Rapid determination of radiostrontium in seawater samples

A new method for the determination of radiostrontium in seawater samples has been developed at the Savannah River National Laboratory (SRNL) that allows rapid pre-concentration and separation of strontium and yttrium isotopes in seawater samples for measurement. The new SRNL method employs a novel and effective pre-concentration step that utilizes a blend of calcium phosphate with iron hydroxide to collect both strontium and yttrium rapidly from the seawater matrix with enhanced chemical yields. The pre-concentration steps, in combination with rapid Sr Resin and DGA Resin cartridge separation options using vacuum box technology, allow seawater samples up to 10 L to be analyzed. The total ⁸⁹Sr + ⁹⁰Sr activity may be determined by gas flow proportional counting and recounted after ingrowth of ⁹⁰Y to differentiate ⁸⁹Sr from ⁹⁰Sr. Gas flow proportional counting provides a lower method detection limit than liquid scintillation or Cerenkov counting and allows simultaneous counting of samples. Simultaneous counting allows for longer count times and lower method detection limits without handling very large aliquots of seawater. Seawater samples up to 6 L may be analyzed using Sr Resin for ⁸⁹Sr and ⁹⁰Sr with a minimum detectable activity (MDA) of 1–10 mBq/L, depending on count times. Seawater samples up to 10 L may be analyzed for ⁹⁰Sr using a DGA Resin method via collection and purification of ⁹⁰Y only. If ⁸⁹Sr and other fission products are present, then ⁹¹Y (beta energy 1.55 MeV, 58.5 day half-life) is also likely to be present. ⁹¹Y interferes with attempts to collect ⁹⁰Y directly from the seawater sample without initial purification of Sr isotopes first and ⁹⁰Y ingrowth. The DGA Resin option can be used to determine ⁹⁰Sr, and if ⁹¹Y is also present, an ingrowth option with using DGA Resin again to collect ⁹⁰Y can be performed. An MDA for ⁹⁰Sr of <1 mBq/L for an 8 h count may be obtained using 10 L seawater sample aliquots.     

Practicality of Tetragonal Nano-Zirconia as a Prospective Sorbent in the Preparation of 99Mo/99mTc Generator for Biomedical Applications

The feasibility of using tetragonal nano-zirconia (t-ZrO₂) as an effective sorbent for developing a ⁹⁹Mo/^99mTc chromatographic generator was demonstrated. The structural characteristics of the sorbent matrix were investigated by different analytical techniques such as XRD, BET surface area analysis, FT-IR, TEM etc. The material synthesized was nanocrystalline, in tetragonal phase with an average particle size of ~7 nm and a large surface area of 340 m² g^?1. The equilibrium sorption capacity of t-ZrO₂ is >250 mg Mo g^?1. The present study indicates that ⁹⁹Mo is both strongly and selectively retained by t-ZrO₂ at acidic pH and ^99mTc could be readily eluted from it, using 0.9% NaCl solution. A 9.25 GBq (250 mCi) t-ZrO₂ based chromatographic ⁹⁹Mo/^99mTc generator was developed and its performance was repeatedly evaluated for 10 days. ^99mTc could be eluted with >85% yield having acceptable radionuclidic, radiochemical and chemical purity for clinical applications. The compatibility of the product in the preparation of ^99mTc labeled formulations such as ^99mTc-EC and ^99mTc-DMSA was evaluated and found to be satisfactory.     

Rapid method for radiostrontium determination in milk in emergency situations using PS resin

This study describes a new and rapid procedure for radiostrontium determination in milk samples based on the use of plastic scintillation resins (PS resins). The proposed method reduces the time of analysis by at least 2 h by combining separation and measurement preparation into a single step and optimizing the pre-treatment steps. The method is robust and reproducible, with good total recoveries (65% on average) and a relative bias for total radiostrontium activity (⁸⁹Sr + ⁹⁰Sr) below 7%. The minimum detectable activity for 100 mL of milk sample measured for 60 min is about 0.34 Bq L^?1. The proposed method can quantify radiostrontium content in 5 h, which makes it suitable for use in emergency situations.     

Chemical and electrochemical behavior of metallic technetium in acidic media

The chemical and electrochemical properties of technetium metal were studied in 1–6 M HX and in 1 M NaX (pH 1 and 2.5), X = Cl, NO₃. The chemical dissolution rates of Tc metal were higher in HNO₃ than in HCl (i.e. 8.63 × 10^?5 mol cm^?2 h^?1 in 6 M HNO₃ versus 2.05 × 10^?9 mol cm^?2 h^?1 in 6 M HCl). The electrochemical dissolution rates in HNO₃ and HCl were similar and mainly depended on the electrochemical potential and the acid concentration. The optimum dissolution of Tc metal was obtained in 1 M HNO₃ at 1 V/AgAgCl (1.70 × 10^?3 mol cm^?2 h^?1). The dissolution potentials of Tc metal in nitric acid were in the range of 0.596–0.832 V/AgAgCl. Comparison of Tc behavior with Mo and Ru indicated that in HNO₃, the dissolution rate followed the order: Mo > Tc > Ru, and for dissolution potential the order: E _diss(Ru) > E _diss(Tc) > E _diss(Mo). The corrosion products of Tc metal were analyzed in HCl solution by UV–Visible spectroscopy and showed the presence of TcO₄ ^?. The surface of the electrode was characterized by microscopic techniques; it indicated that Tc metal preferentially corroded at the scratches formed during the polishing and no oxide layer was observed.     

A rapid method for determining strontium-90 in urine samples

Rapid bioassay methods for ⁹⁰Sr in urine samples are needed to provide an early estimation of possible internal dose resulting from exposure to radiostrontium in the event of a radiological and nuclear emergency. In this work, a fast column separation method followed by liquid scintillation counting for detection of ⁹⁰Sr in urine was developed. Replicate spike and blank samples were analyzed for performance evaluation of the method. Using this method, a detection limit of ~10 Bq L^?1 for ⁹⁰Sr can be achieved with a sample analysis turn-around time of 4 h for a set of 12 samples. The method is adequate to meet the radiobioassay acceptance criteria and is suitable for quick dose assessment of ⁹⁰Sr exposure following a radiation emergency.     

Preparation of 99Mo/99mTc generator based on alumina 99Mo-molybdate (VI) gel

In this work alumina ⁹⁹Mo-molybdate (VI) gel is evaluated as a column matrix for use in the preparation of small chromatographic column type ^99mTc generator. Alumina molybdate (VI) gel is prepared by dissolving inactive MoO₃ with aluminum foil in 5 M NaOH solution containing ⁹⁹Mo radiotracer. After complete dissolution, 0.5 H₂O₂ was added to the reaction mixture solution and acidified to pH 5.5 with concentrated HNO₃. The formed AlMo precipitate was washed with NaNO₃ solution, dried at 50 °C for 24 h and then packed in the form of a chromatographic column for elution of the generated ^99mTc radionuclide with physiological saline solution (0.9 % NaCl). Greater than 86 % of the generated ^99mTc activity is immediately and reproducibly eluted with passing 10 mL of the saline solution through 2.0 g of alumina ⁹⁹Mo-molybdate column bed at a flow rate of about 1.0 mL/min. The high radiochemical ≥98.6 % TcO₄ ^?, radionuclidic ≥99.90 % ^99mTc and chemical purities of the eluates satisfy the specifications for use in nuclear medicine.     

Development of field portable sampling and analysis systems

A portable, field rugged, sampling and analysis system has been developed for the rapid screening of aqueous samples during scoping and remediation studies. Using field portable equipment, water is pumped through ion selective solid phase extraction (SPE) disks, at a flow rate of 150-250 ml/min, and counted for the radionuclide of interest in the field using portable detectors. SPE disks are currently available to selectively concentrate ⁹⁹Tc, ⁹⁰Sr, radiocesium (¹³⁴Cs and ¹³⁷Cs) and radium isotopes. In the field the radiocesium concentration is determined by gamma-spectrometry, ⁹⁰Sr and ⁹⁹Tc are determined by beta-counting. A one-liter sample can be processed and ready for counting within ten minutes. Using a 5-minute counting time, a detection limit of <50 pCi/l for ⁹⁹Tc or ⁹⁰Sr and ~50 pCi/l for ¹³⁷Cs has been achieved. Up to 10 liters of water have been processed for the analysis of ⁹⁹Tc and ¹³⁷Cs when lower limits of detection were required. The sampling and analysis system has been field tested at the Savannah River Site (SRS), Aiken SC, and the Hanford Site, Richland WA. The SRS H-area tank farm storm water runoff system was analyzed for ⁹⁰Sr and ¹³⁷Cs. Groundwater from the SX tank farm at the Hanford Site was analyzed for ¹³⁷Cs and ⁹⁹Tc. Groundwater from seeps below the 100-H area at Hanford was analyzed for ⁹⁰Sr and ⁹⁹Tc.     

Intercomparison of phosphorus-32, technetium-99m and iodine-131 activity measurement in Pakistani nuclear medical centers and supplier

Technetium-99m and iodine-131 are the leading isotopes used in diagnosis and radiotherapy, respectively in Pakistani nuclear medical centers. Both radionuclides are regularly supplied by Isotope Production Division (IPD), PINSTECH to nuclear medical centers in Pakistan. The packages of ³²P, ⁹⁹Mo → ^99mTc generators, and ¹³¹I, are prepared at IPD and dispatched by air, rail or road according to the desired activity at reference day to nuclear medical centers. An accurate determination of the activity to be administered to the patient plays key role on the quality of the clinical result of a medical investigation. Activity meters are employed for quantification of radioactivity of radionuclides at supplier site as well as by the users. In order to investigate the performance on the quality of activity measurements and elution efficiency of PAKGEN ^99mTc generators, a pilot program of intercomparison was organized between the supplier and the users. The results of intercomparison of the activity measurements showed a maximum difference of ±13 % between supplier and nuclear medical centers.     

Extraction of radiostrontium from the mixture of radionuclides in milk using dicarbolide of cobalt

The possibilities of radiostrontium extraction from milk after the isolation of radiocesium were investigated. The polyhedral complex of the type H⁺[(π-/3/-1,2-C₂B₉H₁₁)₂Co⁻], further referred to as dicarbolide-H⁺, in nitrobenzene was used for the extraction. The increase of values and larger differentiation of distribution ratios in the extraction of ion associates of radionuclides were achieved by hydrophobizing with polyethyleneglycol (PEG). The changes of the distribution ratios of⁹⁰Sr,⁸⁹Sr and⁹⁰Y with PEG concentration were studied, and ranges of the highest synergic effect, changing with different extraction agent concentrations in extraction from water solutions and a fresh milk were determined. The influence of milk dilution and nitric acid concentration upon the value of distribution ratio in radiostrontium extraction using dicarbolide-H⁺ and the optimum PEG concentration were investigated. Under suitably selected conditions of radiostrontium isolation, the distribution ratios of some potential contaminants were determined. The influence of nonisotopic carrier Ca²⁺ upon radiostrontium extraction was examined. The possibilities of purification of isolated radiostrontium using reextraction were investigated. A selective and quick analytical procedure was suggested for radiostrontium isolation after the separation of radiocesium from a mixture of fission and activated radionuclides in milk using extraction by dicarbolide-H⁺. The chemical yield is about 98% and⁹⁰Sr is determined by liquid scintillation technique.     

The study of conditions for the preparation and application of99Mo−99mTc generators starting from irradiated molybdenum metal

⁹⁹Mo−^99mTc generators were prepared starting from irradiated molybdenum metal instead of MoO₃ in order to use reactor irradiation space more economically. The adsorption of molybdenum as sodium molybdate on different kinds of alumina was investigated. The effect of the pH of the column and the aqueous phase concentration of molybdenum were studied and related to the elution yield of^99mTc. A study of the radiation damage effect indicated that generators having a high elution efficiency of 90% could be prepared in the 100–600 mCi range. The losses of⁹⁹Mo were minimized to 10⁻⁵-10⁻⁴% and those of alumina to 2–5 μg/ml eluate.     

Cerenkov spectroscopic assay of fission isotopes

The procedure for the assay of⁹⁰Sr or⁸⁹Sr in the presence of a weak beta-emitter, such as¹³⁷Cs is described which could be applicable in studies on the purification and decontamination of aqueous radioactive wastes by ion-exchangers. The influence of colour-quenchers on the counting efficiency of radiostrontium has also been investigated.     

Determination of radionuclidic impurities in <Superscript>99m</Superscript>Tc eluate from <Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generator for quality control

Technetium-99m is the principal radioisotope used in medical diagnostics; radionuclidic impurity is the major concern of its quality. This work presents a analytical method for sequential determination of all radionuclidic impurities listed in pharmacopoeia including gamma emitters, alpha emitters, ⁸⁹Sr and ⁹⁰Sr. Radioactive decay for removal of ^99mTc, ion exchange and extraction chromatography for removal of ⁹⁹Mo and ⁹⁹Tc are effective for separation of interferences. Gamma spectrometry, LSC with alpha/beta discrimination, and Cherenkov counting using LSC are sensitive methods for measurement of the impurity radionuclides. The detection limits of this method are well meet the requirement of the quality control according to the limitation of the pharmacopoeia.     

The simple radiochemical determination of 90Sr in environmental solid samples by solvent extraction

Determination of ⁹⁰Sr in environmental solid samples is a challenging task because of the presence of so many other radionuclides in samples of interest. This problem was dealt with by radiochemical separation of strontium followed by yttrium separation and Cerenkov counting of the high-energy ??-particle emissions of ⁹⁰Y in order to quantitate ⁹⁰Sr. In this work, an improved method is described for the determination of ⁹⁰Sr in soil samples, through the separation of the daughter ⁹⁰Y at equilibrium. The procedure is based on the HDEHP solvent extraction in combination with liquid scintillation spectrometry (LSS). A low background Quantulus has been optimized for low level counting of Cerenkov radiation emitted by the hard ??-emitter ⁹⁰Y. The analytical quality of the method has been checked by analyzing IAEA Soil-375 reference materials. The analytical method has also been successfully applied to the determination of ⁹⁰Sr for moss-soil samples in inter-laboratory exercises through IAEA??s ALMERA network. The chemical recovery for ⁹⁰Y extraction ranged from 80 to 85% and the counting efficiency was 73% in the window 25?C400 keV.     

Development of rapid solvent extraction based radioanalytical technique for estimation of 90Sr in the presence of natural and anthropogenic radionuclides

A radiochemical technique for estimation of ⁹⁰Sr concentration by direct extraction of ⁹⁰Y with bis (2-ethylhexyl) phosphate (HDEHP) in toluene was developed and standardized. About 90% ⁹⁰Sr retained in the aqueous phase at the optimized pH 4–4.5. Cerenkov counting with 0.04 g mL⁻¹ sodium salicylate as a wavelength shifter resulted in 7.5 fold increase its Cerenkov counting efficiency. The method was validated by estimating ⁹⁰Sr in certified reference materials like, IAEA TEL 2016-03 drinking water, Spruce needle and IAEA-156 Clover. The interference of naturally occurring Ra and its β progeny especially Pb was eliminated by means of Cerenkov counting. Conventional radiochemical separation was used to compare ⁹⁰Sr activity in effluent samples.

     

Simultaneous determination of <Superscript>89</Superscript>Sr and <Superscript>90</Superscript>Sr: comparison of methods and calculation techniques

The presence of ⁸⁹Sr and ⁹⁰Sr in the biosphere constitutes a biological hazard. There are several analytical methods for the determination of ⁸⁹Sr and ⁹⁰Sr. Three analytical methods of various application fields using selective Sr resin for Sr separation and DGA resin for Y separation and measuring techniques, i.e. liquid scintillation spectrometry and Cerenkov counting are discussed in the paper. The calculation techniques are compared in the aspects such as trueness and accuracy of the results and the limit of detection. Uncertainties and detection limits are calculated using the spreadsheet method.     

Effects of ionic strength and humic acid on 99TcO4 ? sorption and diffusion in Beishan granite

In terms of pre-safety assessment of a potential site for high-level radioactive wastes disposal in China, the geochemical behavior of key radionuclides which tend to be released from the repository must be thoroughly investigated. ⁹⁹Tc is a long-lived fission product with appreciable productivity in nuclear fuel, and Tc (+7) has unlimited solubility in near-field geochemical environments. In this study, the effects of ionic strength and humic acid on ⁹⁹TcO₄ ^? sorption and diffusion in Beishan granite were investigated with through-diffusion and batch sorption experiments. Results indicated that the effective diffusion coefficients (D _e) of ⁹⁹TcO₄ ^? in Beishan granite varied from 1.07?×?10^?12 to 1.28?×?10^?12?m²/s without change with ionic strength, while the distribution coefficients (K _d) negatively correlated with ionic strength of the rock/water system. This study also indicates that there is no evident influence of humic acid concentration on the diffusion behavior of ⁹⁹TcO₄ ^? in Beishan granite, due to the limited interaction between humic acid and ⁹⁹TcO₄ ^?.     

Ion exchange of 90Sr and 137Cs into 1‐vinyl‐2‐pyrrolidone–divinylbenzene cation‐exchange resin

Radiotracer batch ion‐exchange experiments were employed to investigate the uptake of ⁹⁰Sr and ¹³⁷Cs radioisotopes by various cation‐exchanged forms of a 30% cross‐linked macroporous 1‐vinyl‐2‐pyrrolidone–divinylbenzene cation‐exchange resin with 1.37 ml g^?1 pore volume, 0.0232 µm pore diameter and 271.2 m² g^?1 surface area. The uptake of ⁹⁰Sr and ¹³⁷Cs was determined by taking liquid aliquots at various time intervals from solutions over solids. The volume‐to‐solid ratio was kept at 200. The results of kinetic experiments for the carrier‐free ⁹⁰Sr and ¹³⁷Cs were evident in all cationic forms of the resin. The percentage uptake and distribution coefficient K_d values with carrier (0.005 M SrCl₂ and 0.01 M CsCl) concentrations were also determined, and the best results were obtained from the Li⁺ and H⁺ forms of the resin. Cerenkov counting (β^?‐counting) was used to observe the initial and final radioactivity in the liquid phase. All the experiments were carried out at room temperature and the radioactivity in each case was corrected for the background counts. Copyright © 2005 John Wiley & Sons, Ltd.     

Separation of no-carrier-added 99mTcO4 − from 99Mo–99mTc equilibrium mixture by PEG based aqueous biphasic separation technique using sodium/potassium salts of citric and tartaric acid

Extraction and separation of no-carrier-added (nca) ^99mTcO₄ ^? from ⁹⁹Mo–^99mTc equilibrium mixture was carried out by environmentally benign polyethylene glycol based liquid–liquid aqueous biphasic systems (ABS) consisting various inorganic salts. Among the various inorganic salt trisodium citrate and potassium sodium tartrate showed the suitable salt rich phase for the best separation in this report. The concentration variation of salt rich phase, temperature and PEG phase also exhaustively studied in paper for the achievement of high separation factor. At 40 °C temperature in 50 % (w/v) PEG-4000-2M Na₃citrate showed the highest separation factor (S _Tc/Mo) 1.2 × 10⁷.     

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.     

Effect of EDTA on the diffusion behavior of 99TcO4 − and ReO4 − in GMZ bentonite

Polyaminopolycarboxylate EDTA with powerful metal-binding property, which often presents in low and intermediate-level waste, can enhance the radionuclide migration. The effect of EDTA on the diffusion behavior of ⁹⁹TcO₄ ^? and ReO₄ ^? in Gaomiaozi (GMZ) bentonite was investigated by using through-diffusion method. For ⁹⁹TcO₄ ^? in present of EDTA, the D _e values was (1.2 ± 0.1) × 10^?11 m²/s, which was 4 times higher than that in absent of EDTA. It can be explained that the complexation between ⁹⁹TcO₄ ^? and EDTA might be formed. By contrast, the D _e values of ReO₄ ^? remained unchanged in present or absent of EDTA, indicating that ReO₄ ^? could not complex with EDTA. However, the diffusion of ReO₄ ^? could be increased in present of EDTA, the D _a value was found to be increased from 1.8 × 10^?10 to 5.4 × 10^?10 m²/s. It demonstrated that ReO₄ ^? need more drastic conditions to form the Re(VII)–EDTA complexes than those used for ⁹⁹TcO₄ ^?. For both ReO₄ ^? and ⁹⁹TcO₄ ^?, the rock capacity factor α is less than the total porosity ε _tot, indicating that they has little retention/sorption on the surface of bentonite.