Comparison of direct kinetic phosphorescence analysis and recovery corrected kinetic phosphorescence analysis for the determination of natural uranium in human tissues

Summary In the analysis of biological samples with sub ng/g uranium concentrations, pre-concentration has been shown to improve the detection limit for the determination of uranium. Recovery corrected kinetic phosphorescence analysis (KPA) combines pre-concentration and separation of uranium by anion-exchange from human tissues dissolved in 6M HCl, with the radiochemical yield determined by alpha-spectrometry, using ²³²U as a tracer. Total uranium is determined by KPA after correction for chemical recovery. Twenty-one randomly selected dissolved tissue samples from the United States Transuranium and Uranium Registries (USTUR) Case 0242 were chosen for comparative analyses. The set of samples included dissolved bone and soft tissues. Uranium concentrations for seven of the samples had not been previously reported. Direct KPA could not be used to determine uranium concentrations of five unreported tissues. Three of these tissues had uranium concentrations at or below the KPA L_Q value of 0.028 ng/ml and two tissues had known matrix interferences. All seven of the unreported tissues were successfully analyzed by recovery corrected KPA; concentrations ranged from 9 to 1380 ng per tissue, including those that could not be analyzed by direct KPA due to matrix problems. Recovery corrected KPA gives results similar to direct KPA where matrix interferences and low detection limits are not encountered. A comparison of the direct method of KPA versus recovery corrected KPA shows marked improvement for the determination of uranium in samples that heretofore either uranium was not detected or the sample had to be drastically diluted to minimize matrix effects in order to measure uranium.     

Characterization and application of SuperLig<Superscript>®</Superscript> 620 solid phase extraction resin for automated process monitoring of <Superscript>90</Superscript>Sr

Characterization of SuperLig^® 620 solid phase extraction resin was performed in order to develop an automated on-line process monitor for ⁹⁰Sr. The main focus was on strontium separation from barium, with the goal of developing an automated separation process for ⁹⁰Sr in high-level wastes. High-level waste contains significant ¹³⁷Cs activity, of which ^137mBa is of great concern as an interference to the quantification of strontium. In addition barium, yttrium and plutonium were studied as potential interferences to strontium uptake and detection. A number of complexants were studied in a series of batch K_d experiments, as SuperLig^® 620 was not previously known to elute strontium in typical mineral acids. The optimal separation was found using a 2 M nitric acid load solution with a strontium elution step of ~0.49 M ammonium citrate and a barium elution step of ~1.8 M ammonium citrate. ⁹⁰Sr quantification of Hanford high-level tank waste was performed on a sequential injection analysis microfluidics system coupled to a flow-cell detector. The results of the on-line procedure are compared to standard radiochemical techniques in this paper.     

Implementation of alpha-spectrometry for uranium isotopes in excreta samples

Summary The measurement of radioactivity concentrations in excreta is an important tool for the monitoring of possible radionuclide intakes by occupationally exposed workers. For this purpose, a radiochemical procedure for the determination of alpha-emitting isotopes of uranium in excreta has been optimized. The main steps involved in this procedure are pre-concentration, dissolution of sample, separation by ion-exchange resin, electrodeposition and alpha-spectroscopy. ²³²U tracer is used to monitor chemical recoveries and correct the results to improve precision and accuracy. The quality control of radiochemical analysis in urine and faecal samples has been performed with participation in intercomparison exercises. The results obtained from these samples, with chemical recoveries (80-95%), are shown to be highly consistent. The method offers good prospects to be applied in routine monitoring programme of workers.     

Systematic radiochemical separation for the determination of <Superscript>99</Superscript>Tc, <Superscript>90</Superscript>Sr, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe and <Superscript>59,63</Superscript>Ni in low and intermediate radioactive waste samples

A simple and rapid separation procedure was systemized for the determination of ⁹⁹Tc, ⁹⁰Sr, ⁹⁴Nb, ⁵⁵Fe and ^59,63Ni in low and intermediate level radioactive wastes. The integrated procedure involves precipitation, anion exchange and extraction chromatography for the separation and purification of individual radionuclide from sample matrix elements and from other radionuclides. After separating Re (as a surrogate of ⁹⁹Tc) on an anion change resin column, Sr, Nb, Fe and Ni were sequentially separated as follows; Sr was separated as Sr (Ca-oxalate) co-precipitates from Nb, Fe and Ni followed by purification using Sr-Spec extraction chromatographic resin. Nb was separated from Fe and Ni by anion exchange chromatography. Fe was separated from Ni by anion exchange chromatography. Ni was separated as Ni-dimethylglyoxime precipitates after the removal of ^134,137Cs and ^110mAg by Cs-phosphotungstate and AgCl precipitation, respectively. Finally, the radionuclide sources were prepared by precipitation for their radioactivity measurements. The reliability of the procedure was evaluated by measuring the recovery of chemical carriers added to a synthetic radioactive waste solution.     

Determination of 90Sr in soil, grass and cereals

⁹⁰Sr was measured in environmental samples in Upper Austria in the year 2005. After the nuclear weapon tests the average deposition of ⁹⁰Sr in Austria amounted to 3.3 kBq/m². In 1986 the average deposition was 0.9 kBq/m² [1]. To assess the actual condition in soil, grass and cereals ⁹⁰Sr was measured in these samples. For all samples oxalate precipitation was conducted and strontium specific columns (Eichrom Industries, Inc.) were used. The calcium concentration in these samples was determined to estimate the amount of resin needed for the preparation. For grass and cereal samples columns were packed with the 100–150 μm resin to gain a lower limit of detection LLD below 2 and below 0.1 Bq/kg_{dry matter} respectively. The prepacked 2 mL columns with particle size 100–150 μm were used for soil (LLD below 2 Bq/kg_{dry matter}). After digestion of soil samples, hydroxide precipitation was used as an additional separation step. The ⁹⁰Sr was measured by liquid scintillation counting. For quality control reasons, first the initial strontium concentration in the sample was determined then a strontium carrier solution was added and after the separation steps the chemical recovery was determined by ICP-MS. Thus, no radioactive tracer and just a small amount of the measuring solution were needed. The results are presented and discussed. These results will be used as reference for further ⁹⁰Sr analyses which will be conducted in a 5 year period to detect any radiological impact of the nuclear power plant Temelin on the environment of Austria.     

Determination of 90Sr / 238U ratio by double isotope dilution inductively coupled plasma mass spectrometer with multiple collection in spent nuclear fuel samples with in situ 90Sr / 90Zr separation in a collision-reaction cell

Strontium-90 is one of the most important fission products generated in nuclear industry. In the research field concerning nuclear waste disposal in deep geological environment, it is necessary to quantify accurately and precisely its concentration (or the ⁹⁰Sr / ²³⁸U atomic ratio) in irradiated fuels. To obtain accurate analysis of radioactive ⁹⁰Sr, mass spectrometry associated with isotope dilution is the most appropriated method. But, in nuclear fuel samples the interference with ⁹⁰Zr must be previously eliminated. An inductively coupled plasma mass spectrometer with multiple collection, equipped with an hexapole collision cell, has been used to eliminate the ⁹⁰Sr / ⁹⁰Zr interference by addition of oxygen in the collision cell as a reactant gas. Zr⁺ ions are converted into ZrO⁺, whereas Sr⁺ ions are not reactive.A mixed solution, prepared from a solution of enriched ⁸⁴Sr and a solution of enriched ²³⁵U was then used to quantify the ⁹⁰Sr / ²³⁸U ratio in spent fuel sample solutions using the double isotope dilution method. This paper shows the results, the reproducibility and the uncertainties that can be obtained with this method to quantify the ⁹⁰Sr / ²³⁸U atomic ratio in an UOX (uranium oxide) and a MOX (mixed oxide) spent fuel samples using the collision cell of an inductively coupled plasma mass spectrometer with multiple collection to perform the ⁹⁰Sr / ⁹⁰Zr separation. A comparison with the results obtained by inductively coupled plasma mass spectrometer with multiple collection after a chemical separation of strontium from zirconium using a Sr spec resin (Eichrom) has been performed. Finally, to validate the analytical procedure developed, measurements of the same samples have been performed by thermal ionization mass spectrometry, used as an independent technique, after chemical separation of Sr.     

Alpha emitters from uranium mining in the environment

Uranium mining and milling activities usually generate an enhancement of radionuclide concentrations in the environment that may cause increased radiological exposure to mankind. For risk assessment and radiological protection of man and environment in these areas, usually, it is needed to implement radiological surveillance of water, soils, agricultural products, aerosols, and mining waste discharges as well. Radionuclides to be monitored in priority are alpha-emitting nuclides of the uranium natural series. Radioactivity analysis of materials from uranium mining areas of Portugal shows departure from secular radioactive equilibrium amongst uranium series radionuclides, thus rendering invalid the assumption of equilibrium and requiring the actual determination of each radionuclide. Radionuclide measurements performed with high resolution alphaspectrometry, as reported herein, produce accurate results on specific radionuclides that are essential in computing radiation doses to critical groups of the population.     

Role of Cyanex-272 as an extractant for uranium in the determination of rare earths by ICP-AES

A chemical separation procedure has been developed for the extraction of uranium from some of the crucially important rare earths using a novel extractant viz. Cyanex-272 (2,4,4-trimethyl pentyl phosphinic acid). The near total extraction of uranium and quantitative separation of rare earth elements has been validated using inductively coupled argon plasma - atomic emission spectrometry (ICP-AES). The recovery of some of the representative elements has been confirmed by radioactive tracer studies. The back extraction of uranium from the organic phase was carried out using a solution of 0.5M Na₂CO₃ which resulted in a near total recovery of uranium into the organic phase. These studies have enabled determination of sub ppm amounts of the analyte elements with a precision of 5% RSD utilizing prior chemical separation of rare earths from 1 g uranium samples in just three extractions with Cyanex-272.     

Production of carrier free <Superscript>188</Superscript>Re radioisotope generator based on aluminum tungstate matrix

Improved radionuclide generator include a substantially insoluble salt of a radioactive parent which may be directly packed in column for subsequent elution of the daughter radionuclide. An improved ¹⁸⁸Re generator was prepared by reacting a radioactive tungsten (¹⁸⁸W) as parent radionuclide incorporated with aluminum chloride to obtain an insoluble radioactive aluminum tungstate matrix. The investigated matrix was characterized on the basis of the chemical composition, IR, thermal analysis and mechanical stabilities. The factors affecting the elution performance were studied such as influence of pH, molar ratio and drying temperature. From the obtained data, the molar ratio W:Al was 1.5:1 at pH = 4, the matrix dried at 105 °C for 2 h. Chromatographic and multichannel analysis has been currently used to investigate the radiochemical and radionuclidic purity respectively on eluted ¹⁸⁸Re. An elution yield more than 80%, with radiochemical purity <98% and radionuclidic purity <99% with a ¹⁸⁸W break through >10⁻⁴% of the column. The Al⁺³ and W contents value were about 2 and 3 μg/mL eluate. The obtained data approved the stability of the prepared generator and its suitability for medical application.     

Determination of <Superscript>90</Superscript>Sr in environment of district Swat,Pakistan

Assessment of ⁹⁰Sr is of great interest owing to the fact that this artificially produced radionuclide has high radiological importance because of its high fission yield, chemical similarity to calcium and its relatively long biological and physical half-life. To assess the likely hazard to population, low level ⁹⁰Sr in environmental samples is determined using pre-equilibrated tri-butyl phosphate (TBP) solvent and extraction-liquid scintillation procedure. ⁹⁰Y is selectively extracted from nitric acid solution into TBP solvent and stripped into aqueous phase as oxalate. The activity is finally measured by low level liquid scintillation counter using Cerenkov radiation. The specific activity is found only in three vegetation samples with average value of 2.86±1.7 Bq·kg⁻¹ of dry weight. In all other samples analyzed, the activity is below the detectable limit, i.e., 0.03 Bq. Results obtained are comparable with other areas of Pakistan. The chemical recovery of ⁹⁰Y varies from 75 to 90% for soil, vegetation and water. The present study provides a general background of the detectable radionuclide for the surveyed area that will be helpful in any radiological emergency.     

Use of chromatographic pre-concentration for routine uranium bioassay analysis by ICP-MS

Routine monitoring of urine is an effective way to detect occupational intake of radioactive material. Historically, determinations of uranium isotopic ratios have been performed by radiochemical separation followed by alpha spectrometry. With recent advancements in technology, inductively coupled plasma-mass spectrometry (ICP-MS) has become widely available for the determination of trace metals as well as radioactive nuclides with long half-lives, such as ²³⁸U in urine. Furthermore, ICP-MS measurements of ²³⁸U do not require radiochemical separation since the number of atoms in the sample is determined instead of the number of alpha particles emitted. However, this method does not provide good sensitivity for the determination of ²³⁵U due to its shorter half-life. An improved procedure using pre-concentration of uranium and determination by ICP-MS decreases the detection limit by a factor of ten or greater with only slight increase in total analysis time. The method also has the capability of accurately determining the isotopic ratio of the sample, which is very important in cases where enriched or depleted uranium is involved.     

Method for the detection of Tc in seaweed samples coupling the use of Re as a chemical tracer and isotope dilution inductively coupled plasma mass spectrometry

Analysis of the artificial radionuclide ⁹⁹Tc in environmental samples requires a chemical separation due to its low concentration, and therefore the use of a chemical yield tracer is peremptory. From a practical viewpoint, Re can be used for this purpose, due to its chemical similarities with Tc. Thus, the use of a radioactive tracer for Tc recovery calculation can be avoided. However, results from a recent intercomparison exercise showed that using of Re as a chemical yield tracer appears to underestimate the Tc concentration relative to the result obtained with isotopes of Tc. In the present work, the methodology used to design a simple separation method for the measurement of ⁹⁹Tc in environmental samples is described. Tc recovery is estimated throughout the Re recovery calculation by the isotope dilution technique coupled with ICP-MS (ID-ICP-MS) technique. For chemical separation, a chromatographic resin is used. Interfering elements are removed using a resin washing step carefully designed to avoid any element fractionation between Re and Tc; the care taken in this step is of major importance to assure the equivalence of the chemical recoveries for both elements. Agreement is tested using five replicates of five seaweed samples. The average recoveries for ^95mTc and Re were 93±6 and 95±7%, respectively, those are within the uncertainty intervals for each other. The results explained here demonstrate the possibility of applying Re chemical recoveries to calculate the Tc concentrations with the advantage of not introducing systematic errors.     

Rapid determination of 226Ra in emergency urine samples

A new method has been developed at the Savannah River National Laboratory (SRNL) that can be used for the rapid determination of ²²⁶Ra in emergency urine samples following a radiological incident. If a radiological dispersive device event or a nuclear accident occurs, there will be an urgent need for rapid analyses of radionuclides in urine samples to ensure the safety of the public. Large numbers of urine samples will have to be analyzed very quickly. This new SRNL method was applied to 100 mL urine aliquots, however this method can be applied to smaller or larger sample aliquots as needed. The method was optimized for rapid turnaround times; urine samples may be prepared for counting in <3 h. A rapid calcium phosphate precipitation method was used to pre-concentrate ²²⁶Ra from the urine sample matrix, followed by removal of calcium by cation exchange separation. A stacked elution method using DGA Resin was used to purify the ²²⁶Ra during the cation exchange elution step. This approach combines the cation resin elution step with the simultaneous purification of ²²⁶Ra with DGA Resin, saving time. ¹³³Ba was used instead of ²²⁵Ra as tracer to allow immediate counting; however, ²²⁵Ra can still be used as an option. The rapid purification of ²²⁶Ra to remove interferences using DGA Resin was compared with a slightly longer Ln Resin approach. A final barium sulfate micro-precipitation step was used with isopropanol present to reduce solubility; producing alpha spectrometry sources with peaks typically <40 keV FWHM (full width half max). This new rapid method is fast, has very high tracer yield (>90 %), and removes interferences effectively. The sample preparation method can also be adapted to ICP-MS measurement of ²²⁶Ra, with rapid removal of isobaric interferences.     

Methodology for determination of <Superscript>89</Superscript>Sr and <Superscript>90</Superscript>Sr in radiological emergency: II. Method development and evaluation

In this work a method for the determination of both ⁸⁹Sr and ⁹⁰Sr is presented. The method can potentially be used in radiological emergency and deliver results shortly after an incident. The method development was based on theoretical calculations of potential interferences from other fission products and how these could be discriminated when applying different chemical separation schemes. Validation was done on reactor coolant water containing short-lived fission products, and on a reference material. The results indicate that correct results of ⁸⁹Sr and ⁹⁰Sr can be obtained 4 and 9 days, respectively, after an incident.     

The effect of physicochemical parameters on the separation recovery of plutonium and uranium from aqueous solutions by cation exchange

The effect of physicochemical parameters such as pH, salinity (e.g. [NaCl]) and competitive cation (e.g. Ca²⁺ and Fe³⁺) concentration on the separation recovery of plutonium and uranium from aqueous solutions by cation exchange has been investigated. The investigation was performed to evaluate the applicability of cation exchange as separation and pre-concentration method prior to the radiometric analysis of uranium and plutonium isotopes in natural water samples. Application of the method to test solutions of constant radionuclide concentration and variable composition (0.1, 0.5 and 1 M NaCl; 0.1 and 0.5 M Ca(NO₃)₂; 0.1 and 1 mM FeCl₃; 10) has generally shown that: (1) the optimum pH is 4.5 for uranium and plutonium, (2) increasing salinity results in slightly lower for uranium and significantly higher chemical recovery plutonium and (3) the presence of Ca(II) cations doesn’t significantly affect the chemical recovery of both radionuclides. Contrary, the presence of Fe(III) cations ([Fe(III)] > 0.1 mM) results in significantly lower chemical recovery for both radionuclides (<50%). The later is attributed to the formation of Fe(III) colloids, which present increased chemical affinity for uranium and plutonium and hence compete with the radionuclide binding by the resin. Nevertheless, the results indicate that the method could be successfully applied to a wide range of natural waters.     

Separation of neptunium, plutonium, americium and curium from uranium with di-(2-ethylhexyl)-phosphoric acid (HDEHP) for radiometric and ICP-MS analysis

The possibility of using di-(2-ethylhexyl)-phosphoric acid (HDEHP) in solvent extraction for the separation of neptunium, plutonium, americium and curium from large amounts of uranium was studied. Neptunium, plutonium, americium and curium (as well as uranium) were extracted from HNO₃, whereafter americium and curium were back-extracted with 5M HNO₃. Thereafter was neptunium back-extracted in 1M HNO₃ containing hydroxylamine hydronitrate. Finally, plutonium was back-extracted in 3M HCl containing Ti(III). The method separates²³⁸Pu from²⁴¹Am for α-spectroscopy. For ICP-MS analysis, the interferences from²³⁸U are eliminated: tailing from²³⁸U, for analysis of²³⁷Np, and the interference of²³⁸UH⁺ for analysis of²³⁹Pu. The method has been used for the analysis of actinides in samples from a spent nuclear fuel leaching and radionuclide transport experiment.     

Biosorption of 241Am by immobilized Saccharomyces cerevisiae

More than half of the world's annual production of radionuclides is used for medical purposes such as diagnostic imaging of diseases and patient therapy. Using aqueous homogeneous solution reactor technology, production quantities of medical radioisotopes ⁹⁹Mo and⁸⁹Sr, can be extracted from one reactor cycle. ⁹⁹Mo may be produced directly from UO₂SO₄ uranyl sulfate in an aqueous homogeneous solution nuclear reactor in a manner that produces high purity radionuclides, making efficient use of the reactor's uranium fuel solution. The process is relatively simple, economical, and waste free, eliminating uranium targets. The short-lived radioisotope ^99mTc is eluted from ⁹⁹Mo for diagnostic imaging. Radioisotope ⁸⁹Sr infusion is a therapeutic modality that reduces reliance on narcotic analgesia through palliation of metastatic bone pain caused by metastases of the cancer to the bone. Painful disseminated osseous metastases are common with carcinomas of the lung, prostate, and breast. Synergistic interleaving of two manufacturing processes, one producing ⁹⁹Mo and another producing ⁸⁹Sr in the same production cycle of an aqueous homogeneous solution reactor makes full and efficient use of the time for both the neutron irradiation stage and the extraction stage of each radionuclide. Interleaving the capture of ⁸⁹Sr radioisotope with production processing of ⁹⁹Mo radioisotope is achieved, since the extraction and subsequent elimination of radionuclide impurities occurs during separate parts of the reactor cycle. The process applies to either HEU or LEU nuclear fuels in an aqueous homogeneous solution reactor.     

Sorbent separation and enrichment method for cobalt ions determination by graphite furnace atomic absorption spectrometry in water and urine samples using multiwall carbon nanotubes

A new cobalt ions pre-concentration method, optimised by fractional factorial design, using multiwall carbon nanotubes (MWCNTs) with further Graphite Furnace Atomic Absorption Spectrometry (GFAAS) quantification is described. The method explores the high chemical and physical stability of MWCNTs for improving the detectability of GFAAS. It is based on off-line pre-concentration of 20.0 mL of sample previously buffered (pH 8.82) on MWCNTs at a flow rate of 10.0 mL min^?1. After the pre-concentration procedure, the elution step was carried out with 500 µL of 0.524 mol L^?1 HNO₃ solution at a flow rate of 2.0 mL min^?1. Fractional factorial designs and response surface methodology were employed for optimisation of all chemical parameters involved in the pre-concentration procedure, including pre-concentration flow rate, buffer and eluent concentration, sample pH and elution volume. The method provides a linear calibration range from 0.03 up to 7.00 µg L^?1 with linear correlation coefficient higher than 0.9994 and limits of detection and quantification of 0.01 and 0.03 µg L^?1, respectively. Repeatability of the six measurements was found to be 2.38 and 1.84% for 0.3 and 4.5 µg L^?1 cobalt concentration, respectively. By pre-concentrating 20.0 mL of sample, a pre-concentration factor (PF) of 19.10-fold and a consumption index of 1.05 mL were obtained. The pre-concentration efficiency (PE) was found to be 9.55 min^?1. The proposed method was successfully applied for the pre-concentration and determination of cobalt in water and urine samples with satisfactory recovery values.     

Determination of alpha-emitting isotopes of uranium and thorium in vegetables and excreta

A radiochemical procedure for the determination of alpha-emitting isotopes of uranium and thorium in vegetables and excreta has been optimized, involving sample dissolution, separation by ionic exchange resin, electrodeposition and alpha-spectroscopy. Uranium and thorium isotopes were determined separately to prevent interference of ²²⁸Th from ²³²U tracer with ²²⁸Th from natural series of 232Th. This procedure was applied to faeces from people living in the Poços de Caldas plateau, a high natural radioactivity region of Brazil, and vegetables from the Laboratory of Environmental Monitoring (EML/DOE). Results show a chemical recovery of 80–95% for uranium and 46–72% for thorium.     

Automated determination of uranium(VI) at ultra trace levels exploiting flow techniques and spectrophotometric detection using a liquid waveguide capillary cell

Rapid and fully automated multisyringe flow-injection analysis (MSFIA) with a multi-pumping flow system (MPFS) coupled to a long path-length liquid waveguide capillary cell (LWCC) is proposed for the determination of uranium(VI) at ultra trace levels. On-line separation and pre-concentration of uranium is carried out by means of a TRU resin. After elution, uranium(VI) is spectrophotometrically detected after reaction with arsenazo-III. Combination of the MSFIA and MPFS techniques with the TRU-resin enables the analysis to be performed in a short time, using large sample volumes and achieving high selectivity and sensitivity levels. A detection limit of 12.6 ng L⁻¹ (ppt) is reached for a 100-mL sample volume. The versatility of the proposed method also enables pre-concentration of variable sample volumes, enabling application of the analysis to a wide concentration range. Reproducibility of better than 5% and a resin durability of 40 injections should be emphasized. The developed method was successfully applied to different types of environmental sample matrices with recoveries between 95 and 108%.