycolamides synthesis, extraction agents used in separation procedures in various fields of nuclear chemistry such as waste management, spent nuclear fuel reprocessing, radioecology or radiopharmacy. The synthesis is assisted by coupling agent, propylphosphonic anhydride, commercially known as T3P®. Using this agent allows to perform the synthesis in a single step and to synthesize also novel types of diglycolamides that have not been described, yet. Thus, this approach helps to widen the spectrum of prepared diglycolamides and possible use of these extraction agents. Prepared diglycolamides insoluble in water were immobilized on polymer matrix and used as solid extractants in pilot batch experiments with Eu(III) ions in order to confirm the quality of known diglycolamides and asses the quality of novel ones. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Comprehensive Summary: Exploring the physiochemical properties and expanding the applications of actinide‐containing materials is paramount to address the escalating challenge of radioactive waste accumulation. However, unlocking the full potential of these materials is largely crippled by the radiotoxicity of the actinides. We report here two porous and luminescent thorium‐based metal‐organic frameworks (Th‐BITD‐1 and Th‐BITD‐2) that serve as a bifunctional platform for sequencing and sensing of radioiodine, a much more radioactive fission product discharged during the nuclear fuel reprocessing. In particular, Th‐BITD‐1 displays better iodine uptake performance than Th‐BITD‐2 via the solution‐based process and vapor diffusion with the maximum adsorption capacities of 831 and 1099 mg/g, respectively. Furthermore, Th‐BITD‐1 can function as a highly sensitive luminescence sensor for iodate with a quenching constant (KSV) of 6.6(5) × 103 M−1 and a detection limit of 2.02 μM, respectively, outperforming 2.96(6) × 103 M−1 and 10.5 μM of Th‐BITD‐2. Moreover, a positive correlation between the sensing efficacy and the iodate adsorption capacity has been revealed. This work highlights the opportunity in designing novel actinide‐based MOFs for their potential applications in radiological fields, e.g., radionuclide separation and detection. [ABSTRACT FROM AUTHOR]

Bituminized waste products (BWPs) were produced by conditioning in bitumen the co‐precipitation sludge resulting from industrial reprocessing of spent nuclear fuel. Underground geological disposal is a solution for the long‐term disposal of some intermediate level long‐lived (ILW‐LL) categorized BWPs in France. After one or several hundred thousand years, the water from the host rock will fully saturate the disposal cells. By an osmotic phenomenon enabled by the semi‐permeable capacity of the bituminous matrix, water in direct contact with BWPs could cause them to swell, which would lead to pressure on the host rock and may damage the disposal facility. Therefore, the BWPs' swelling behavior has to be taken into account in safety studies for disposal facilities. This paper presents an experimental and numerical investigation of BWPs' swelling behavior due to water uptake. Experimentally, water uptake, swelling and released salts were monitored for about 4 years during free leaching tests of simplified BWPs. The numerical model presented is extended from an existing one that incorporates transport mechanisms (diffusion, permeation, osmosis) with mechanics via Maxwell's viscoelastic model. An original coupled homogenization of transport terms is proposed to better capture the role of the semi‐permeable membrane played by the bitumen and the porosity dependency of coupled transport coefficients during water uptake. The presented numerical model is able to reproduce the experimental results of free leaching tests and shows its validity for the entire leaching process. [ABSTRACT FROM AUTHOR]

Deep eutectic solvents (DES) are considered a novel class of environmentally benign molecular solvents that are considered as potential solvents for nuclear fuel reprocessing, material recycling, and many other technological applications in both research and industry. However, there is a complete dearth of understanding pertaining to the behavior of metal ions in DES. Herein, we have investigated the speciation, complexation behavior, photochemistry, and redox properties and tried to obtain insight into the chemical aspects of the europium ion in DES (synthesized from heptyltriphenylphosphonium bromide and decanoic acid). The same has been probed using time-resolved photoluminescence (TRPL), cyclic voltammetry (CV), synchrotron-based extended X-ray absorption fine structure (EXAFS) spectroscopy, and density functional theory (DFT) calculations. TRPL indicated the stabilization of europium in the +3 oxidation state, favoring the potential of the Eu(III)–DES complex to emit red light under near UV excitation and the existence of inefficient energy transfer between DES and Eu3+. EXAFS analysis revealed the presence of Eu–O and Eu–Br, which represent the local surroundings of Eu3+ in the Eu(III)–DES complex. TRPL measurement has also suggested two distinct local environments of europium ions in the complex. DFT calculations supported the EXAFS findings, confirming that the Eu(III)–DES structure involves not only the oxygen atom of decanoic acid but also the oxygen atoms from the nitrate ions, contributing to the local coordination of Eu(III). Electrochemical studies demonstrated that the redox reaction of Eu(III)/Eu(II) in DES displays quasi-reversible behavior. The reaction rate was observed to increase with higher temperatures. The findings of this study can contribute to the understanding of the fundamental properties and potential applications of this luminescent and electrochemically active complex and pave the way for further studies and the development of novel materials with enhanced luminescent and electrochemical properties. [ABSTRACT FROM AUTHOR]

The applicability of bentonite clay-based ceramics for immobilization of waste from pyrochemical reprocessing of spent nuclear fuel–spent electrolyte, which consists of alkali metal chlorides, was studied. Main matrix characteristics such as the phase composition, mechanical durability, leaching rate and microstructure were studied. The results obtained were compared to the requirements to the cement compound, formulated in NP-019-15 document. The matrix was considered to be suitable for the radioactive waste disposal. [ABSTRACT FROM AUTHOR]

Alternative procedures for the dissolution of the solid phase (filter cake) of sludge formed by filtration of a spent nuclear fuel solution through cartridge and bulk filters were tested. Quite successful dissolution of the filter cake from the cartridge filter regeneration is ensured by successive treatment with alkali and acid solutions. Attempts to dissolve the filter cake from the bulk filter regeneration failed. The chemical and radiochemical composition of the filter cakes was evaluated. The results obtained will be used for assessing the nuclear and radiation safety and for developing procedures for the management of filter cakes from the filter regeneration. [ABSTRACT FROM AUTHOR]

This study aims at understanding the cause of vibrations observed during the operation of high speed centrifuges in a radioactive environment and materializing ways to mitigate these vibrations to ensure dynamic stability of centrifuges during operation. For this, a condition-based monitoring study was carried out for the centrifuge and its supporting structure using dynamic co-relation study in frequency domain during the operation of centrifuge at its operating speed of 19 700 rpm. Dynamic characteristics of the support structure were analyzed using standard finite element code. It was observed that high vibration amplitudes in the original support structure, before modification, owed to its low structural stiffness. Hence, structural modifications were carried out to reduce the dynamic crossover response of the support structure at the critical speeds and operating speeds of centrifuge. It was found that structural modification resulted in considerable reduction in the vibration amplitude of the support structure, compared to its initial configuration. [ABSTRACT FROM AUTHOR]

Conventional U/Th separation is carried out using tri-n-butyl phosphate (TBP) as the extractant. While processing the thorium-based feed solutions with TBP, third phase formation is a major concern due to the presence of large quantities of Th(IV). In a systematic investigation to overcome the limitations and retaining the properties of TBP, a branched isomer of higher homologue of TBP, tris(2-methylbutyl) phosphate (T2MBP) has been explored as a potential extractant for nuclear fuel reprocessing with lesser third phase formation tendency. In this context, T2MBP has been investigated for the complete recovery and separation of actinides (U, Th) from the THOREX and monazite ore processing feed solutions in cross current mode, and the results are compared with TBP-based systems. The co-extracted thorium in the separation process was further improved by scrubbing. The influence of the structural effect, organic-to-aqueous (O/A) phase ratio and scrub acid concentration on the U/Th separation process have been determined using T2MBP as the extractant. Overall, these studies indicated that T2MBP exhibits higher U/Th separation factor, and it has the ability to recover and separate individual actinides (U, Th) without third phase formation compared to TBP. [ABSTRACT FROM AUTHOR]

Rhodium is one of the scarcest, most valuable, and useful platinum group metals, a strategically important material relied on heavily by automotive and electronics industries. The limited finite natural sources of Rh and exponentially increasing demands on these supplies mean that new sources are being sought to stabilise supplies and prices. Spent nuclear fuel (SNF) contains a significant quantity of Rh, though methods to recover this are purely conceptual at this point, due to the differing chemistry between SNF reprocessing and the methods used to recycle natural Rh. During SNF reprocessing, Rh partitions between aqueous nitric acid streams, where its speciation is complex, and insoluble fission product waste streams. Various techniques have been investigated for Rh recovery during SNF reprocessing for over 50 years, including solvent extraction, ion exchange, precipitation, and electrochemical methods, with tuneable approaches such as impregnated composites and ionic liquids receiving the most attention recently, assisted by more the comprehensive understanding of Rh speciation in nitric acid developed recently. The quantitative recovery of Rh within the SNF reprocessing ecosystem has remained elusive thus far, and as such, this review discusses the recent developments within the field, and strategies that could be applied to maximise the recovery of Rh from SNF. [ABSTRACT FROM AUTHOR]

The immobilization of cesium elements in cubic zirconia ceramic matrix is essential for studying fission product release and spent fuel reprocessing. However, existing loading methods suffer from uneven distribution and low loading of cesium elements. This paper presents an improved impregnation adsorption method that uses wet gel microspheres with a three-dimensional porous gel network to absorb cesium nitrate efficiently. The wet gel microspheres were calcined at 400 °C. Yttrium-stabilized cubic zirconia microspheres with 17.8 wt% cesium elements loading and 0% cracking rate were obtained. The adsorption process and uniform distribution of cesium elements were investigated by FTIR, Raman, TG-DSC, SEM, and TEM. The results showed that the improved impregnation adsorption method could achieve uniform high cesium elements loading in cubic phase yttrium-stabilized zirconia microspheres, promising for nuclear fuel waste reprocessing. • A new approach successfully achieves the immobilization of cesium elements in cubic zirconia ceramic microspheres matrix. • The loading of cesium elements in cubic phase yttrium-stabilized zirconia ceramic microspheres matrix reaches 17.8 wt%. • Cesium elements are mainly present as cesium nitrate in cubic phase zirconia microspheres matrix after calcination at 400 °C. [ABSTRACT FROM AUTHOR]

The status of nuclear power plants in the world is analyzed. Trends in the development of nuclear power in several countries are identified. Many reactors were shut down when their lifetimes expired. Issues with reprocessing of spent nuclear fuel and minimization of its amount must be resolved. High- capacity (up to 70 wt%) porous polymer composite sorbents with ultrahigh molecular weight polyethylene as the matrix and montmorillonite and zeolite NaX-n fillers have been produced using gel technology. The sorption kinetics of 137Cs, 90Sr, 90Y by polymer-inorganic composite sorbents containing montmorillonite and zeolite NaX-n were studied. The sorption efficiency of these radionuclides was shown to be an order of magnitude greater than that of cation-exchanger KU-2-8, which allowed the quality of remediated radioactively contaminated water to be improved. [ABSTRACT FROM AUTHOR]

Slow strain rate tensile (SSRT) experiments were applied to comprehensively examine the strain rate susceptibility of the stress corrosion cracking behaviour of commercial Zr702 used in the spent nuclear fuel reprocessing environment. Results revealed that layers formed on Zr702 in boiling HNO3 solutions was composed of monoclinic ZrO2 (m-ZrO2) and tetragonal ZrO2 (t-ZrO2). With strain rates above 10−4 s−1, oxide layers ruptured rapidly, evidenced by the drop down of opened circuit potential (OCP). However, the OCP curve displayed fluctuation with 10−5 s−1 during straining, indicating that oxide layer alternately ruptured and repaired, resulting in the thickening of the oxide layer and increasement of stress corrosion cracking susceptibility. [ABSTRACT FROM AUTHOR]

A shrouded air sampling probe was designed and fabricated to meet the effluent monitoring requirements of an upcoming reprocessing facility at IGCAR. Performance evaluation of the probe was carried out using standard polydisperse aerosols up to 2 µm at the high-efficiency particulate air filter test facility. Its performance was also verified for higher size particles till 10 µm using the "Deposition 2001a" code. Based on the results, the shrouded probe was found to qualify the air sampling requirements and the same was fixed at the sampling point in the exhaust duct. This paper describes the details of design, testing, and installation of the shrouded probe. [ABSTRACT FROM AUTHOR]

Increased focus on carbon neutral energy has generated a resurgence of interest in nuclear power, and in particular advanced reactors which are likely to utilize high assay low enriched uranium (HALEU). This in turn could increase the economic attractiveness of recovering still partially enriched uranium from used nuclear fuel. Concomitant to development of advanced reactors, advanced reprocessing schemes should be developed which address the disadvantages to well established reprocessing schemes. The present study focuses on using di-(2-ethylhexyl)-iso-butyramide (DEHiBA) under high metal loading conditions for the reprocessing of used nuclear fuel. The elements examined in the study include the dominant transuranic actinides (Np, Pu, Am) as well as the often-problematic Tc and Zr. By increasing the concentration of the extractant from the more commonly reported 1.0 M – 1.5 M, the extraction of hexavalent actinides is substantially increased, while maintaining effective rejection of tri, tetra, and pentavalent actinides, particularly in the presence of high loadings of uranium. The extraction of Zr by 1.5 M DEHiBA is noted to be negligible by comparison to tributyl phosphate (TBP), however the coextraction of Tc with U is observed to be nominally twice the quantity that is extracted by TBP indicating a need for effective Tc management. [ABSTRACT FROM AUTHOR]

Lanthanides (Ce, Pr, Nd and Gd) 1, 3, 5 and 10 wt% of each were loaded into a pristine sodium aluminium phosphate glass matrix in order to simulate the immobilization of minor actinides (Am, Np, Cm) produced from spent nuclear fuel reprocessing. The optical parameters pertaining to these glasses were derived from the diffuse reflectance spectra. The systematic dependence of the physical and structural properties of these glasses on the dopants and the concentration of the latter was studied by XRD and Raman spectroscopy. The results on Rare Earth (RE) doped sodium aluminium phosphate are being reported for the first time. The physical parameters such as density, molar volume and oxygen packing density were evaluated. The XRD patterns showed the absence of any crystalline phase in these glasses. The relative % areas of (PO3)−2(Q2) bonds increase, whereas Q0 and Q1 tend to decrease with the addition of lanthanide oxides as compared to the pristine glass. The addition of different lanthanide oxides was found to depolymerize the network by forming more Q2 short chains. The results showed that the band gap energy (Eg) depends strongly on the cation field strength (CFS) of the lanthanides. The band gap energy decreases with an increase in the lanthanide concentration in the glass matrix. Praseodymium-doped glasses showed the highest Eg values (5.28–5.23 eV). The photoluminescence spectra confirm the characteristic excitation and emission peaks for the cations Ce+3, Pr+3 and Gd+3 and their variation with concentration was analysed. [ABSTRACT FROM AUTHOR]

Better ways are needed to capture radioactive volatile fission products (Kr, Xe, Br, I, Te, Rb, and Cs) discharged during the reprocessing of spent nuclear fuel in order to reduce the volumes of produced waste and minimize environmental impact. Using density functional theory, we examine the efficacy of a one-dimensional yttrium silicide electride (Y5Si3:e−) as a host matrix to encapsulate these species. Endoergic encapsulation energies calculated for Kr, Xe, Rb, and Cs imply they are not captured by Y5Si3:e−. Encapsulation is exoergic for Br, I, and Te with respect to their atoms and dimers as reference states, meaning that they can be captured effectively due to their high electronegativities. This is further supported by the formation of anions due to charge transfer between Y5Si3:e− and Br (I and Te). The selectivity of this material for these volatile species makes it promising for use in nuclear filters. [ABSTRACT FROM AUTHOR]

The separation of actinides from lanthanides in spent nuclear fuel reprocessing is a vital step of nuclear fuel cycle process. As one class of mature industrial extractants, the organophosphorus extractants have been widely used for the extraction and separation of actinides and lanthanides in spent fuel reprocessing due to their strong extraction ability and low‐cost acquisition. In this concept, the application scope of tributyl phosphate (TBP), bis(2‐ethylhexyl) phosphate (HDEHP), octyl(phenyl)‐N,N‐diisobutylcarbamoylmethylphosphine oxide (CMPO), trialkyl phosphine oxide (TRPO), and purified Cyanex 301 (bis(2,4,4‐trimethylpentyl) dithiophosphinic acid, HA301) are introduced, and their extraction mechanism, as well as structure‐function relationships for separation of actinides over lanthanides are also discussed. Furthermore, the design criteria, extraction properties and mechanism of several typical newly developed organophosphorus extractants (CMPO‐modified calixarene/pillarene, phenanthroline‐derived organophosphorus extractants, and phosphate‐modified carborane) based on pre‐organized skeletons are briefly reviewed. Finally, the important role played by those organophosphorus extractants is emphasized and potential applications in separation of actinides over lanthanides in future advanced nuclear fuel cycle are identified. [ABSTRACT FROM AUTHOR]

85Kr is a major radioactive byproduct released during nuclear fuel reprocessing and an important atmospheric indicator for nuclear safety and environmental monitoring. In this work, we report on online monitoring of atmospheric 85Kr activities with time resolution of 1.5 hours. An automated Kr purification system and an atom trap based 85Kr measurement system work synchronously in the experiment. The Kr purification system features temperature-controlled cryogenic distillation and gas chromatography, which allows direct sampling of ambient air and continuous operation. The separated Kr is fed into an atom trap capable of measuring the abundance of 85Kr with micro-L size Kr samples. The sample consumption is 3 L STP of air per measurement with a 1.5 hour processing time. Compared to the conventional monitoring method based on proportional counting, both the measurement time and sample size (10 m3 and 1 week) are reduced by several orders of magnitude. The 85Kr analysis technology demonstrated in this work allows analysis of atmospheric 85Kr activities with hourly resolution, enabling applications such as online monitoring of nuclear fuel reprocessing plants and tracing of atmospheric circulation patterns. [ABSTRACT FROM AUTHOR]

At the VNIINM (Rosatom) installation, the sorption chromatographic technique for separation of 244Сm and 241,243Аm was tested as part of the research work of the Proryv project in the field of fractionation using sulfonic cation exchange resin SPS SAC (8%)-50 (manufactured by Technosorbent LLC). The separation was carried out using a concentrate of REE–TPE oxides from the VVER-1000 spent nuclear fuel reprocessing at the Mayak Production Association. In summary, 1.39 g of Am with a purity of more than 99.9% was yielded in one cycle. The combined Cm fraction contained 36.4 mg of Am (about 2.6% of the original). The amount of Am in fractions containing also Eu or Cm did not exceed 4% of its initial amount. [ABSTRACT FROM AUTHOR]

Small modular reactors represent a promising technology for power generation, offering solutions to the energy crisis and mitigating greenhouse gas emissions. As Ukraine considers the deployment of NuScale, UK SMR, and SMR-160, it is crucial to address the safe management of spent nuclear fuel. This study focuses on evaluating the radiological characteristics of spent nuclear fuel from the selected small modular reactors and for comparison, from the VVER-1000 reactor. Using the Monte Carlo code Serpent, depletion calculations were performed for an assembly in an infinite 2-D geometry, and the activity, decay heat, and inhalation toxicity of the spent nuclear fuel were assessed. We determined the main nuclides contributing to the radiological characteristics and quantified the mass content of these nuclides. The total number of spent nuclear fuel assemblies produced during the entire life of each small modular reactor type was estimated. The radiological characteristics assessed for the three small modular reactors do not exceed those observed for WER-1000 reactors currently operating in Ukraine. So, spent nuclear fuel generated by the selected small modular reactors will introduce no new challenges to Ukraine's radioactive waste management system. The results of this work provide valuable insights for identifying the optimal small modular reactor technologies for Ukraine concerning safe spent nuclear fuel management. [ABSTRACT FROM AUTHOR]

We analyzed the spatiotemporal variations in the 137 Cs activity concentrations in global ocean surface seawater from 1956 to 2021 using the HAMGlobal2021 (Historical Artificial radioactivity database in Marine environment, Global integrated version 2021) and other published data. The global ocean was divided into 37 boxes. When observing the 0.5-year median value of 137 Cs in each box in the Pacific Ocean, we noticed that the values gradually increased or had almost constant levels in the 1950s and 1960s, and then decreased exponentially in 1970–2010, immediately before the Fukushima Daiichi Nuclear Power Plant Station (F1NPS) accident. In the northern North Atlantic Ocean and its marginal sea, the 0.5-year median values of 137 Cs showed large variations in the directly discharged 137 Cs from the reprocessing plants. The 137 Cs inventory in the surface mixed layer in 1970, when 137 Cs was released into the surface seawater, was estimated to be 184±26 PBq. In 1975 and 1980, the 137 Cs inventory increased to 201±27 and 214±11 PBq, respectively, due to direct discharge from the Sellafield and La Hague nuclear fuel reprocessing plants. In 2011, the 137 Cs inventory in the global ocean mixed layer increased to 50.7±7.3 PBq compared to that before the F1NPS accident, in which the contribution from the accident was estimated to be approximately 15.5±3.9 PBq. Mass balance analysis indicates that 137 Cs deposited by the global fallout in the western North Pacific Ocean moved to the eastern North Pacific Ocean. Subsequently, 137 Cs was transported southwards, followed by westward transport in the subtropical and equatorial Pacific Ocean, and then inflowed into the Indian Ocean via the Indonesian Archipelago. The longer apparent half-residence times in the Indonesian Archipelago (36.7 years from 1973 to 1997) and central Atlantic Ocean (38.0 years from 1992 to 2016) also support the interpretation of the global-scale transport of 137 Cs from the western North Pacific Ocean to the Indian (20–30 years) and Atlantic oceans (30–40 years). In the northern North Atlantic Ocean and its marginal sea, 137 Cs discharged from nuclear reprocessing plants is transported to the North Sea, Barents Sea and coast of Norway, and Arctic Ocean on a decadal scale. The dataset is available at 10.34355/CRiED.U.Tsukuba.00085 (Aoyama, 2021), 10.34355/Ki-net.KANAZAWA-U.00149 (Inomata and Aoyama, 2022a), 10.34355/Ki-net.KANAZAWA-U.00150 (Inomata and Aoyama, 2022b), and 10.34355/Ki-net.KANAZAWA-U.00151 (Inomata and Aoyama, 2022c). [ABSTRACT FROM AUTHOR]

The recovery of neptunium and plutonium play an important role in the spent nuclear fuel reprocessing process. Neptunium and plutonium are multiple oxidation states stable in solution and the varying conditions of the involved processes make it very difficult to effectively restrict them within desired streams in the relevant separation processes, and to precisely depict their species and ideally control their distribution in the environment. In order to control the valence of neptunium and plutonium in nitric acid and 30% TBP-nd, the structures and properties need to be deeply understood, especially the coordination behavior with nitrate. The special 5f orbital electron transition of actinides results in the characteristic absorption spectra of neptunium and plutonium in the visible-near infrared region, therefore, the valence and species structure of neptunium and plutonium in the relevant process points can be studied by absorption spectroscopy. In this paper, the species composition of neptunium in spent fuel reprocessing process, especially the distribution of Np (IV), (V), (VI) in the aqueous and organic phase in dissolution and co-decontamination unit were studied by absorption spectroscopy, the coordination structure of neptunium with nitrate and the spectra in 30% TBP- n-dodecane were also studied. On the other hand, the distribution of Pu (III), (IV), (VI) in the process were studied by absorption spectroscopy method. [ABSTRACT FROM AUTHOR]

A multisensor system for quantitative determination of minor actinides and other transuranic elements jointly with rare earth elements in multicomponent solutions formed in the course of spent nuclear fuel reprocessing and radioactive waste management was developed and tested. A cell providing the process control in real time was fabricated and tested. [ABSTRACT FROM AUTHOR]

Electroanalytical determination of chlorides in molten salts are often found to not have the accuracy needed for nuclear material monitoring. This was also observed in our studies on samarium trichloride. It was determined that a poorly soluble and relatively stable oxychloride that formed during our studies accounted for the loss of accuracy in the quantification of samarium trichloride. In this study, the spectroscopic and electrochemical properties of synthesized samarium oxychloride were investigated as they relate to the pyrochemical reprocessing of used nuclear fuel in the molten LiCl–KCl of eutectic composition. A qualitative, in situ investigation of oxychloride formation was conducted using electroanalytical voltammetry and Raman spectroscopy. These results were confirmed using synthesized oxychloride. The nature of oxychloride formation, solubility in the molten eutectic, implications in electrochemical processing, and the management of nuclear material in high-temperature systems are discussed. [ABSTRACT FROM AUTHOR]

Radiolytic stability is one of the main requirements of the ligands for the reprocessing of spent nuclear fuel. The prediction of radiation stability based only on the 2D molecular structural formula allows us to accelerate and simplify the development of new ligands. Here, we used quantum chemistry to investigate the radiolytic behavior of water-soluble diglycolamides as one of the most popular ligands for spent nuclear fuel reprocessing. The accurate accounting of conformational mobility in the descriptors based on the Frontier Orbital Fukui theory allowed us to obtain a good correlation between theoretical and experimental data. [ABSTRACT FROM AUTHOR]

Pulsed disc and doughnut columns (PDDC) have become an integral part of the nuclear fuel reprocessing and hydrometallurgical industries for quite some time. In this study, five unified correlations have been proposed for the prediction of slip velocity, characteristic velocity, flooding throughput and flooding slip velocity over a wide range of operating parameters, column dimensions and chemical systems in a PDDC. Before formulation of the correlations, a detailed cross validation of the existing correlations with the reported and the present experimental results have been performed. Then the empirical equations have been formulated by correlating the reported results by various researchers along with the present experimental data. The total number of data points used for predicting the slip velocity, characteristic velocity, flooding throughput and flooding slip velocity is 584, 129,185 and 128, respectively. For slip velocity and slip velocity at flood point, two new correlations have been proposed. For characteristic velocity and flooding throughput, the reported and the present data points have been refitted using existing correlations. Good accuracy has been achieved between the predicted and experimental results. [ABSTRACT FROM AUTHOR]

Complexing of organic ligands with actinides plays a key role in the processes of spent nuclear fuel reprocessing in the final stages of the nuclear fuel cycle. N,O-Donor ligands based on phenanthroline diamides are promising for group isolation and separation of actinides in hydrometallurgical approaches that are alternative to the PUREX process (such as the GANEX process). Understanding of the specific features of actinide ions complexing in an organic solvent may explain the mechanisms of extraction and be useful to find an "ideal" extractant for uranium extraction. Specific features of metal coordination in the presence of perchlorate anion have been studied by the example of a UO2(ClO4)2 complex with pyrrolidine-substituted 1,10-phenanthroline-diamide in an organic solvent, and the structure of the complex compound has been compared with that for a similar compound with nitrate anion as a counter ion. [ABSTRACT FROM AUTHOR]

Two new approaches to measure Np concentration in dissolved used nuclear fuel simulant (aqueous feed for PUREX process) by spectrophotometry are developed. The first approach is based on chemical reduction of Np in the feed to its tetravalent state using ascorbic acid with simultaneous conversion of Pu(IV) to Pu(III). Interfering effects from light absorbing fission and corrosion products are accounted for by measuring optical absorbance spectrum of aqueous raffinate after extraction of U, Np, and Pu by tributyl phosphate in dodecane. The second approach uses no chemical treatment at all and relies on spontaneous valency adjustment of Np to Np(V) by dilution of the feed with water to reduce its acidity to low decimolar range of nitric acid concentration. Results of Np determination in the feed by spectrophotometry are in good agreement with its concentration measured by ICP-MS. [ABSTRACT FROM AUTHOR]

Titanates and zirconates of light rare earth elements (REE): REE2TiO5, REE2Ti2O7, REE4Ti9O24, and REE2Zr2O7, are of interest as matrices for isolating the REE actinide fraction of high-level waste from the reprocessing of irradiated nuclear fuel. Data on the incorporation of impurities (Zr, U, Ca) into Nd and La titanates are examined. They display limited isomorphism toward these elements, including by the reaction 2REE3+ ↔ Ca2+ + U4+, which is common for minerals and their synthetic analogues. The reasons for the low solubility of Zr and U in Nd titanates and the role of the crystal chemical factor in the choice of crystalline matrices for the immobilization of the REE actinide fraction are considered. [ABSTRACT FROM AUTHOR]

We report a comprehensive study on lanthanide monouranate–based glass–ceramic (GC) composites as potential waste forms for the immobilization of the lanthanide actinide fraction waste arising from the reprocessing of spent nuclear fuel (SNF). Although the crystalline LnUO4 precursor prepared via a nitrate route can be well stabilized in a sodium aluminoborosilicate glass, the in situ crystallization of EuUO4 in glass from oxide precursors (Eu2O3 and U3O8) is rather robust with regards to various processing conditions such as waste loadings, ceramic‐to‐glass ratios, and cooling rates. Scanning electron microscopy and transmission electron microscopy investigations revealed the detailed microstructures, where ∼1–5‐μm spheres for NdUO4 and EuUO4, and ∼1–5‐μm rectangular crystals for DyUO4 and HoUO4 were observed in residual glasses. As designed, the pentavalent uranium has been confirmed by diffuse reflectance spectroscopy. Overall, LnUO4‐based GC composite waste forms are chemically durable, offering flexible processing options with wide operating windows for SNF and process waste stream management. [ABSTRACT FROM AUTHOR]

We analysed the spatiotemporal variations in the 137Cs activity concentrations in global ocean surface seawater from 1956 to 2021 using the HAMGlobal2021: Historical Artificial radioactivity database in Marine environment, Global integrated version 2021. The global ocean was divided into 37 boxes. The 0.5-yr average value of 137Cs in each box, except in the northern North Atlantic Ocean and its marginal sea, decreased exponentially in 1970-2010, immediately before the Fukushima Nuclear Power Plant (F1NPS) accident. The 137Cs inventory in the surface mixed layer in 1970 was estimated to be 187 ± 26 PBq. In 1975 and 1980, the 137Cs inventory increased to 201 ± 28 and 210 ± 12 PBq, respectively, due to direct discharge from the Sellafield and La Hague nuclear fuel reprocessing plants. In 2011, the 137Cs inventory in the global ocean mixed layer increased to 48.1±12.1 PBq compared to that before the F1NPS accident, in which the contribution from the accident was estimated to be approximately 15.5 ± 3.9 PBq. The distribution and variation in 137Cs in global surface seawater reflect basin-scale or global-scale transport. Mass balance analysis indicates that 137Cs deposited by the global fallout in the western North Pacific Ocean moves to the eastern North Pacific Ocean. Subsequently, 137Cs is transported southwards, followed by westwards transport in the subtropical and equatorial Pacific Ocean and inflow into the Indian Ocean via the Indonesian Archipelago. The longer apparent half residence times in the Indonesian Archipelago (36.7 years from 1973 to 1997), South Atlantic Ocean 37.0 years from 1973 to 2004), and Central Atlantic Ocean (43.5 years from 1993 to 2016) also support the interpretation of the global-scale transport of 137Cs from the western North Pacific Ocean to the Indian and Atlantic Oceans. In the northern North Atlantic Ocean and its marginal sea, 137Cs discharged from nuclear reprocessing plants is transported to the North Sea, Barents Sea and coast of Norway, and Arctic Ocean on a decadal scale. The dataset is available at http://dx.doi.org/10.34355/CRiED.U.Tsukuba.00085 (Aoyama, 2021), http://dx.doi.org/10.34355/Ki-net.KANAZAWAU. 00149 (Inomata and Aoyama, 2022a), http://dx.doi.org/10.34355/Ki-net.KANAZAWA-U.00150 (Inomata and Aoyama, 2022b), http://dx.doi.org/10.34355/Ki-net.KANAZAWA-U.00151 (Inomata and Aoyama, 2022c). [ABSTRACT FROM AUTHOR]

Hexagonal near-alpha titanium alloy-like Ti–6Ta (in wt%) is widely used as a structural material in spent nuclear fuel (SNF) reprocessing plant equipment. In this work, the detailed microstructure and texture of the Ti–6Ta alloy were analyzed by using EBSD technique, followed by the Gamma-ray (γ-ray) irradiation experiments with a 60Co isotope facility to simulate the γ-ray radioactivity of the short-lived fission fragments of SNF. Results showed that γ-ray irradiation-induced crystalline defects in hexagonal titanium was firstly observed and characterized. TEM results revealed the dislocation loops within α-titanium grains, whereas the loop size and number density peaked at the irradiation dose of 100 kGy, and then were obvious declined as the γ-ray radiation dose increased. The lattice parameters for the irradiated Ti–6Ta samples were slightly changed for the different irradiation dose cases. Moreover, Monte Carlo (MC) simulations further explained that the energy of primary knock-on atom (PKA) could reach the threshold energy of titanium or tantalum atom to induce the consequent vacancies and interstitials transformation into Frank dislocation loops in Ti–6Ta alloy. [ABSTRACT FROM AUTHOR]

One of the relevant tasks of nuclear power industry is the reprocessing of spent nuclear fuel. Such processing implies the separation of actinides from uranium fission products. One of the processing methods can be plasma mass separation. In the last 10 years, research aimed at the development of various aspects related to plasma mass separation has been actively conducted at the Joint Institute for High Temperatures of the Russian Academy of Sciences. The article provides an overview of the main results of these studies in four areas: numerical calculations and analysis of separation schemes; generation of background plasma and formation of plasma potential distribution; plasma source for injection of a mixture of separated substances; and demonstration of the model substances separation. [ABSTRACT FROM AUTHOR]

The danger of the wastes from the reprocessing of spent nuclear fuel is largely determined by the presence in them of long-lived actinides, first of all, americium and curium. The solution to the problem could be to involve together with plutonium its attendant actinides in the fuel cycle of fast reactors. This article considers the technologies (reactor, fuel fabrication, and spent fuel reprocessing) that are necessary for burning actinides as part of mixed uranium-plutonium nitride fuel. It is shown that the concentration of actinides, produced in a fast reactor, in the fuel will not exceed 0.5% h.a. For the transmutation of actinides separated from VVER spent fuel simultaneously with plutonium, it is expedient to increase the americium proportion in the fuel to 1.2%. Consideration is given to the possibility of using carbothermal synthesis of the initial powders and tablet technology for the manufacture of mixed uranium-plutonium nitride fuel, as well as the combined technology of spent-fuel reprocessing, planned for realization at a pilot-demonstration power complex with the BREST-300 reactor, and their suitability for involving neptunium and americium in the fuel cycle along with plutonium and uranium. [ABSTRACT FROM AUTHOR]

Radiochemical methods before measurement are indispensable to determine Iodine-129 by Accelerator Mass Spectrometry (AMS) in environmental samples. In particular, in sediments, iodine is associated with both inorganic and organic matter, and it presents in several chemical forms. Therefore, the radiochemical procedure must be designed to extract all the different iodine species from the matrix, whilst being as robust and time effective as possible. In this work, different microwave digestion methods, together with the processing of the iodine species to be measured by AMS, were tested and their performances are presented here. Two sediment cores, one with muddy sediment (core A) and one with permeable sandy sediment (core I), collected in the Celtic Sea, were used and the results were evaluated to determine an optimized iodine extraction. The method consisted of a microwave digestion with concentrated nitric acid followed by a double liquid–liquid extraction and the final precipitation of silver iodide (AgI), required to measure iodine by AMS. Back extraction of iodine in a reducing solution during purification was carried out with hydrazine (NH 2 -NH 2). Subsequent additional steps were added to optimize all iodine species extraction. First, hydrochloric hydroxylamine (NH 2 OH·HCl) and sodium bisulfite (Na 2 S 2 O 5) were used to homogenize all iodine species before extraction through a redox process. Second, sodium hypochlorite (NaOCl) was added to remove any final organic matter after digestion. The final method incorporated sodium hydroxide (NaOH) combined with NaOCl to facilitate reduction of iodine in solution and enhance the NaOCl effect. The different methods were validated against an internal standard and the last method was chosen, as it showed the most reproducible and accurate results. As practical application, Iodine-129 concentration from both cores were compared, showing concentrations between 0.19 × 1012 at/kg and 7.16 × 1012 at/kg for core A and between 0.28 × 1012 at/kg and 2.40 × 1012 at/kg for core I. Despite the 129I diffusion detected in the deeper layers, depth profiles accurately reproduced Sellafield discharges, which is the nuclear fuel reprocessing plant closest to the cores. [ABSTRACT FROM AUTHOR]

A new general synthetic route to selective actinide extracting ligands for spent nuclear fuel reprocessing has been established. The amide-functionalized ligands separate Am(III) and Cm(III) from the lanthanides with high selectivities and show rapid rates of metal extraction. The ligands retain the advantages of the analogous unfunctionalized ligands derived from camphorquinone, whilst also negating their main drawback; precipitate formation when in contact with nitric acid. These studies could enable the design of improved solvent extraction processes for closing the nuclear fuel cycle. [ABSTRACT FROM AUTHOR]

The reprocessing of spent nuclear fuel produces high level liquid waste (HLLW). Due to the decay heat, the concentrated nitric solutions containing fission products (FP) are stored in cooled tanks. Actually, the loss-of-cooling accident on these HLLW storage tanks could lead to releases of radioactive materials to the environment, especially ruthenium volatile species. After a general presentation on the physico-chemical behaviour of ruthenium in concentrated nitric acid, this paper successively details the different transfer modes of ruthenium reported in the literature, and the available trapping techniques. Based on this review, research perspectives are presented aiming at updating data concerning ruthenium behaviour in this context. [ABSTRACT FROM AUTHOR]

A passive neutron based assaying system for plutonium bearing solid wastes has been designed and developed. The assaying of the plutonium in the drum containing the wastes is determined by as positioned and by rotating the drum to 180 degree and average counts are obtained. The average counts were used for deriving the counting statistics. This system was calibrated with plutonium typically obtained in fast reactor fuel reprocessing facilities with known isotopic compositions. The minimum detection limit has been estimated to be 35 mg. It has been determined that a γ background of up to 2000 mR/h can be tolerated without loss of accuracy. [ABSTRACT FROM AUTHOR]

The physicochemical parameters of tris(2-methylbutyl) phosphate (T2MBP), a molecule possessing the essential characteristics required to be considered as a promising extractant in the solvent extraction stage of fast reactor fuel reprocessing, have been evaluated in the present study. In this context, the density, viscosity and interfacial tension (IFT), which are considered to be few among the important solvent selection criteria prior to its deployment in the plant scale have been measured for T2MBP in n-dodecane (n-DD) based systems. Similar studies have been performed with its straight chain isomer, tri-n-amyl phosphate (TAP) as well as the widely utilized commercial extractant, tri-n-butyl phosphate (TBP) so as to emphasize on the structural effects. Though the trends on the physicochemical behaviour of TBP based systems (temperature and ligand concentration effects) are well known in the literature, the data on TBP have been generated in the present work for comparison with the T2MBP results under similar experimental conditions. Solutions of trialkyl phosphates (TalP) in n-DD of different concentrations have been used to generate the data on the variation of density, IFT and viscosity with the change in the ligand concentration. In addition, the transformation in their properties has been assessed after subjecting the solvents to various levels of gamma absorbed doses. Overall, it has been observed that there was no significant variation in the density of the irradiated TalPs; however, there was a significant rise in the viscosity and reduction in the IFT of the TalP samples upon irradiation. The data on the Gibbs energy change of activation of various 1.1 M TalP/n-DD solutions have been generated by fitting their respective dynamic viscosity value as a function of temperature using Andrade's equation. Moreover, from the IFT value, the parameters relating to the interfacial activity of TalP/n-DD solutions have been determined using the Szyszkowski adsorption isotherm. [ABSTRACT FROM AUTHOR]

Spin-driven thermoelectric (STE) generation based on the combination of the spin Seebeck effect (SSE) and inverse spin Hall effect is an alternative to conventional semiconductor-based thermocouples as it is large scale, low cost, and environment-friendly. The STE device is thought to be radiation hard, making it attractive for space and nuclear technology applications. By using magnetometry, transmission electron microscopy, and the hard X-ray photoemission spectroscopy (HAXPES) measurements, we show that an STE device made of single-crystalline Y3Fe5O12 (YIG)/Pt heterostructures has tolerance to irradiation of high-energy heavy ion beams. We used 200 MeV gold ion beams modeling cumulative damages due to fission products emitted from the surface of spent nuclear fuels. By varying the dose level, we confirmed that the thermoelectric and magnetic properties of the single-crystalline YIG/Pt STE device are finite at the ion-irradiation dose up to 1012 ions/cm2 fluence. In addition, the HAXPES measurements were performed to understand the effects at the interface of YIG/Pt. The HAXPES data suggest that the chemical reaction regarding Fe and O that diminishes the SSE signals is promoted with the increase of the irradiation dose. The understandings of the damage analysis in YIG/Pt are beneficial for developing better STE devices applicable to harsh environmental usages. [ABSTRACT FROM AUTHOR]

As an alternative approach to the transmutation of Np, Am, and Cm in fast power reactors, the use of a specialized facility is considered, which includes a molten-salt burner reactor and module for fuel reprocessing and fabrication. Based on the requirements for the consumer properties of the installation, a comparative analysis of transmutation for LiF–NaF–KF and LiF–BeF2 salts was performed. Calculations of the transmutation characteristics were performed for model systems. The computational results show the advantages of a reactor based on the LiF–NaF–KF salt on account of the higher solubility of actinide fluorides. For a given transmutation capacity, a burner reactor based on LiF–NaF–KF consumes several-fold less power-grade plutonium, right up to its absence in the equilibrium regime, as well as lower thermal power and lower production of fission products. Of note is a particular aspect of the molten-salt incinerator reactor that is associated with the accumulation of even-numbered plutonium isotopes during its operation whereby the plutonium extracted from the facility is actually converted into long-lived radioactive waste. [ABSTRACT FROM AUTHOR]