Sequential separation of <Superscript>99</Superscript>Tc, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe, <Superscript>90</Superscript>Sr and <Superscript>59/63</Superscript>Ni from radioactive wastes

A sequential separation procedure has been developed for the determination of ⁹⁹Tc, ⁹⁴Nb, ⁵⁵Fe, ⁹⁰Sr and ^59/63Ni in various radioactive wastes generated from nuclear power plants. Ion exchange and extraction chromatography were adopted for individual separation of the radionuclides. Precipitation was supplementarily utilized for both purification of the individual radionuclides and preparation of the radionuclide sources for use in a radioactivity measurement. The chromatographic separation behavior of the radionuclides both from the sample matrix metals and from one another was investigated using stable metals, Re (as a surrogate of ⁹⁹Tc), Nb, Fe, Sr and Ni. The validity of the procedure for reliability and applicability was evaluated by measuring the recovery of the metal carriers added to synthetic radioactive waste solutions. The recoveries by the chromatographic separation were in the range of 84.8 to 102.2% with 2s of less than 8.6%, the recoveries by the precipitation being in the range of 84.3 to 97.3% with 2s of less than 10.9%.     

Experimental and FLUKA study relevant to the production of <Superscript>124</Superscript>I through <Superscript>nat</Superscript>Te(p,xn) reaction at a medium-sized cyclotron

A radiochemical methodology for the determination of ⁹⁴Nb in low-level radioactive wastes from nuclear power plant was proposed. Although ⁹⁴Nb is a strong gamma emitter, its concentration in radioactive waste samples is usually several orders of magnitude lower than that of other β–γ emitters, whose emissions interferes in the detection of the emission lines of ⁹⁴Nb. The procedure involves acid digestion, precipitation, cation exchange chromatography by using Amberlite IRA120 resin, extraction chromatography by using TEVA resin to isolate the Nb and the gamma spectrometry to its measurement. The chemical yield was 70% in average. Samples of evaporator concentrate and spent resin were analyzed. For the samples of the evaporator concentrate, the results obtained were below L _D = 9.59 × 10^?4 Bq g^?1. For the spent resin an average result of 1.54 × 10² Bq g^?1 was obtained.     

Preparation of <Superscript>26</Superscript>Al, <Superscript>59</Superscript>Ni, <Superscript>44</Superscript>Ti, <Superscript>53</Superscript>Mn and <Superscript>60</Superscript>Fe from a proton irradiated copper beam dump

The station for pions cancer therapy was operated at PSI from 1980 to 1992. After a cooling time of 12 years it’s made of copper beam dump was cut and samples were taken for analytical purposes. The sampling collected about 500 g of high active copper chips that can be used for separation of exotic radionuclides. The analyses by gamma spectrometry, LSC and AMS showed main nuclides present to be ⁶⁰Co, ⁵⁴Mn, ²²Na, ⁶⁵Zn, ²⁶Al, ⁵³Mn, ⁵⁹Ni, ⁶³Ni, ⁵⁵Fe and ⁶⁰Fe and ⁴⁴Ti with a daughter nuclide ⁴⁴Sc. In the frame of ERAWAST project a procedure combining selective precipitation and ion exchange for the separation of the rare radionuclides from the copper beam dump was developed. The proposed separation procedure is easy for remote controlled implementation in a hot cell. The ion exchange separation of Ni, Al, Mg, Ti and Fe was complete and high decontamination factors for copper and cobalt were achieved. Based on the developed procedure a remotely controlled system for separation of exotic radionuclides from the copper chips was set up. The full scale system was installed in a hot cell where high activity levels can be handled. In order to evaluate the reliability and functionality of the system extensive tests have been done. During the test period 13.86 g in total of the proton irradiated copper beam dump were processed for separation of ²⁶Al, ⁵⁹Ni, ⁵³Mn, ⁴⁴Ti and ⁶⁰Fe. The results showed that the system was operational and the radionuclide separation was selective with high chemical yield. The procedure manages as well the generated liquid wastes containing high level of ⁶⁰Co activity.     

Extraction chromatographic separation of iron from complex liquid samples and the determination of <Superscript>55</Superscript>Fe

Summary Iron separation is described from liquid samples with a high concentration of ions that enables simple determination of ⁵⁵Fe. One of the described methods consists of iron precipitation from a large volume seawater by sodium hydroxide and/or ammonium carbonate and separation from other elements (Ca, Sr, Cu, Mg, etc.) on a TRU column with 4M HCl or 8M HNO₃. In the other procedure iron is separated directly from a mixture of seawater samples and HCl on a TRU column. In both methods, the iron recovery is almost 100%. After separation, ⁵⁵Fe is determined by counting with a liquid scintillation counter. The binding of Fe and Zn on TEVA, U/TEVA and TRU resins from seawater solutions of HCl and HNO₃depends on the type of the resin, concentration of acid and other ions. Iron and zinc can be separated from seawater on a U/TEVA column with 2M HCl.     

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.     

Separation and determination of <Superscript>90</Superscript>Sr in low- and intermediate-level radioactive wastes using extraction chromatography and LSC

A methodology for the determination of ⁹⁰Sr in low- and intermediate-level radioactive wastes from nuclear power plants is presented in this work. It is a part of a methodology developed for the sequential radiochemical separation of radionuclides difficult-to-measure directly by gamma spectrometry in these radioactive wastes. The separation procedure was carried out using precipitation and extraction chromatography with Sr Resin, from Eichrom and the ⁹⁰Sr was measured by liquid scintillation counting (LSC). Optimum conditions for the pretreatment, separation and LSC measurements were determined using simulated samples, which were prepared using standard solutions and carriers. The procedure showed to be rapid and achieved a good chemical yield, in the range 60–90%, and a detection limit of 6.0 × 10⁻⁴ Bq g⁻¹. The method was also tested by participation in a national intercomparison program, with aqueous samples, with good agreement of results.     

Measurement of <Superscript>90</Superscript>Sr in primary coolant of pressurized water reactor

The present procedure for analysing ⁹⁰Sr combines the use of a non radioactive carrier with high pressure ion chromatography separation, conductivity detection of the carrier and optimized external counting by liquid scintillation. This improvement with respect to traditional methods led to a more rapid and efficient purification stage. The present work proves that activities of ⁹⁰Sr as low as 3 Bq/L can be measured in highly contaminated pressurized water reactor primary coolant matrix without any observed radiochemical interference. The approach shows promise for the analysis of other emitters of low energy radiation, or isotopes subject to high background or matrix effects in a PWR primary coolant.     

Interaction of <Superscript>99</Superscript>Tc onto bentonite and influence of Fe(II) and (Ca,Mg) ion exchange in its interlayers

⁹⁹Tc is a redox active radionuclide, which is present as contaminant at a number of sites where nuclear fuel cycle operations have been carried out. The aim of our research was study of bentonite interaction with Fe and Fe²⁺ cation and these influences on the migration behavior. Radioanalytical and analytical methods were used for the concentration and chemical forms of ⁹⁹Tc determination. It was found that Fe²⁺–Ca²⁺, Mg²⁺ ion exchange should be considered as a process of reaction among the corrosion products of carbon steel container and bentonite in the environment of radioactive waste repository.     

Determination of <Superscript>237</Superscript>Np, <Superscript>93</Superscript>Zr and other long-lived radionuclides in medium and low level radioactive waste samples

The majority of long-lived radionuclides produced in the nuclear fuel cycle can be regarded as “difficult-to-measure” nuclides, hence chemical separation is needed before the nuclear measurement of them. A combined radiochemical procedure that enables the simultaneous determination of some “difficult-to-measure” nuclides in medium and low level radioactive wastes has been developed in our laboratory. Recently, this method has been extended for determination of ²³⁷Np and ⁹³Zr. ²³⁷Np and ⁹³Zr are pre-concentrated by co-precipitation on iron(II) hydroxide and zirconium oxide, separated by extraction chromatography using UTEVA, and measured by inductively coupled plasma mass spectrometry (ICP-MS). As even traces of polyatomic ions and isotopes at m/z 237 or 93 cause considerable interferences during ICP-MS detection, a purification step by extraction chromatography was needed. Analyzing real samples (evaporation concentrates of a nuclear power plant) 66–99% and 31–99% chemical yields were achieved for Np and Zr, respectively.     

Atmospheric depositional fluxes of cosmogenic <Superscript>32</Superscript>P, <Superscript>33</Superscript>P and <Superscript>7</Superscript>Be in the Sevastopol region

⁹⁰Y was separated from ⁹⁰Sr using an extraction chromatographic resin consisting of 4, 4′(5′)-bis-t-butylcyclohexano-18-crown-6 (DtBuCH₁₈C₆), 1-ethyl-3-methylimidazolium bis(trifluoromethanesulfonyl) amide (C₂mimNTf₂), and a polymer (Amberlite XAD-7). Ionic liquid was introduced into the column to improve the separation efficiency. The column showed an excellent performance for the separation of Y from Sr. After the separation, the ratio of ⁹⁰Sr/⁹⁰Y was <2.0 × 10^?5; the column was recycled for >18 times. This study provides preliminary results on columns to produce ⁹⁰Y with a high purity in radiopharmaceuticals.     

PC-controlled radiochemistry system for preparation of no-carrier-added <Superscript>64</Superscript>Cu

A homemade modular PC-controlled radiochemistry system, used to separate the no-carrier-added ⁶⁴Cu from irradiated electroplated solid cyclotron target ⁶⁴Ni layers was developed. Upon irradiation the target layer is dissolved in diluted nitric acid, followed by selective liquid–liquid extraction of the ⁶⁴Cu. The final purification step is achieved making use of an anion exchange column. All the separation procedures can be carried out remote-controlled with a minimized risk of operator errors within less than 2 h, with excellent processing yield (>95%).     

Separation of bulk Y from 89Y(n,p) produced 89Sr by extraction chromatography using TBP coated XAD-4 resin

⁸⁹Sr was produced using the ⁸⁹Y(n, p) ⁸⁹Sr reaction by irradiating yttria target in the fast breeder test reactor (FBTR). An analytical scale procedure was standardized for the removal of the bulk target yttrium by solvent extraction using the tri-n-butyl phosphate (TBP). The final purification of ⁸⁹Sr source was carried out by ion exchange chromatography. However, extraction chromatography is preferred to solvent extraction due to its low waste generation and ease of operation. This paper reports the separation of Sr from bulk Y and other radioactive impurities produced during irradiation by extraction chromatography using TBP coated XAD-4 resin. Initially the separation procedure was standardized using ⁸⁵Sr and ⁸⁸Y tracers. The ⁸⁹Sr in the dissolver solution of the irradiated yttria target was separated under the same standardized conditions. The study established the separation of Sr from Y in the dissolver solution of the irradiated target yttria by the TBP coated XAD-4 column. However the evaluation of the column after every use for about three separation studies exhibited the reduction in its breakthrough capacity for yttrium.     

Migration processes of <Superscript>137</Superscript>Cs and <Superscript>90</Superscript>Sr in compartments of a lake ecosystem

Summary The dispersion of radioactive substances in the environment following nuclear weapon tests in atmosphere since 1954 and accidents to nuclear plants, like that in Chernobyl in 1986, have allowed us to study the migration processes of some radionuclides in complex ecosystems such as lakes are. In the present paper the behavior of ¹³⁷Cs and ⁹⁰Sr in different compartments of the Monterosi Lake (central Italy) was assessed. The ¹³⁷Cs concentration was measured in lake water as well as sediment, stream water, aquatic plant and fish samples. ⁹⁰Sr concentration in water and sediments was also determined. A total inventory of 4206±76 Bq ^. m^-2 and 958±79 Bq ^. m^-2 (on 27/6/01) has been found for ¹³⁷Cs and ⁹⁰Sr, respectively. The experimental data presented here allow to calibrate theoretical models predicting the temporal trend of radionuclide concentration in similar ecosystems. Moreover, information on cesium and strontium migration processes can be extended to other pollutants having similar environmental behavior.     

Automated radiochemical analysis of total <Superscript>99</Superscript>Tc in aged nuclear waste processing streams

Summary A fully automated analysis procedure and instrument for the measurement of total ⁹⁹Tc in aged nuclear waste has been developed. The overall analysis approach is based on a fully automated wet radiochemical analysis method. Microwave-assisted sample oxidation is used prior to a chemical separation step in order to oxidize all of the non-pertechnetate species to pertechnetate. Separation of the pertechnetate from interfering radioactive and stable matrix species is carried out using an anion-exchange column. The separated ⁹⁹Tc is quantified using a flow-through solid cell scintillation detector. The instrument is capable of an analysis time of <13 minute per sample with a detection limit of 2000 dpm/ml. Nuclear waste samples from the Hanford site with a high content of non-pertechnetate species were successfully analyzed using this method.     

Recovery of <Superscript>65</Superscript>Zn from waste solutions from gallium targets at Brookhaven Linac Isotope Producer

A procedure for recovery of ⁶⁵Zn from acidic Ga-rich solutions generated in ⁶⁸Ge production process was developed and implemented in a “hot cell” environment. A one step simple separation was carried out on anion exchange column filled with Biorad AG1-X8 resin. With ⁶⁵Zn recovery yields of more than 95% this procedure is currently routinely used at Brookhaven Linac Isotope Producer.     

Combination of developed in-house method and application of Eichrom Sr resin to determine the radioactivity of <Superscript>90</Superscript>Sr in environmental sample

A method is described for the determination of ⁹⁰Sr in environmental samples using combination of developed in-house method, Eichrom Sr resin and Beta Counter. Strontium was efficiently, rapidly and simply separated from Ca and other interfering matrix components by Eichrom Sr resin. All the results in general showed good accuracy, high precision, reliable and in good agreement between these two measured and certified value of SRM (i.e. IAEA-375, IAEA-326, IAEA-152 and IAEA-414). As a whole, the procedure described in this work notably improves conventional methods in particular concerning the time needed, sample volume, safer and others. Thus, the introduced method was successfully performed and will be applied to actual sample for the determination of ⁹⁰Sr in different environmental materials such as soil, sediments, milk, biological sample, water etc.     

Behavior of <Superscript>134</Superscript>Cs, <Superscript>90</Superscript>Sr,and <Superscript>238</Superscript>Pu in different Syrian soils

The experiment aimed to evaluate the vertical migration of ¹³⁴Cs, ⁹⁰Sr and ²³⁸Pu in the main types of Syrian soils; entisol, inceptisol, alluvial (rock outcrops) and gypsiferous soils, using soil columns through which the aqueous solution of the radionuclides percolated. The results show that the vertical migration of the studied radionuclides through the soil profile depend on the radionuclide and the soil type. More than 97% of ¹³⁴Cs and ²³⁸Pu concentrated in the upper 2 cm of the entisol, inceptisol, and alluvial soils, whereas only 46.2% to 68.6% of the ⁹⁰Sr was retained in the upper 2 cm of these soils. The vertical migration of the studied radionuclides in the gypsiferous soil was different from the other soils. The distribution of the radionuclides in the gypsiferous soil was irregular through the soil profile and reached the deeper layer of the soil. This may be due to its physical characteristics; poor structure stability, high permeability and low retention capacity.     

A novel electrochemical <Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generator

A novel electrochemical process to avail clinical grade ^99mTc from (n,γ)⁹⁹Mo has been demonstrated. The electrochemical parameters were optimized to maximize the ^99mTc yield with minimal ⁹⁹Mo contamination. ⁹⁹Mo/^99mTc generators containing up to 29.6 GBq (800 mCi) ⁹⁹Mo were developed and their performance were extensively evaluated for 10 days without changing the operating conditions. Very high radioactive concentration of ^99mTcO₄ ⁻ of acceptable quality, commensurate with hospital radiopharmacy requirements could be availed from the system with >90% yield. The compatibility of the product for the formulation of ^99mTc labeled radiopharmaceuticals such as ^99mTc-DMSA and ^99mTc-EC was found to be satisfactory in terms of high labeling yields. The proposed route represents an important step for enhancing the scope of accessing clinical grade ^99mTc from low specific activity (n, γ)⁹⁹Mo.     

Validation of <Superscript>90</Superscript>Sr separation method using molecular recognition product AnaLig<Superscript>®</Superscript> Sr01 gel and extraction chromatography Sr<Superscript>®</Superscript> resin

Two separation techniques for strontium determination using AnaLig^® Sr01 molecular recognition technology and extraction chromatography Sr^®  resin were tested. The methods performance was investigated by analysis of NPL (High Alpha–Beta 2003) intercomparison sample. The results obtained for both procedures were compared in terms of activities and recoveries. Data analysis proved a good agreement with the reference values. The AnaLig^® Sr01 separation method for ⁹⁰Sr determination was successfully validated with the same performance as the Sr^® resin method.     

A selective separation method for <Superscript>93</Superscript>Zr in radiochemical analysis of low and intermediate level wastes from nuclear power plants

The zirconium isotope ⁹³Zr is a long-lived pure β-particle-emitting radionuclide produced from ²³⁵U fission and from neutron activation of the stable isotope ⁹²Zr and thus occurring as one of the radionuclides found in nuclear reactors. Due to its long half life, ⁹³Zr is one of the radionuclides of interest for the performance of assessment studies of waste storage or disposal. Measurement of ⁹³Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. A radiochemical procedure based on liquid–liquid extraction with 1-(2-thenoyl)-3,3,3-trifluoroacetone in xylene, ion exchange with Dowex resin and selective extraction using TRU resin has to be carried out in order to separate zirconium from the matrix and to analyze it by liquid scintillation spectrometry technique (LSC). To set up the radiochemical separation procedure for ⁹³Zr, a tracer solution of ⁹⁵Zr was used in order to follow the behavior of zirconium during the process by γ-ray spectrometry through measurement of the ⁹⁵Zr. Then, the protocol was applied to low level waste (LLW) and intermediate level waste (ILW) from nuclear power plants. The efficiency detection for ⁶³Ni was used to determination of ⁹³Zr activity in the matrices analyzed. The limit of detection of the 0.05 Bq l⁻¹ was obtained for ⁶³Ni standard solutions by using a sample:cocktail ratio of 3:17 mL for OptiPhase HiSafe 3 cocktail.