Die Photokernreaktionen der Edelgase mit 15-MeV-bremsstrahlung

The rare gases Xe, Ar, Kr and Ne under normal pressure in quartz ampoules were irradiated with 15-MeV-Bremsstrahlung for two hours in the Mikrotron (23–25 μA) of the Joint Institute of Nuclear Research Dubna. Identification of the generated radionuclides was carried out by means of measured γ-energies and half-life. The rare gases Xe, Kr and Ar except Ne are generating with 15-MeV-Bremsstrahlung several radionuclides. Relative sensitivities and detection limits for determination of rare gases are estimated and possibilities of applications are discussed.      

Determination of long-lived fission products and actinides in Savannah River Site HLW sludge and glass for waste acceptance

Savannah River Site (SRS) is immobilizing the radioactive, high-level waste sludge in Tank 51 into a borosilicate glass for disposal in a geologic repository. A requirement for repository aceeptance is that SRS report the concentrations of certain fission product and actinide radionuclides in the glass. This paper presents measurements of many of these concentrations in both Tank 51 sludge and the final glass. The radionuclides were measured by inductively coupled plasma mass spectrometry and α, β, and γ counting methods. Examples of the radionuclides are⁹⁰Sr,¹³⁷Cs,²³⁸U and,²³⁹Pu. Concentrations in the glass are 3.1 times lower due to dilution of the sludge with a nonradio-active glass forming frit in the vitrification process. Results also indicated that in both the sludge and glass the relative concentrations of the long lived fission products insoluble in caustic are in proportion to their yields from the fission of²³⁵U waste in the SRS reactors. This allowed the calculation of a fission yield scaling factor. This factor in addition to the sludge dilution factor can be used to estimate concentrations of waste acceptance radionuclides that cannot be measured in the glass. Examples of these radionuclides are⁷⁹Se,⁹³Zr, and¹⁰⁷Pd.     

Chemotopological study of positron emitter radionuclides used in PET diagnostic imaging: physical, physico-chemical, dosimetric, quantum and nuclear properties

We carried out a chemotopological study of two groups of positron emitter radionuclides of current and potential use in positron emission tomography (PET) diagnostic imaging. The aim was to look for potential β⁺ radionuclides not yet in use for PET imaging, taking into account the similarities of these radionuclides with radionuclides already used in PET imaging. The similarity was studied through physical, physico-chemical, dosimetric, quantum and nuclear properties of the radionuclides. We applied cluster analysis (two similarity functions and three grouping methodologies) and generated six dendrograms. One dendrogram was selected in each group, which was used to build up a basis for a topology. From the calculation of topological properties of several subsets of interest, we propose β⁺ radionuclides ⁴⁷V, ⁴⁸V, ⁶³Zn, ⁷⁰As, ⁹⁰Nb, ¹⁰⁶Ag, ¹¹⁵Sb, ¹¹⁶Sb, ¹²⁰Sb, ¹³⁰Cs, ¹³⁴La and ¹⁴⁰Pr, as potential new radionuclides of use in PET imaging.     

External dose-rates for natural gamma emitters in soils from an agricultural land in West Anatolia

The surveys of natural gamma-emitting radionuclides in soils from three basins of West Anatolia intensively used for agricultural purposes were conducted during 1998–2003. In the present study, part of the survey, the activity concentrations of ²³⁸U, ²³²Th and ⁴⁰K in the soil samples from 43 sites distributed all over the agricultural land known as Büyük Menderes basin were determined by scintillation gamma spectrometry. The average activity concentrations and ranges of the relevant radionuclides in the soils were as follows: ²³⁸U was 29 (7–84); ²³²Th, 22 (10–48) and ⁴⁰K, 464 (100–864) Bq kg⁻¹. The corresponding absorbed dose rates in air from all those radionuclides were in the range of 17–81 nGy h⁻¹ with a mean value of 46 nGy h⁻¹ and did not exceed the world-wide average values. All dosimetric calculations were performed based on the guidance of UNSCEAR 2000 report [1].     

Application of radiochemical methods and dispersion model in the study of environmental pollution in Brazil

Methodologies for analysis of anthropogenic and natural radionuclides in marine samples were developed and applied in environmental samples. Results of systematic measurements of these radionuclides have showed that artificial radioactivity levels are in agreement with the values from the regions not affected directly by nuclear accidents or nuclear reprocessing plant discharges and are due to the global fallout. The average concentration of¹³⁷Cs is of 1.4 Bq·m⁻³ in seawater, ranges from 13 to 220 mBq·kg⁻¹ in fish, and from 0.4 to 1.8 Bq·kg⁻¹ for sediments.⁹⁰Sr levels in seawater are of 1.8 Bq·m⁻³ and in fish vary from 19 to 75 mBq·kg⁻¹. Sediments present concentrations of⁹⁰Sr lower than 0.8 Bq·kg⁻¹ and for^239+240Pu of 0.03 to 0.18 Bq·kg⁻¹.²¹⁰Po levels in fish range from 0.5 to 5.3 Bq·kg⁻¹. The data generated represent reference values for our country and are used to estimate the intake levels of these radionuclides by consuming of marine products.     

Leaching of uranium,radium and thorium from vertisol soil by ground water

Shallow land burial is routinely used for the disposal of low-level radioactive waste. Natural processes causing leaching of radionuclides can lead to contamination of surrounding ground water and soil by the radionuclides. The comparative leachability of radionuclides U_(nat), ²²⁶Ra, ²²⁸Ra and Th_(nat) from the soil of a radioactive waste disposal site, by ground water was evaluated. The probability of leaching was obtained in the following order Ra (≈77%) > U (≈40%) > Th (≈20%). Observed ratios (OR) were calculated to correlate leachability of radionuclides to that of major cations Ca²⁺ and Mg²⁺. The leaching of the radionuclides was seen to be dependent on Ca²⁺ and SO₄²⁻ leached from the soil. This study provides sitespecific leachability of radionuclides, that can be used as indicator of the tendency for migration or retention in soil. It can play an important role during an unforeseen accident like breach of containment at the waste disposal site leading to contamination of soil and ground water and causing hazard to public via drinking water route.     

Alpha spectroscopic analysis of actinides (Th,U and Pu) after separation from aqueous solutions by cation-exchange and liquid extraction

The radioactivity concentration of ²³⁶Pu, ²³²U and ²²⁸Th in aqueous samples has been determined by means of alpha spectroscopy after chemical separation and pre-concentration of the radionuclides by cation exchange and liquid–liquid extraction using the Chelex-100 resin and 30% TBP/dodecan, respectively. Method calibration using a ²³⁶Pu standard solution containing the daughter radionuclides results in a detector efficiency of 18% and in a chemical recovery for cation-exchange which is (30 ± 7)%, (90 ± 5)% and (20 ± 5)% for plutonium, uranium and thorium, respectively. The chemical recovery for liquid–liquid extraction is found to be (60 ± 7)%, (50 ± 5)% and (70 ± 5)%, for plutonium, uranium and thorium, respectively. The differences in the efficiencies can be ascribed to the oxidation states, the different actinides present in solution. Taking into account that the electrodeposition of the radionuclides under study is quantitative, the total method efficiency is calculated to be (18 ± 15)%, (46 ± 7)% and (15 ± 5)%, for plutonium, uranium and thorium, respectively, at the mBq concentration range. The detection limit of the alpha spectrometric system has been found to be 0.2 mBq/L, suggesting that the method could be successfully applied for the radiometric analysis of the studied radionuclides and particularly uranium in aqueous samples.     

A 228Ra-212Pb tandem generator for potential application in biomedical studies

A new tandem generator system (cation-exchange generator column coupled to anion-exchange), based on the ²²⁸Ra as starting radionuclide, for ²¹²Pb/²¹²Bi production in the solutions suitable to direct application in biomedical investigations has been developed. Optimum conditions have been found for retention of Ra, Th and Ac parent radionuclides on the cation-exchange column and repeated elution of Pb with subsequent concentration on anion-exchange column using HBr and mixed HBr/CH₃OH solutions. It was shown that Pb/Bi could be eluted from anion — exchange column with a small volume of EDTA, DTPA, NaCl solutions or heated H₂O. The advantages of the tandem generator system for production of short-lived Pb/Bi radionuclides and their application for biomedical studies are discussed.     

The transfer of natural (226Ra,?232Th, 40K) and artificial (137Cs, 90Sr) radionuclides from soil to the wood of a Scots pine (Pinus sylvestris L.)

On the Stabatishke site near to Ignalina Nuclear Power Plant a near surface repository for low and intermediate-level short-lived radioactive waste is being constructed to store the waste which was produced during the decommissioning of the Ignalina Nuclear Power Plant. A possible spread of radionuclides from the near surface repository and the radiation level of expression are one at the most important aspects while evaluating the safety of the repository. The article examines the specific activity of artificial (¹³⁷Cs, ⁹⁰Sr) and natural (²²⁶Ra, ²³²Th, ⁴⁰K) radionuclides in the soil of Stabatishke site of Ignalina NPP in pinewood; the spread of these radionuclides is determined in the system “soil-tree”. The change of long term soil pollution with artificial radionuclides was estimated. This is a background soil pollution which had been there before the exploitation of the near surface waste repository. The transfer factors of natural (²²⁶Ra, ²³²Th, ⁴⁰K) and artificial (¹³⁷Cs, ⁹⁰Sr) radionuclides from the 0-to-20-cm soil layer to different annual pinewood rings were measured. After estimation of the position of pine roots in accordance to the soil vertical and the annual change of the plant biomass, the transfer factor of ¹³⁷Cs movement to different annual pinewood rings from the separate 0 to 20-cm soil layers was determined.     

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.     

Direct and precise determination of environmental radionuclides in solid materials using a modified Marinelli beaker and a HPGe detector

A simple but precise detection method was studied for the determination of natural radionuclides using a conventional HPGe detector. A new aluminium beaker instead of a plastic Marinelli beaker was constructed and examined to reach radioactive equilibrium conditions between radon and its daughter elements without the escape of gaseous radon. Using this beaker fifteen natural radionuclides from three natural decay series could be determined by direct γ-ray measurement and sixteen radionuclides could be determined indirectly after radioactive equilibrium had been reached. Analytical results from ground water were compared with those from conventional α spectroscopy and the results agreed well within 12% difference. Nitrogen gas purge was used to replace the surrounding air of the detector to obtain a stable background and reducing the interference of radon daughter nuclides in the atmosphere. The use of nitrogen purging and the aluminium Marinelli beaker results in an approximately tenfold increase of sensitivity and a decrease of the detection limit of ²²⁶Ra to about 0.74 Bq kg^–1 in soil samples. Received: 12 October 2000 / Revised: 12 December 2000 / Accepted: 13 December 2000     

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.     

Optimization study on sample pretreatment of spent fuel storage rack

In order to evaluate radionuclide inventories as an essential item for the permanent disposal of spent fuel storage racks, chemical conditions for a sample pretreatment of a spent fuel storage rack were studied. Especially, the surface microstructure and the radionuclide distributions for the spent fuel storage rack were investigated by using a SEM–EDX and γ-spectrometer for minimizing the matrix effect which could affect a chemical separation process of some β-emitting radionuclides. The samples were pretreated with a mixed solution of 5 M HCl and 2 M HNO₃ by an ultrasonic surface leaching method. Some radionuclides in the raw racks showed the radioactivity of 10²–10³ Bq for about 10 g of sample weight. From the sample pretreatment, it was confirmed that almost all radionuclides in the rack were completely extracted from the rack when the dissolved thickness of the rack became a maximum 15 μm by the ultrasonic surface leaching method. The established pretreatment method was applied for all spent fuel storage rack generated from Korean NPPs to determine the scaling factor. The radioactivities of ⁶⁰Co and ¹³⁷Cs radionuclides in the pretreated solutions were in the range of 4.9E−1~1.5E+2 and 1.2E−1~9.0E+0 Bq/g, respectively.     

Identification and determination of natural radioactive impurities in rare earth chlorides

²²⁷Ac,²²⁸Th,²²⁶Ra,²¹⁰Po and²¹⁰Pb can be present at rare earth chlorides. A radiochemical procedure is presented for the identification and determination of natural radioactive impurities in rare earth chlorides. The determination limits for these radionuclides were 1.5·10^–4 to 3·10^–1 Bq/g. The relative standard deviations for determining 10^–2 Bq/g radionuclides were usually less than ±7%.     

Distribution of <Superscript>137</Superscript>Cs, <Superscript>40</Superscript>K, <Superscript>232</Superscript>Th and <Superscript>238</Superscript>U in coastal marine sediments of Margarita Island,Venezuela

This work presents the results of ¹³⁷Cs, ⁴⁰K, ²³²Th and ²³⁸U concentration (Bq kg⁻¹) values in coastal marine sediments collected from 38 sites along the coastline of the island of Margarita, Venezuela. The purpose was to determine baseline values for these radionuclides in surface marine sediments and to detect if there were any anomalously high concentration values. Only three of the 38 sediments analyzed had measurable values above the detection limit of 0.9 Bq kg⁻¹ for ¹³⁷Cs and the highest only being 1.4 Bq kg⁻¹. While, the concentration (Bq kg⁻¹) ranges for the primordial radionuclides, ⁴⁰K, ²³²Th and ²³⁸U were as follows: 12.2–211.7, <1.5–9.8 and <4.4–20.7, respectively. These concentration ranges for the primordial radionuclides can be considered as baseline values for surface marine sediments for areas that are considered not polluted by man or contaminated by nature. Finally, the concentration range of ¹³⁷Cs can also be employed as baseline values, which only seem to have been the result of the atmospheric testing of nuclear weapons in the past.     

A simple method for the activation analysis of mercury in fish by using iodinated resins

After irradiation and decay periods, samples are dissolved in 20 min in a sulfuric-nitric acid mixture containing a metavanadate salt. The aqueous phase is neutralized and passed through an iodinated resin which selectively isolates mercury from all other interfering radionuclides. The activity of¹⁹⁷Hg is determined by γ-spectrometry using a thin Na(Tl) detector. A sensitivity limit of 2.5 ppb^* is obtained and the relative standard deviation of the method is 6.7% at a level of 85 ppb.     

Study of natural radionuclides and its radiation hazard index in Malaysian building materials

The main component of most building materials in Malaysia is rock. All rocks are known to contain natural radionuclides such as ²³⁸U and ²³²Th series as well as ⁴⁰K. In order to estimate the radiological impact to the dweller, the level of radionuclides present in various building materials available in Malaysia were analyzed using γ-spectrometry. The radiation hazard indexes were calculated based on the above results. The results showed that the activity concentration of natural radionculides ²³⁸U, ²³²Th, ⁴⁰K was between 19.0–42.2 Bq/kg, 16.5–28.8 Bq/kg and 243.3–614.2 Bq/kg, respectively. On the whole, the radionuclides concentration was still below the global average of 50 Bq/kg, 50 Bq/kg and 500 Bq/kg for ²³⁸U, ²³²Th, and ⁴⁰K, respectively The radiation hazard indexes of the building materials were also lower than the maximum value suggested.     

New developments in the production of theranostic pairs of radionuclides

A brief historical background of the development of the theranostic approach in nuclear medicine is given and seven theranostic pairs of radionuclides, namely ^44gSc/⁴⁷Sc, ⁶⁴Cu/⁶⁷Cu, ⁸³Sr/⁸⁹Sr, ⁸⁶Y/⁹⁰Y, ¹²⁴I/¹³¹I, ¹⁵²Tb/¹⁶¹Tb and ¹⁵²Tb/¹⁴⁹Tb, are considered. The first six pairs consist of a positron and a β^?-emitter whereas the seventh pair consists of a positron and an α-particle emitter. The decay properties of all those radionuclides are briefly mentioned and their production methodologies are discussed. The positron emitters ⁶⁴Cu, ⁸⁶Y and ¹²⁴I are commonly produced in sufficient quantities via the (p,n) reaction on the respective highly enriched target isotope. A clinical scale production of the positron emitter ^44gSc has been achieved via the generator route as well as via the (p,n) reaction, but further development work is necessary. The positron emitters ⁸³Sr and ¹⁵²Tb are under development. Among the therapeutic radionuclides, ⁸⁹Sr, ⁹⁰Y and ¹³¹I are commercially available and ¹⁶¹Tb can also be produced in sufficient quantity at a nuclear reactor. Great efforts are presently underway to produce ⁴⁷Sc and ⁶⁷Cu via neutron, photon and charged particle induced reactions. The radionuclide ¹⁴⁹Tb is unique because it is an α-particle emitter. The present method of production of ¹⁵²Tb and ¹⁴⁹Tb involves the use of the spallation process in combination with an on-line mass separator. The role of some emerging irradiation facilities in the production of special radionuclides is discussed.     

Gas chromatographic determination of dissolved noble gases in liquids

A gas chromatographic technique for the determination of noble gas content in inorganic an organic liquids is described. Experimental data given to demonstrate the ability of the method indicate that the lower concentration limit of noble gases dissolved in 1 g of a liquid that can be determined by the described analytical technique is about of 6·10⁻³ cm³ (STP) of Kr and 2·10⁻³ cm³ (STP) of Xe. In addition, it is shown that this method together with the standard sample preparation applied is suitable for the study of gas solubilities in liquids.     

Concentration of natural radionuclides in and around Padubidri on the coastal Karnataka,India

The activity concentrations of natural radionuclides viz. ²³²Th, ²²⁶Ra and ⁴⁰K were measured in soil samples of Padubidri on the coastal Karnataka, the site for the proposed coal based thermal power station, using gamma-ray spectrometry to establish a baseline data on radioactivity levels in the environment of the region. The activity concentration of ²³²Th varies in the range of 38.5–115.5 Bq·kg⁻¹ with a mean value of 66.0 Bq·kg⁻¹, the activity concentration of ²²⁶Ra varies in the range 35.3–72.5 Bq·kg⁻¹ with a mean value of 53.5 Bq·kg⁻¹ and that of ⁴⁰K varies in the range of 307.5–550.9 Bq·kg⁻¹ with a mean value of 419.3 Bq·kg⁻¹. The radium equivalent activity varies in the range of 140.0–242.9 Bq·kg⁻¹ with a mean value of 180.2 Bq·kg⁻¹. The correlation between ²²⁶Ra and ⁴⁰K, ²³²Th and ⁴⁰K and ²²⁶Ra and ²³²Th was studied from the results of the concentration of these naturally occurring radionuclides. The calculated dose rates in air due to these naturally occurring radionuclides varies in the range of 66.0–110.0 nGy·h⁻¹ with a mean dose rate of 83.1 nGy·h⁻¹.