Actinide Content Research Articles

Searching for incinerating the accumulated plutonium around the world by mixing it with thorium and using this mixture as a nuclear fuel in the CANDU-6

This work investigates the possibility of getting rid of the accumulated plutonium around the world by mixing it with fertile materials and using it as a nuclear fuel for CANDU-6. Using plutonium isotopes with thorium as a nuclear fuel for CANDU-6 is a means of preventing the production of other Pu isotopes. MCNPX has been used to design a three dimensional model of the CANDU-6 bundle. Three fuel types including (238U, rgPu)O2, (232Th, rgPu)O2 and (232Th, wgPu)O2 have been examined as nuclear fuel in the designed model and their results were compared with UO2 as a standard fuel. The fuel burnup parameters such as kinf, fissile inventory ratio, plutonium concentration and minor actinides concentration have been analyzed for the suggested fuels. Some of the most related safety parameters such as effective delayed neutrons (βeff), moderator temperature coefficient, coolant temperature coefficient and Doppler constant have been studied. The excess thermal neutrons in the CANDU-6 maximized the benefit of using plutonium-based fuels, where a significant amount of plutonium has been burned during the fuel cycle. From the neutronic and safety point of view, thorium fuel mixed with reactor grade plutonium has proven to be the most promising candidate among the investigated fuels.

Predictive phase stability of actinide-bearing hollandite waste forms from first-principles calculations

In this work, we investigated the stability trends of actinide-bearing hollandites using density functional theory calculations. Incorporating actinides generally decreased the stability of cesium end-members relative to their barium equivalents. Among the actinides, neptunium-containing structures were consistently the least stable. Curium was the most stable actinide in the +3 oxidation state, followed by americium, plutonium, uranium, and neptunium. Conversely, in the +4 state, uranium-bearing hollandites were the most stable. An inverse correlation between actinide content and stability was consistently observed.

Monitoring for Actinide Neutron Emissions From Spent Nuclear Fuel Under Extreme Gamma Fields Using Centrifugally Tensioned Metastable Fluid Detector Sensor Technology

Abstract This paper presents outcomes from research studies conducted independently at Purdue University as part of a collaborative project with the staff of Pacific Northwest National Laboratory (PNNL). Research work focused on assessments for meeting the challenge of monitoring actinide content in spent nuclear fuel (SNF) via characteristic neutron emissions [from spontaneous fission and (α,n) reactions] using the centrifugally tensioned metastable fluid detector (CTMFD) sensor technology using an inert organic decafluoropentane (DFP)(C5H2F10) sensing fluid. Traditional detectors readily saturate and/or cannot monitor neutron emissions under the expected 1012:1 gamma to neutron radiation environment. A challenge problem was posed to examine if a CTMFD could operate reliably over 1 h for conducting neutron spectroscopy at a 1 m standoff from a 30-y cooled SNF, in a ∼1012:1 gamma:neutron intensity environment resulting in a 150 Gy (15 kRad) accumulated dose for the CTMFD. The impacts on reliable operability were studied separately under expected gamma radiation energy and intensity for possible effects of: (i) radiolysis in the CTMFD sensing fluid from absorbed (<1.5 MeV) gamma radiation; (ii) photoneutron contamination signals from < 3 MeV high energy gamma photons interacting with the sensing fluid; and, (iii) malfunction of CTMFD component electronics from the absorbed gamma radiation. A Co-60 gamma irradiator was used for dose accumulation in the CTMFD electronic components and sensing fluids. A 14 MeV DT accelerator was used with a NaCl target to produce 3–4 MeV photons from activated 37S (via. neutron absorption in 37Cl) at SNF-commensurate intensities from SNF at 1-m standoff. Our examinations revealed the absence of any significant impact on CTMFD performance for meeting and exceeding the challenge problem metric. That is, we validated for no discernible impact of: 3–4 MeV gamma-produced photoneutrons when combined with a fission neutron source and radiolysis in the DFP sensing fluid through a 150 Gy absorbed dose. Past research results at Purdue University have validated for survivability of the key electronic components for absorbed gamma doses above the targeted 150 Gy level. This paper also provides extended evidence for survivability (from radiolysis) at higher gamma doses through 750 Gy with a borated DFP-sensing fluid formulation-based CTMFD.

Maximizing the Processing of Polymetallic Concentrates via Actinide Separation and Rare Earth Retrieval

During the last decades, the growing demand for rare earth elements (REEs) has led to numerous recent studies to recover these elements from various bearing ores and wastes. Therefore, the recovery of REEs from Ras Baroud polymetallic concentrate has been investigated in the current study. Physical beneficiation for the Ras Baroud pegmatite sample was carried out, yielding a concentrate for euxenite (Y), fergusonite (Y), xenotime (Y), monazite (Ce), allanite, thorite, uranothorite, and Hf-zircon, which resulted in raising the concentrations of rare earth elements, Th, Zr, U, and Ti in the sample. Fusion digestion processes with sodium hydroxide were studied using the Conceived Predictive Diagonal (CPD) technique. The three experimental digestion groups proved the dissolution of 99.9, 95.6, 99.9, 52.5, and 0.47% for REEs, Th, U, Ti, and Zr, respectively, under fusion conditions of 723 K, 120 min, 1/1.5 ore/alkali ratio, and − 100-μm particle sizes. Fusion kinetics, isotherms, and thermodynamics were investigated using several suggested models, namely, pseudo reversible first order, uptake general model, and shrinking core model which matched well with the experimental digestion results. Selective recovery of actinide content from REE content of the digested concentrate chloride solutions was accomplished using solvent extraction with di-2-ethyl hexyl phosphoric acid. About 99.9, 99.9, and 4.2% extraction efficiencies for Th, U, and REEs were performed, respectively, using 0.3 mol/L solvent concentration in kerosene as a diluent, 1/2 organic to aqueous ratio, an aqueous pH of 0.2, and 15-min contact time. Thorium and uranium ions were stripped with sulfuric acid solution 2.5 mol/L with 94 and 98% stripping efficiency, respectively. A highly purified REE precipitate was obtained from the raffinate solutions. Zircon mineralization tailings were obtained as a by-product through the alkaline digestion process.

Intrinsic background of EJ-309 liquid scintillator detectors

In deep-underground experiments, intrinsic detector background can exceed the external one, limiting the detection sensitivity. Intrinsic background in liquid scintillator neutron detectors consists of α particles emitted from radioactive impurities in the detector housing (and possibly in the liquid). We present the results of long background measurements of 12 EJ-309 liquid organic scintillators detectors in the deep-underground Gran Sasso National Laboratory of the Italian Institute of Nuclear Physics. These detectors are a part of the detection array for the SHADES ERC project that aims to measure the very low cross section (down to picobarn) of the astrophysically important reaction 22Ne(α,n)25Mg. With an exposure of ∼35 days we identify an actinide concentration in the detector housing of 8.4(1.8)sta(1.4)sys×10−2ppm and 1.62(0.57)sta(0.03)sys×10−1ppm of 238U and 232Th, resulting in 64 α particle counts per hour. The half-life of 214Po was also determined as 163.7(11)μs.

Discussing the possibility of using thorium-based fuels as an alternative fuel to uranium dioxide fuel for APR-1400 reactor

This work investigates the possibility of using thorium-based fuels as an alternative fuel for advanced power reactor APR-1400. MCNPX code version 2.7 with cross-section library ENDF.VII has been used to design an APR-1400 fuel assembly. This code has been used to study the neutronic performance of the proposed thorium-based Fuel types (0.944Th, U)O2, (0.955Th, 233U)O2, (0.934Th, rgPu)O2. The fuel burn-up parameters such as infinity multiplication factor (kinf), initial heavy metals concentrations, Minor actinides concentration and fission products concentration have been analyzed during 1500 effective full power days (FPDSs) for thorium-based Fuel types and compared with the common fuel. The analysis of the horizontal thermal power and neutron flux distribution provides valuable insights into the behavior and performance of the suggested Fuel types in the APR-1400 assembly. The analysis of the neutronic results ensures the viability of using the proposed thorium-based fuel types as an alternative fuel to UO2 because they achieved acceptable safety parameter values and provided a good power distribution through the fuel assembly compared to UO2.

Investigation of potential polyatomic interferences on uranium isotope ratio measurements for the LS-APGD-Orbitrap MS system

The determination of actinide (e.g., U and Pu) content and isotopics is of importance to the nuclear forensics and safeguards communities. However, in the analysis of environmental samples, such as those collected by the International Atomic Energy Agency, uranium measurements can be complicated by isobaric interferences from polyatomic variants of heavy elements (e.g., Pb). This leads to complex, time-consuming sample manipulations (i.e., separations) before determining isotope ratios. An alternative strategy to sample pretreatment is to use high-resolution mass spectrometric platforms during the analysis to fully resolve the uranium isotopes from potential polyatomic interferences, negating the need for prior chemical separation. The liquid sampling - atmospheric pressure glow discharge (LS-APGD) coupled with an Orbitrap mass spectrometer provides a high-resolution (>70,000 m/Δm at m/z 200) inorganic mass spectrometry platform. As a demonstration of the power of this instrumental platform, and indeed, the high-resolution approach in general, uranium isotope ratios were determined in the presence of elemental impurities (e.g., Pb, Pt, Ta, W) commonly encountered with environmental sample swipe analysis, without any prior treatment. Even at elemental impurity concentrations of 1000–5000× relative to uranium, no interference was observed with the 235U or 238U signal. In addition, the 235U/238U isotope ratio for the samples with the concomitants present are within 2 standard deviations of the values obtained without their addition, indicating that these impurities do not impact the determined uranium isotope ratio. These findings represent a significant first step in leveraging the high resolution of the LS-APGD-Orbitrap-MS to overcome isobaric and molecular interferences instead of relying on chemical separations.

New Waste Management Options Created by iMAGINE through Direct Operation on Spent Nuclear Fuel Feed

The demand for improving the nuclear waste management has since long been identified as one of the major hurdles for widespread use of nuclear energy. Nuclear waste management, through partitioning and transmutation (P&T), has been researched since the 1990s with partitioning being a prerequisite for the process. Recently, an innovative approach of reactors directly operating on spent, or nowadays often called used nuclear fuel, iMAGINE has been proposed which could deliver on the aims of P&T as a side effect to more efficient and sustainable nuclear energy production in the future. A HELIOS model of the core has been used to analyze the long-term operation of a molten salt reactor including the investigation of the minor actinide accumulation over the entire burnup period. The results shown here confirm that long-term reactor operation is possible, even with higher amounts of vitrified waste loaded. Thus, it is possible to achieve the aims of P&T without prior partitioning, but it is certainly less efficient since the high concentration of minor actinides (MAs), required for efficient burning, is impossible to obtain in a short operational time. On this basis, the proposed nuclear waste management approach will be a long-term effort when it is accomplished without partitioning/separation technologies. However, none of the analyses contradicts this effort. The key points are: (a) when the technology for treating the waste is possible and reliable, the time horizon will not be a major concern; (b) the waste management is now intrinsically linked with energy production instead of requiring dedicated costly facilities, delivering a promising economic basis; (c) the waste management is now associated with long-term energy production and massively improved resource utilization. The study of feedback effects has shown that the modeled system has a strong negative feedback effect of ~−6 pcm/K, and even with spent nuclear fuel feed reduces to ~−3.8 pcm/K, ensuring the basis for a safe operation. In summary, it has been demonstrated that the objectives of P&T are achievable without prior partitioning, an approach which was never even discussed in the past. These ground-breaking results and the new insights will allow or even require rethinking the nuclear waste management of the future.

Accurate prediction of gaseous isotopes Xe and Kr concentrations from the burnup of nuclear fuel using simple regression algorithm.

Mathematical techniques for modeling and simulating dangerous or complex systems, such as nuclear technology systems, often require high-performance computing to process and analyze available data. In this paper, simple and quick method to support studies and research related to nuclear fuel is presented. This reasonably simple method helps to predict different concentrations of actinides and fission products in nuclear fuels without the need for expensive specialized programs and highly-trained researchers. The great importance of this approach is the speed of predicting the components of nuclear fuel concentrations, which in turn leads to quick decision-making, such as the possibility of operating fuel at higher burnup values, predicting the amount of gases resulting from nuclear fission (which may accumulate and cause problems in nuclear fuel such as volume swells), and other important decisions in nuclear fuel technology. The predicted equations have been generalized for higher values of burnup and compared with comparable results from MCNP codes. The equations deduced in calculating the different concentrations of xenon and krypton isotopes resulting from fission in burnup of nuclear fuel showed very precise results with discrepancies (magnitude of an error between the data points and the corresponding predicted ones) less than 2%. The suggested method offers a great advantage for researchers, which are the use one of any simple or common computational programs available to most researchers and do not need much experience to deal with, such as MATLAB, Excel that are easy to use for regression analyses. In this paper, the advantages of the proposed method are explained along with the limitations of its use.

Assessment of the Concentration of 51 Elements in the Liver and in Various Parts of the Human Brain-Profiling of the Mineral Status.

The anthropogenic environment and diet introduce many metals into the human body, both essential and toxic. Absorption leads to systemic exposure and accumulation in body fluids and tissues. Both excess and deficiency of trace elements are health hazards. The primary aim of the present study was to evaluate the concentration of 51 elements in liver samples and 11 selected brain regions obtained at post-mortem examination from a population of adults living in south-eastern Poland (n = 15). A total of 180 analyses were performed by inductively coupled plasma mass spectrometry in two independent replicates. The collected data show very high individual variability in the content of the investigated elements. Macroelements such as sodium, magnesium, phosphorus, potassium, calcium, iron, and zinc occurred in the highest concentrations and with the greatest statistically significant variations. Although the elemental content of the brain and liver differed significantly, the strongest positive correlation between liver and polus frontalis was observed for the essential element selenium (0.9338) and the strongest negative one for manganese (-0.4316) and lanthanum (-0.5110). The brain areas studied have different requirements for phosphorus, manganese, iron, and molybdenum. In addition, males had a significantly (p < 0.05) higher brain content of lanthanides and actinides than females. The results of this study show that the inhabitants of south-eastern Poland are exposed to a fairly uniform accumulation of aluminum and vanadium in the brain, which have the highest affinity to the thalamus dorsalis. This result proves that there is environmental exposure to these elements.

Radioactive Fission Waste of The Conversion of High-Enriched Uranium to Low-Enriched Uranium Target on 99Mo Production

In the past, 99mTc produced by PT. Industri Nuklir Indonesia (PT. INUKI) was carried out through the processing of a 99Mo parent radioisotope that was produced using high enriched uranium target. Due to restrictions on the use of high-enriched uranium (HEU) by the International Atomic Energy Agency, PT. INUKI has converted the process using a low enriched uranium (LEU) target. This study aims to determine the change of radioactive fission waste characteristics in 99Mo production due to the conversion process. The characteristics of radioactive fission waste generated from low-enriched uranium targets were calculated using ORIGEN 2.1 program. The results show that the radioactive fission waste contains radionuclides of activation products, actinides, and fission products. The conversion produced smaller activity of 99Mo compared to that of using the HEU target, which was 397 compared to 1010 Ci. The conversion generated radioactive fission waste with a smaller content of remaining 235U but greater content of actinides, particularly 239Pu. The activity of 239Pu in the radioactive fission waste from LEU was 29.1 μCi, approximately 19 times higher compared to that of using HEU, which was 1.52 μCi. After 50 years of decay, this radioactive fission waste was calculated to have a specific activity of 6.54x108 Bq/g, lower than that of the highly enriched uranium target, which was 3.01x109 Bq/g. This radioactive fission waste requires management with a high safety level.

Trace Actinide Signatures of a Bulk Neptunium Sample.

In nuclear forensic analyses, measurements of actinide elements in a sample can assist with identifying interdicted or unknown materials. While these radiochemical signatures have been extensively investigated in uranium materials, less is known about bulk neptunium samples. This paper describes the measurement of trace actinide concentrations and isotopic profiles in a 237Np oxide sample. Uranium, plutonium, americium, and curium concentrations and isotopic profiles in the sample were determined and deemed potentially useful for distinguishing different sources of 237Np. Several different potential radiochronometry systems were also investigated; discordant results indicate that the Np sample was never completely purified of other actinide elements, or that subsequent contamination of the sample occurred. Few prior studies of neptunium materials have been reported, and these data suggest that trace actinide constituents could provide unique signatures to identify material out of regulatory control.

Utilizing deep eutectic solvent for facile, selective and sustainable sequestration of plutonium

In the recent surge of green and cheap energy for future generation; nuclear industry is going to play a pivotal role and reprocessing of nuclear fuel through safe and systematic approach is the key. In that direction recent literature has shown that there is lack of better and improved methods towards efficient extraction of highly radiotoxic and hazardous plutonium from nuclear waste. Herein, we have utilized a greener and cost effective indigenously synthesized deep eutectic solvent (DES) for the selective extraction of plutonium from (Am, Pu, U) waste. The maximum D value achieved for Pu was 67.42 at 5 mol L−1 HNO3 media and the equilibrium uptake capacity was found to be 107.28 ± 5 μg of Pu/g of C12DES. This was followed by solidified floating organic drop micro-extraction (SFODME) technique for its quantification. High radiation stability of this DES (upto 750 kGy) highlighted its great potential in nuclear fuel reprocessing. Several other reprocess parameters were studied and optimized such as DES volume, alkyl chain length attached to hydrogen bond acceptor (HBA), actinide concentration and speciation and its acidity to achieve best extraction of plutonium ion form the waste stream. Regeneration capability and reusability of the DES employing stripping agents gave encouraging results as 95% of Pu could be stripped in a single cycle using 1 M hydroxylamine hydrochloride as the stripping agent. The developed method was applied for recovery and quantification of Pu from real samples, viz. (U,Pu) Carbide fuel solution, radioactive research waste and synthetic dissolver solution. We believe this work is a way forward in resolving a variety of issues that exist in health physics and reprocessing laboratories of nuclear installation pertaining to selective and efficient extraction of Pu (IV).

Validation and application of the Dragon5 lattice code for neutronics and burnup analysis of VVER-1000 pin cell and assembly model

This work deals with the validation and assessment of neutronics and burnup analysis capacity of the deterministic lattice code Dragon5 for VVER-1000 hexagonal pin and assembly model configuration. Various VVER-1000 hexagonal pin-cell benchmark cases including the fresh uranium fuel and mixed oxide fuel of ORNL are simulated and compared with the HELIOS and OpenMC code. The data library differences of certain nuclides will lead to large deviations of the results. Accurate results can be presented by the Dragon5 lattice code when compared with OpenMC code. The OECD LEU and MOX assembly benchmarks are computed to further validate the prediction ability for VVER-1000 assembly, reasonable agreement between Dragon5 solutions and published values of other codes is observed for burnup parameters including infinite multiplication factors, concentrations of important actinides and fission product nuclides. Moreover, the post-irradiation of UO2 sample experiment of the Russian Kalinin-1 VVER-1000 reactor is simulated and compared with laboratory measured values to provide the experiment validation of the Dragon5 lattice code. The predicted concentrations of actinides up to Cm244 are compared with HELIOS and Serpent Monte Carlo code. The results show that the Dragon5 lattice code presents better prediction accuracy for Calculated-to-Experimental ratios (C/E's) of most actinide concentrations than the HELIOS and Serpent code. Therefore, the École Polytechnique de Montréal Dragon5 lattice physics code can be applied to the neutronics and burnup analysis of VVER-1000 hexagonal pin-cell and assembly model.

Effects of base salt additives on NaCl–UCl3–PuCl3 fuel systems: Insights from CALPHAD simulations

The overall performance of molten salt reactor (MSR) fuels is governed by their chemical composition: the actinide content regulates fuel criticality; the base salt determines the fuel melting temperature, and the solubility of chemical species that evolves during the reactor’s lifetime. In this study, Calculation of Phase Diagrams (CALPHAD) simulations were performed to investigate the effects of non-unary base salts on the performance of NaCl–UCl3–PuCl3 fuel systems. We specifically focused on the fuel liquidus temperature and the fuel response to compositional changes, such as the breeding of PuCl3 and the evolution of fission products. Our calculations indicate that non-unary base salt combinations can effectively decrease the liquidus temperatures of the proposed fuel systems up to the eutectic actinide composition of the KCl–UCl3–PuCl3 system. Moreover, we found that the response of the fuel liquidus to shifts in fuel chemistry depends largely on the interactions between the base salt and actinide content. Our results imply that an appropriate modification of the base salt chemistry in the fuel can bolster the safety, performance, and economics of molten salt reactors.

Electrochemical Deposition with Redox Replacement of Lanthanum with Uranium in Molten LiCl-KCl

Electrochemical recovery of dilute concentrations of actinides from spent nuclear fuel would reduce the longevity of storing high-level nuclear waste. Electrochemical deposition with redox replacement (EDRR) is used in a molten salt medium for the selective electrochemical recovery of uranium in the presence of excess concentrations of lanthanum. In each EDRR cycle, after a short electrodeposition pulse, the deposited lanthanum is spontaneously replaced by uranium at open circuit. After repeated cycles, uranium metal was obtained on a tungsten electrode immersed in a LiCl-KCl melt that contained 1 wt% lanthanum chloride—0.15 wt% uranium (IV) chloride. Scanning electron microscopy and energy dispersive X-ray spectroscopy (SEM-EDS) analysis revealed uranium particles approximately 0.5−1 μm with well-defined rectangular shapes; and with 20–60 times more uranium recovered on the surface of the electrode than lanthanum.

Simultaneous quantification of Zr, Mo, U, Np and Pu in technological solutions of spent nuclear fuel reprocessing with a potentiometric multisensor system

The chemical analysis of spent nuclear fuel (SNF) reprocessing is a very challenging research field due to the high radioactivity of process streams and their complex composition. The process managing requires thorough chemical analysis to ensure safe and efficient performance. Traditional ways of chemical control of SNF reprocessing are based on sampling methods and involve long and tedious procedures, failing to provide immediate information on the process status. Recent literature suggests that potentiometric multisensor systems can be employed for on-line control. However, up to now these systems were only studied in very simple model media. This work reports on the potentiometric multisensor array applied for the first time for the simultaneous quantification of the content of actinides (uranium, plutonium, neptunium) and several other analytes (zirconium, molybdenum and nitric acid) in the complex samples obtained from the pilot extraction unit for SNF reprocessing. The same samples were also analyzed by UV-Vis spectrometry (another prospective method for on-line control) and the analytical performance of these two methods was compared using established analytical figures of merit (sensitivity and analytical sensitivity metrics) for multivariate calibration. In spite of the extremely challenging analytical task, both simple methods provide the characteristics suitable for real industrial applications.

Synthesis and feasibility studies of doping U at Ti site of Y2Ti2O7 as a radioactive waste immobilization matrix.

In pursuit of clean and green nuclear energy one of the major challenges is to effectively immobilize the nuclear waste. In this context A2B2O7 type pyrochlore owing to its structural flexibility, ability to accommodate ions at both A/B-sites and high radiation tolerance has demonstrated excellent capability to store highly radioactive actinide ions. To fill the major gap area of actinide doping at the B site we have taken up the challenge of doping uranium ions at the Ti site of Y2Ti2O7 type pyrochlore. An yttria titanate (Y2Ti2-xUxO7; x = 0.05, 0.075, 0.1, 0.2, and 0.3) based matrix with uranium doped at the Ti site was synthesized using a simple gel combustion route under an air atmosphere. Rietveld refined X-ray diffraction (XRD) demonstrated that Y2Ti2O7 can accommodate U up to 5 mol% in the Ti site without any phase separation, which was further confirmed using Raman spectroscopy. Y2Ti2O7 based matrices are found to be radiation stable up to 1000 kGy and at the same time they are moderately thermally stable and on a par with the values reported for pyrochlores. Uranium in Y2Ti2O7 stabilizes in +6 oxidation state in the form of uranyl ion distributed near and far off from titanium vacancies with distinct excited state lifetime. This work could provide a smart and strategic way for selecting a suitable advanced ceramic matrix for immobilization of high level waste with additional and important information on solubility limit, actinide speciation, radiation/thermal stability, actinide concentration, etc.

Actinides Th and U in Atmospheric Particulate Matter in Yakutsk

The distribution of Th and U was studied in the near-ground atmosphere of the city of Yakutsk. The chemical compositions of particulate matter (PM) in the summer atmosphere and in the solid and soluble phases of the snowpack were studied in 2019–2020. In the summer atmosphere, Th and U are concentrated mostly in the dust fraction of PM, together with a group of siderophile elements. The wintertime PM is characterized by a more homogeneous size of its particles. PM minerals with which concentrations of actinides correlate are amphiboles and pyroxene for Th and carbonates for U. The Th and U concentrations were found out to systematically decrease in the sequence soil–summer PM (dust)–winter PM (snow), at the dominance of Th, but the aqueous phase of the snow contains U concentrations one order of magnitude higher than those of Th. The bulk (close to 90%) of actinides is deposited during the warm season. The total amount of actinides falling out in Yakutsk is 63 mg/(m2 year), with two-thirds of this amount being Th. This value may locally increase to 200 mg/(m2 year) and more in areas of contrasting anomalies, with the predominance of Th. The main sources of actinides entering the near-ground atmosphere of Yakutsk are dusting from the soil surface and emissions from vehicles, with lesser amounts coming from energy-producing and construction-industry facilities.

High-Throughput Determination of Ultratrace Actinides in Urine by In-Line Extraction Chromatography Combined with Quadrupole Inductively Coupled Plasma Mass Spectrometry (EC-ICP-MS).

Determining actinides in urine is vital for occupational exposure monitoring and radiological emergency response because of the toxicity and radiological dose effects of actinides on human health. Traditional radiochemistry analytical methods used to determine actinide concentrations in urine are time-consuming (sample analysis takes several days) and are hindered by a variety of technical and instrumentation-related obstacles. A high-throughput, fully automated, precise, and accurate in-line method was developed for determining five actinides (241Am, 239Pu, 237Np, 232Th, and 238U) at ng/L levels in urine using extraction chromatography combined with quadrupole inductively coupled plasma mass spectrometry (EC-ICP-MS). In this method, the five actinides were successfully separated with the required sensitivity, peak shape, and resolution using a simplified single Eichrom TRU column with a Dionex ICS-5000 system. The separated actinides were subsequently injected into an in-line PerkinElmer (PE) NexION 300D ICP-MS for quantitative determination. The sample-to-sample run time was 23 min for automatic chemical separation and quantification using only 0.5 mL of urine. The limits of detection (LOD) obtained using this method were 0.015, 0.022, 0.039, 4.5, and 2.4 ng/L for 241Am, 239Pu, 237Np, 232Th, and 238U, respectively. The method routinely had a chemical yield of >84% as well as a linearity (R2) coefficient of ≥0.999 for the calibrators. The method proved to be rapid, reliable, and effective for actinide quantification in urine and therefore is appropriate for radiological emergency response incidents.