Methodology for determination of 89Sr and 90Sr in radiological emergency: I. Scenario dependent evaluation of potentially interfering radionuclides

Early determination of ⁸⁹Sr and ⁹⁰Sr in radiological emergency is hampered by the presence of interfering short-lived fission products. In this study, three commonly used radioanalytical strategies for ⁸⁹Sr and ⁹⁰Sr were evaluated theoretically considering their suitability in a nuclear explosion scenario. The methods were evaluated with respect to the need for decay time of interfering short-lived strontium and yttrium isotopes, and reduction of other known interfering nuclides prior to measurement. The strategy shown to be most successful included initial separation of strontium and determination of ⁸⁹Sr, followed by an yttrium separation and counting of ⁹⁰Y. ⁸⁹Sr and ⁹⁰Sr could be determined about 5 and 9 days after a nuclear explosion, respectively.     

Nuclear decay studies of rare-earth fission-product nuclides using fast radiochemical separation techniques

A facility for the rapid radiochemical separation of individual rare-earth fission product nuclides from mixed fission products has been developed at the Idaho National Engineering Laboratory (INEL). This facility, called the INEL ESOL (elemental separation on-line) facility, includes an electroplated spontaneously fissioning²⁵²Cf source, a He jet transport system to deliver short half-life fission products from the²⁵²Cf hot cell to the radiochemistry laboratory, and a high pressure liquid chromatograph for the individual separation of rare-earth radioelemental fractions. The fission product collection and separation instrumentation is interfaced to a microprocessor that controls valves, motors and other devices and monitoring instrumentation. The data acquistion instrumentation can be controlled from a signal originating from the microprocessor or initiated manually. The results of some recent decay scheme studies on¹⁵³Pm and¹⁵⁴Pm are reported.     

Separation of <Superscript>142</Superscript>La from fission products by SISAK technique

A rapid separation system based on SISAK technique was established to isolate ¹⁴²La successfully from fission products. SISAK technique is often applied in the separation of nuclides with the half-life of seconds or minutes. Here it was used to separate the parent nuclide of ¹⁴²La, which the half-life is in the magnitude of several seconds. According to the separation procedure designed in the paper, the activity of ¹⁴²La acquired is more than 10⁴ Bq and the decontamination factors for most γ-emitters are higher than 10³.     

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.     

Separation of95Zr and95Nb from each other and from other fission products by adsorption and desorption on silica gel using hydrochloric acid solutions

The silica gel adsorption behaviour of zirconium, niobium, ruthenium and cerium in hydrochloric acid has been investigated by batch and column techniques. A satisfactory radiochemical separation of zirconium and niobium from each other and from other fission products has been achieved by a two column technique. The recommended procedure consists of sorption of all the nuclides on a primary silica gel column. Fifteen per cent of⁹⁵Nb, all of the zirconium and all of the other fission products are eluted first by washing with 5.5 M HCl. A second elution with concentrated hydrochloric acid then recovers the⁹⁵Nb (free from other products). The solution from the first elution after evaporation to 1 ml is then passed through another silica gel column and successively washed with 0.5M HCl, 5.5M HCl and concentrated HCl to obtain three fractions—other fission products—⁹⁵Zr free from other products—⁹⁵Nb free from other products, respectively.     

Determination of radioisotopes of Ce,Eu, Pu,Am and Cm in low-level-wastes from power reactors

Long-lived isotopes of lanthanides and actinides are very important for the disposal of low-level radioactive wastes. These nuclides serve for risk calculations of accidents. Their determination requires radiochemical separation from high activity main nuclides. Supervision of waste vessels is done by direct non-destructive -spectrometry of the key nuclides:⁶⁰Co for corrosion products and¹³⁷Cs as for fission products as for transuranic elements. The activity ratios of the long-lived nuclides to the key nuclides are called scaling factors. They have to be determined radiochemically in the laboratory in representative samples of each waste type. They are used for activity calculations of long-lived nuclides in the waste vessels. The scaling factors determined are of the order of magnitude of 10^–6 and due to the fact that we have used low-level measurement techniques, we could performe the necessary chemical separations in a laboratory not exceeding the 10-fold free-level limit.     

Radiostrontium separation from sodium molybdate solution and its measurement using LSA: an application to radiopharmaceutical analysis

Technetium (^99mTc), a decay product of molybdenum (⁹⁹Mo), is employed as radioisotope in nuclear medicine. Several practical devices known as generators are commercially available which enable the user to separate the daughter from the parent radionuclide. The present study is focused on quality control of chromatographic technetium generator. A properly constructed generator should comply with international requirements of radionuclide purity of ⁹⁰Sr/⁹⁹Mo ≤ 6 × 10^?8 and ⁸⁹Sr/⁹⁹Mo ≤ 6 × 10^?7. For this purpose an analytical method was optimized to quantify radiostrontium (⁸⁹Sr and ⁹⁰Sr) in sodium molybdate [Na ₂ ⁹⁹ MoO₄] solution, a fission product used for ⁹⁹Mo/^99mTc generators. Dowex 1 × 8 and alumina were used in sequence followed by tributyl phosphate extraction for radiostrontium separation. Cerenkov measurement of ⁸⁹Sr and ⁹⁰Sr (through its descendent ⁹⁰Y) was performed using Perkin Elmer Tricarb LSA 3170 with detection efficiency of 42 and 14 %, respectively. Since efficiency of Cerenkov counting is sensitive to presence of color, spectral index of sample was used to correct the counting efficiency. The chemical recovery for strontium was 22 % and for yttrium was 80 % as determined by inductively coupled plasma optical emission spectrometry. Lower limit of detection was found to be 6.3 and 14.4 Bq L^?1 for ⁹⁰Sr and ⁸⁹Sr, respectively with 60 min counting time. Hence method can be applied successfully to analyze ^89,90Sr in fission molybdenum used as radiopharmaceutical with a relative error of <10 %.     

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO₂, U₃O₈, and solid solutions of actinides in UO₂ in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL^?1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

Alpha-spectrometry in nuclear fuel analysis

A survey is given of the analysis of actinide nuclides by means ofa-spectrometry in high-level radioactivity process solutions resulting from reprocessing of spent U-Th fuel. Separation of fission products and isolation of the nuclides²²⁸Th,²³¹Pa,²³²U,²³⁷Np and²³⁸Pu are performed by adsorption, ion-exchange, extraction chromatography and extraction techniques. After separation, samples for quantitative determination are prepared by electrodeposition and measured using a silicon-surface barrier detector combined with a multichannel analyzer. An error estimation is given.     

Nuclide analysis of a tantalum target irradiated with high-energy protons

First results of γ-spectrometric measurements of radionuclides produced in a tantalum target irradiated for 500 days with 800 MeV protons are presented. The activities of the γ-ray-emitting nuclides measured after a cooling period of 2–4 years differ by more than four orders of magnitude. Within this cooling period about 95% of the activity of the target is determined by¹⁷²Lu,¹⁷³Lu,¹⁷⁴Lu,¹⁷²Hf (daughter nuclide of¹⁷²Lu) and¹⁸²Ta nuclides. These highly active nuclides, together with some other medium-active nuclides, were measured with good precision by γ-ray spectrometry directly in the sample solution. Weakly active nuclides could not be measured in this way because of the high Compton background in the γ-ray spectrum. For the sensitive measurement of many other nuclides present in weaker activities a simple chemical separation procedure was developed to separate Lu, Hf and Ta. With the separation of these elements and simultaneously of the high-active nuclides, the Compton background in the γ-ray spectra could be drastically reduced and many weakly active nuclides additionally measured.     

Separation of Y from Sr by zirconium vanadate gel ion-exchanger sorbent: kinetics and thermodynamic study

A new zirconium vanadate (Zr–V) ion-exchanger was synthesized and characterized for fast and selective separation procedure of ⁹⁰Y from ⁸⁹Sr. The method was based on ⁹⁰Y(III) sorption from aqueous HCl solution containing ⁸⁹Sr(II) onto Zr–V gel exchanger. The kinetics of Y(III) sorption from HCl solution by Zr–V exchanger was subjected to Weber–Morris, Lagergren, Bhattacharya and Venkobachar, and Bt models. Initially, the uptake of Y(III) onto the exchanger was fast followed by kinetically first-order sorption with an overall rate constant, K _Lager = (3.55 ± 0.03) × 10^?4 min^?1. Film and intraparticle transport are the two steps that might influence Y(III) sorption. The negative values of ΔG of ⁹⁰Y retention dictate that, the process is a spontaneous. The negative values of ΔH and ΔS reflect the exothermic nature of ⁹⁰Y(IIsorption and the random uptake of ⁹⁰Y(III) onto Zr–V sorbent. Zr–V exchanger offers unique advantages of ⁹⁰Y(III) retention over conventional solid sorbents in rapid and effective separation of traces of ⁹⁰Y(III) from Sr. The exchanger was successfully packed in column for an effective separation of ⁹⁰Y.     

Rapid determination of 239,240Pu, 238Pu, 241Am and 90Sr in high contaminated samples waste using combined SPE sorbents AnaLig® Pu-02, AnaLig® Sr-01 and DGA® Resin

A rapid separation method has been developed which allows measurement of plutonium, americium and strontium isotopes in the high active sample from CBRN Training and Testing Centre in Zemianske Kostolany (Slovakia) with high chemical recoveries and effective removal of matrix interferences. This method uses different commercial products stacked AnaLig^® Pu-02, AnaLig^® Sr-01 and DGA^® Resin cartridges from IBC Advanced Technologies and Eichrom Technologies. The method allows rapid separation of plutonium, strontium and americium using a single multi-stage column in the vacuum box (cartridge technology) with rapid flow rates to minimize sample separation time. The ^239,240Pu, ²³⁸Pu and ²⁴¹Am were determined by alpha spectroscopy, ⁹⁰Sr was counted on TRICARB 2900 TR by Cerenkov counting of its progeny ⁹⁰Y.     

Optimization of 89Zr production using Monte Carlo simulations

Zirconium-89 (⁸⁹Zr) can be produced in a cyclotron by focusing the proton beam on an yttrium-89 (⁸⁹Y) foil target. Optimal combination of beam energy and target assembly configuration enables maximum production of ⁸⁹Zr while minimizing the formation of contaminant nuclides such as ⁸⁸Zr and ⁸⁸Y to allow efficient and effective radiopharmaceutical labeling. Accurate modeling of the proton beam and the target is therefore an essential step to assure the best beam and target specification. We used the radiation transport code MCNPX to simulate the transport of protons through the irradiation assembly and the nuclear reaction code TALYS to obtain the production cross sections of various nuclides from proton-⁸⁹Y reactions. Results from simulating the irradiation of 14 mm diameter targets with aluminum (Al) degrader at 19.8 mA for 1 h suggested that the 0.15 mm thick one would produce 227 MBq while the 0.3 mm thick one would produce 413 MBq of ⁸⁹Zr with less than 1 % uncertainty. These results show excellent agreement with experimental work in literature. This work provides the basis for further experimental and theoretical assessments of the use of ⁸⁹Zr as radiopharmaceutical labels.     

The half-lives of 90mY and 97Ru

In an attempt to improve the accuracy of neutron activation analysis, gamma-ray spectrometry and least-squares methods developed for other nuclides have been applied to redetermine the decay constants of ^90mY and ⁹⁷Ru. Several irradiated samples of each pure element were counted continually through several half-lives, acquiring up to 100 spectra with good statistics. Photopeaks from each nuclide were evaluated with a region of interest integration routine, and each data point corrected for rate-related losses before fitting to an exponential function by a nonlinear least-squares procedure. The half-lives found are 3.204 ± 0.008 h for ^90mY and 2.836 ± 0.014 days for ⁹⁷Ru, where the uncertainty is an estimate of the 95 % confidence level. Although (by Chi squared criteria) the fits to the decay curves were very good and the uncertainty of the resulting half-lives an order of magnitude smaller than the previously published uncertainty, care is necessary before assigning a best value with a totally defensible uncertainty to the overall data set.     

Use of the Sorben-Tec system for rapid dosimetric evaluation of <Superscript>222</Superscript>Rn level in drinking water

The Sorben-Tec system was tested for rapid dosimetric evaluation of ²²²Rn level in drinking water in domestic conditions using a dosimeter of beta radiometer as a measurement equipment. It was shown that the method is cheap, rapid and very simple, therefore it can be used by population for rapid radiation safety assessment of drinking water. The sorption-active Sorben-Tec system contains iron hexacyanoferrate allowing for separation of approximately 40% of both ²¹⁴Pb and ²¹⁴Bi, the short-lived decay products of ²²²Rn from 5 L water sample. It was assessed that the Sorben-Tec system provides detection limits of radon in 5 L water samples of 35–40 and 10 Bq L^?1 for dosimetric and radiometric measurements respectively. The total time consumption of analysis does not exceed 1 h excluding the time for ²¹⁴Pb and ²¹⁴Bi ingrowth in the water sample (min. 3 h). Due to an insignificant sorption of radon, it is possible to reuse spent Sorben-Tec system again 3–4 h after previous analysis.     

Separation of heavier rare earths from neutron irradiated uranium targets

A radiochemical method is described for the separation of heavier rare earths from the fission of uranium. The method is particularly suitable for the separation of low yield (10⁻⁵%–10⁻⁷%), highly asymmetric rare earth fission products viz.^179,177Lu,¹⁷⁵Yb,¹⁷³Tm,^172,171Er,¹⁶⁷Ho and^161,160Tb in the neutron induced fission of natural and depleted uranium targets. Additional separation steps have been incorporated for decontamination from²³⁹Np (an activation product) and^93-90Y (a high fission-yield product) which show similar chemical behaviour to rare earths. Separation of individual rare earths is achieved by a cation exchange method performed at 80°C by elution with α-hydroxyisobutyric acid (α-HIBA).     

Application of low energy gamma-spectrometry in rapid actinide analysis for emergency preparedness

For preparedness purposes, a fast and reliable method is essential to quickly assess radioactive fallout in the environment. The rapid determination of certain nuclides such as alpha-emitting actinides is necessary to make initial environmental and agricultural advisories. Therefore, a method using a preconcentration resin and low energy gamma-spectrometry was developed to a fast determination of certain nuclides in soil samples. The preconcentration resin allows samples to be partially purified and then directly measured by gamma-spectrometry without further extraction or separation. The initial gamma-measurement provides fast and accurate determination of certain nuclides such as ²⁴¹Am and ²³⁵U which are normally analyzed by alpha-spectrometry, but require additional time-consuming purification and separation steps. After gamma-spectrometry, the sample may be further processed and analyzed by traditional methods to determine actinides or other nuclides more precisely.     

The ambient gamma dose-rate and the inventory of fission products estimations with the soil samples collected at Canadian embassy in Tokyo during Fukushima nuclear accident

In this study, soil samples were collected at Canadian embassy in Tokyo (about 300 km from Fukushima) on 23 March and 23 May of 2011 for purposes of estimating concentrations of radionuclides in fallout, the total fallout inventory, the depth distribution of radionuclide of interest and the elevated ambient gamma dose-rate at this limited location. Some fission products and actinides were analyzed using gamma-ray spectrometry, alpha spectrometry and liquid scintillation counting. The elevated activity concentration levels of ¹³¹I, ¹³²I, ¹³⁴Cs, ¹³⁷Cs, ¹³⁶Cs, ¹³²Te, ¹²⁹mTe, ¹²⁹Te, ¹⁴⁰Ba and ¹⁴⁰La were measured by the gamma-ray spectrometer in the first sample collected on 23 March. Two months after the accident, the ¹³⁴Cs and ¹³⁷Cs became only detectable nuclides. A mass relaxation depth of 3.0 g/cm² was determined by the activities on the depth distribution of ¹³⁷Cs in a soil core. The total fallout inventory was thus calculated as 225 kBq/m² on March sampling date and 25 kBq/m² on May sampling date. The ambient gamma dose-rates in the sampling area estimated by the fallout fission products inventory and ¹³⁷Cs depth distribution ranged from 184 to 38 nGy/h. There was no detectable americium or plutonium in the soil samples by alpha spectrometry. Although ⁹⁰Sr or ⁸⁹Sr were detected supposedly as a result of this accident, it was less than the detection limit, which was about 0.4 Bq/kg in the soil samples.     

Mass–yield distributions of fission products from photo-fission of natPb induced by 50–70 MeV bremsstrahlung

The mass–yield distributions of various fission products have been determined in the 50-, 60- and 70-MeV end point bremsstrahlung induced fission of ^natPb using off-line γ-ray spectrometric technique in the electron linac at Pohang Accelerator Laboratory, Korea. The mass–yield distributions are symmetric with average mass of 102.34, 102.25 and 102.03 and FWHM of 21, 22 and 23 mass unit, respectively. From the present data and literature data in the 50–85 MeV bremsstrahlung induced fission of ²⁰⁹Bi the following observations were obtained: (i) The average masses of the yield distributions in the 50–85 MeV bremsstrahlung induced fission of ^natPb and ²⁰⁹Bi are around 102.25 ± 0.25 and 103 ± 0.5, respectively. (ii) The FWHM of the mass–yield distributions increases from about 21 mass units at 50 MeV to 23 mass units at 70–85 MeV, which is explained from the point of increase in multi-chance fission probability with increasing excitation energy. (iii) Within the bremsstrahlung energy range of 50–85 MeV, the role of nuclear structure effect in the mass–yield distribution was observed in the photo-fission of ²⁰⁹Bi, whereas it was not seen in case of ^natPb. This may be due to the presence of so many isotopes in ^natPb unlike mono-isotopic ²⁰⁹Bi.     

Modified 2-window approach for rapid determination of 90Sr and 90Y at low quench level using liquid scintillation counting

⁹⁰Sr is a product of nuclear fission, the radioactivity of which can be determined by liquid scintillation counting (LSC). Because the LSC spectra of ⁹⁰Sr and its daughter ⁹⁰Y overlap each other, the following methods are usually used: (1) measuring immediately after ⁹⁰Sr/⁹⁰Y separation; (2) waiting to reach radioactive equilibrium; (3) adopting the conventional 2-window approach; and (4) using the spectra deconvolution technique. The first one requires ⁹⁰Sr/⁹⁰Y separation and immediate measurement; the second one is time-consuming; the third one is valid only for samples with the same quench level as the calibration standard; the last one is somewhat complicated, and in some cases it is not convenient to export the experimental data to some deconvolution software. Therefore, we have developed a modified 2-window approach to rapidly determine ⁹⁰Sr and ⁹⁰Y in either equilibrium or disequilibrium at low quench level. The key modification of the approach is to provide an LSC spectrum of pure ⁹⁰Y with the same quench level as the sample to be determined. This modification eliminates the need to conduct ⁹⁰Sr/⁹⁰Y separation for the sample itself, to prepare the quench curves, and to fit the LSC spectra with some deconvolution software.