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The impact of low doses of ionising radiation on biological and environmental systems have been historically difficult to study. Modern biological tools have provided new methods for studying these mechanisms but applying these tools to a dose–response relationship may require refinement of dosimetric techniques that incorporate a detailed understand of radionuclide accumulation in biological cells, particularly when assessing the impact of low doses of ionising radiation. In this work Pseudomonas putida (KT2440) grown in liquid culture was exposed to low dose rates (10–20 mGy d−1) of 239Pu and 55Fe, both alone and in combination, for a period of 20 days, and the accumulation of 239Pu and 55Fe in cell pellets was analysed via liquid scintillation counting. The study also considered of cells grown with 239Pu and stable Fe (primarily 56Fe). In addition to the analysis of cell pellet and media samples, this work includes analysis of the radiological content of ribonucleic acid extraction samples to examine uptake of radionuclides. Results indicate that 239Pu inhibited the uptake of 55Fe, and that the presence of stable and radioactive isotopes of Fe in cultures may promote pathways for Fe accumulation that are used by 239Pu. The work herein provides foundational insight into future dosimetric models for our work with environmental bacteria.

Tritium extraction from radioactively contaminated cement mortar samples was performed using heating and liquid scintillation counting methods. Tritiated water molecules (HTO) can be present in contaminated water along with water molecules (H2O). Water is one of the primary constituents of cement mortar dough. Therefore, if tritium is present in cement mortar, the buildings and structures using this cement mortar would be contaminated by tritium. The radioactivity level of the materials in the environment exposed to tritium contamination should be determined for their disposal in accordance with the criteria of low-level radioactive waste disposal facility. For our experiments, the cement mortar samples were heated at different temperature conditions using a high-temperature combustion furnace, and the extracted tritium was collected into a 0.1 M nitric acid solution, which was then mixed with a liquid scintillator to be analyzed in a liquid scintillation counter (LSC). The tritium extraction rate from the cement mortar sample was calculated to be 90.91% and 98.54% corresponding to 9 h of heating at temperatures of 200 °C and 400 °C, respectively. The tritium extraction rate was close to 100% at 400 °C, although the bulk of cement mortar sample was contaminated by tritium.

Radiocarbon (14C) dating by liquid scintillation (LS) spectroscopy (also known as LS counting or LSC) provides an alternate method of 14C analysis where accelerator mass spectrometry (AMS) analysis is less desirable. The past, present, and future applications of the method are discussed.

A rapid analytical method for sequential separation and determination of the most significant α- and β-emitting radionuclidic impurities in 99Tc radiopharmaceutical raw material is presented. Successive HTiO co-precipitation steps were first conducted for effective removal of 99Tc to avoid its spectral interference on the determination of β-emitting nuclides by liquid scintillation counting. A sequential chromatographic method was applied for simultaneous separation and purification of actinide and lanthanide isotopes. A series of K99TcO4 solution spiked with known quantities of 238Pu, 239Pu, 241Pu, 237Np, 241Am, 244Cm and 147Pm were analyzed for method evaluation, demonstrating that the method would meet the analytical requirements for radionuclidic impurity evaluation of K99TcO4 raw material.

In this study, the dependence of TDCR Cerenkov counting technique on sample geometry, vial material and instrument type was studied. Various volume sets of color quenched samples spiked with yttrium-90, phosphorus-32, strontium-89, and bismuth-210 were prepared in 20-mL polyethylene scintillation (PE) vial, low-potassium borosilicate glass scintillation vial and 7-mL PE vial and then were measured by TDCR liquid scintillation analyzer (LSA), Hidex 300SL, Hidex 300SLL and LSA 3000. The results showed that in case of low color quench, dependence of TDCR Cerenkov counting technique on sample volume (in the volume range of 8–20 mL), vial type (7 and 20-mL) was negligible. The effect of vial material on the TDCR Cerenkov counting technique varied from decay energy of radionuclides and type of TDCR LSA. Additionally, TDCR Cerenkov curves of Hidex LSA agreed well with each other, and deviated from those of LSA 3000.

Responding to a radiological or nuclear incident may require assessing tens to hundreds of thousands of people for possible radionuclide contamination. The measurement of radioactive Sr is important because of its impact on people’s health. The existing analytical method for urine Sr-90 analysis using crown ethers is laborious and involves possible exposure to concentrated acids; therefore, this work is devoted to the development of the automated Sr-90 separation process, which became possible with the prepFast pre-analytical system (Elemental Scientific, Inc).

There is currently general concern over low-level radioactive wastewater from the development of nuclear industry. In this paper, a method based on an ultralow-level liquid scintillation spectrometer for measuring uranium radioactivity in low-level radioactive wastewater is proposed. This method can easily and quickly measure the radioactivity level of uranium in samples and can even distinguish the main isotopes of uranium. The liquid scintillation method directly provides results in units of radioactivity activity concentration, which are more convenient for comparison with relevant national standards to determine whether the emission standards are met. The lowest limit of detection of this method is 0.014 Bq l−1 within 600 min.

Tritium measurement from aqueous fraction of the wines of known vineyard can be a useful tool in reconstruction of the tritium activity in different areas and makes possible to date the time of the harvest of the grapes. The goal of this study was to develop a suitable method for purification of the aqueous phase from the wine samples, obtained from different type of wines, in order to determine tritium activity concentration using liquid scintillation counting. In this work the influence of the purification method applied to the aqueous samples resulted from wine distillation was assessed by deuterium and tritium measurements, content of organic carbon, pH and conductivity. The most reliable purification method used for wine aqueous fractions, according with the results obtained for pH, and total organic carbon content is chemical treatment followed by lyophilization.

Objectives. To define the levels of volume activity of tritium compounds and fraction of organically bound tritium in urine of chemical production workers of “Mayak Production Association” in present-day conditions; to identify the relationship between volume activity levels and professional occupation (department and profession). Material and methods. 245 urine samples from 171 workers of Mayak Production Association tritium production facility were collected in the period from 2017 to 2019. Volume activity of tritium compounds was measured by liquid scintillation method using spectrometer Quantulus-1220. The samples were distilled or dried and then combusted using an automatic preparation and oxidization system Sample Oxidizer A307. The “R” software was used for statistical analysis and for processing the measurement results and occupational factors. Chaddok’s scale was applied to determine the degree of correlation. The significance level was taken equal to 5%. Results. The value of total tritium volume activity in the urine and tritium volume activity in the water phase varied within 4 orders of magnitude (from ~30 Bq/dm3 to ~250 kBq/dm3). The value of volume activity of organically bound tritium in the urine varied within 2 orders of magnitude (from ~6 Bq/dm3 to ~3000 kBq/dm3). The fraction of organically bound tritium in the urine of the workers was within the range from 0,07% to 74%, and did not differ statistically significantly from lognormal distribution with parameters GM=2,7% and GSD=3,7. Very high rank correlation was detected between total volume activity of tritium compounds and tritium volume activity in the water phase in the urine. Noticeable rank correlations were detected between the total volume activity of tritium compounds in the urine and volume activity of organically bound tritium, as well as between tritium volume activity in the water phase and volume activity of organically bound tritium in the urine. The total tritium volume activity and tritium volume activity in the water phase in the urine of the workers of the 1st department and of the analytical laboratory of Mayak Production Association tritium production facility were statistically significantly higher than in the workers of the 2nd department according to median values. Statistically significant differences between medians of the total tritium volume activity in the urine and tritium volume activity in the water phase related to profession were observed only in the workers in the 1st department. Conclusion. Estimation of tritium volume activity in the water phase by the level of total tritium volume activity in the urine without sample preparation is possible with 95% reliability within limits of one order of magnitude towards the model value. Estimation of volume activity of organically bound tritium in the urine without sample preparation by the level of total tritium volume activity in the urine without sample preparation is possible with 95% reliability within limits of two orders of magnitude towards the model value. The effect of occupational factors to the levels of volume activity of tritium compounds in the urine of Mayak Production Association professional workers was detected.

Uranium is a radioactive element with a special presence in the rocks, waters, sediments, soils, and plants at the state of Chihuahua. The activity ratio of 238U/234U is used to explain the uranium transport by surface water and its deposition in arid environments. In this work, the activity concentration of natural U isotopes is determined by PERALS liquid scintillation and high-resolution alpha spectrometry (AS, Canberra camera 7401VR) in the Environmental Radiological Surveillance Laboratory (LVRA) at CIMAV, Chihuahua. Uranium is extracted from soils through the scintillating liquid extractor URAEX for PERALS, with chemical recovery (CR) of 80 - 85 %, and through the ion exchange resin UTEVA + electrode position by the Hallstadius method, with CR of 85 - 90 %, for AS. The procedures of 234U and 238U activity concentration (AC) determination in soils were validated by their application to the certified reference material IAEA-375. The resulting values were in the reference range of the certified or informative values. Both procedures were applied to 6 representative soil samples, with AC of the same order of magnitude or greater, and similar CR and compatible results. Both procedures are satisfactory for the purposes of LVRA research and in general.

The features of beta spectra standardization for modern liquid scintillation counting (LSC) us-ing Quantulus 1220TM alpha beta spectrometer are considered. The range of tasks is tritium meas-urement, beta-spectrometric determination of strontium-90, determination of strontium-90 by Che-renkov counting, determination of lead-210 in aerosol filters, radiocarbon studies. The purpose of the study was to summarize and analyze the peculiarities of the application of the methods of spectra standardization of beta-emitters in the study of radioactivity of environmental objects based on liquid scintillation counting. Materials and methods: liquid scintillation counting, LSC, Cherenkov counting, radiochemi-cal preparation of samples, spectra decomposition of beta-emitters. Results: Methods for spectra standardization of beta-emitters samples were analyzed in the study of 3H, 14C, 90Sr, 210Pb in environmental objects using a modern Quantulus 1220TM liquid scintil-lation spectrometer. The acceptability, effectiveness and scope of application of the methods of stand-ardization and spectra decomposition for beta-emitters for liquid scintillation counting and Cerenkov counting are grounded.

Purpose [18F]Fluoromethylcholine ([18F]FMCH) is a radiopharmaceutical used in positron emission tomography (PET) imaging for the study of prostate, breast, and brain tumors. It is usually synthesized in cyclotron facilities where 18F is produced by proton irradiation of [18O]H2O through 18O(p,n)18F reaction. Due to the activation of target materials, the bombardment causes unwanted radionuclidic impurities in [18O]H2O, that need to be removed during the radiopharmaceutical synthesis. Thus, the aim of this study is to quantify the radionuclide impurities in the 18F production process and in the synthesized [18F]FMCH, demonstrating the radionuclidic purity of this radiopharmaceutical. Methods Long-lived radionuclide impurities were experimentally assessed using high-resolution gamma and liquid scintillation spectrometries, while short-lived impurities were monitored analyzing the decay curve of the irradiated [18O]H2O with an activity calibrator. As spectrometric radionuclide library, a Geant4 Monte Carlo simulation of the 18F-target assembly was previously performed. Results 3H, 52,54Mn, 56,57,58Co, 95m,96Tc, 109Cd, and 184Re were found in the irradiated [18O]H2O, but no radionuclide was found in the non-irradiated [18O]H2O neither in the final [18F]FMCH solution with an activity concentration greater than the minimum detectable activity concentration. A total impurity activity 99.9999998% was estimated. Finally, the decay curve of the irradiated [18O]H2O revealed a very low maximum of 13N activity ( Conclusions This study demonstrated the radionuclidic purity of [18F]FMCH according to the EU Pharmacopeia.

The quantity of liquid organic radioactive wastes produced by the use of radioactive materials in nuclear research facilities is small compared to aqueous radioactive waste, but a special and low-cost treatment method is needed. Here we investigated the adsorption performance of five materials, namely: KU-2 resin, bentonite, charcoal (M&S) and clay adsorbents for the successful removal of 90Sr/90Y from liquid scintillation cocktail waste. The batch adsorption technique (influence of pH, contact time, and temperature), sequential, and column technique were investigated. The efficiency of these adsorbents for the removal of 90Sr/90Y is in this order, resin > bentonite > clay with removal efficiency 90 ± 5.2, 68 ± 3.25, and 65 ± 5.3%, respectively. While charcoal has lower affinity for the sorption processes. Purification of liquid scintillation (LS) cocktail by separation of 90Sr/90Y was successfully carried out by packed column with KU-2 resin. The exhausted loaded column with 90Sr/90Y is successfully regenerated by 25 mL, 1 M HNO3. Characterizations of the original and the purified LS cocktail were carried out using FTIR analysis. The efficiency of the purified liquid scintillation waste (LSW) for the determination of radionuclide is about 62.67 ± 4.8.

A method of measuring the tritium in seawater based on electrolytic enrichment and ultra-low background liquid scintillation counting techniques was established. The different factors influencing the detection limit were studied, including the counting time, the electrolytic volume of the seawater samples, the selection of background water, scintillation solution and their ratio. After optimizing the parameters and electrolyzing 350 mL volume of samples, the detection limit of the method was as low as 0.10 Bq/L. In order to test the optimization of system for this method, of the 84 seawater samples collected from the Arctic Ocean we measured, 92% were above the detection limit (the activity of this samples ranged from 0.10 Bq/L to 1.44 Bq/L with an average of (0.30±0.24) Bq/L). In future research, if we need to accurately measure the tritium activity in samples, the volume of the electrolytic samples will be increased to further reduce the minimum detectable activity.

Following a radiological or nuclear incident, the National Response Plan has given the Department of Health and Human Services / Centers for Disease Control and Prevention the responsibility for assessing population's contamination with radionuclides. In the public health response to the incident, valuable information could be obtained in a timely and accurate manner by using liquid scintillation counting techniques to determine who has been contaminated above background for alpha and beta emitting radionuclides. The calibration plays a major role in this process therefore, knowing the effect of quench agents on calibration is essential.

Aim: Human 14C radiotracer studies provide information-rich data sets that enable informed decision making in clinical drug development. These studies are supported by liquid scintillation counting after conventional-sized 14C doses (50–200 μCi) or complex accelerator mass spectrometry (AMS) after microtracer-sized doses (∼0.1–1 μCi). Mid-infrared laser-based ‘cavity ring-down spectroscopy’ (CRDS) is an emerging platform for the sensitive quantitation of 14C tracers. Results & methodology: We compared the total 14C concentrations in plasma and urine samples from a microtracer study using both CRDS and AMS technology. The data were evaluated using statistical and pharmacokinetic modeling. Conclusion: The CRDS method closely reproduced the AMS method for total 14C concentrations. With optimization of the automated sample interface and further testing, it promises to be an accessible, robust system for pivotal microtracer investigations

The triple-to-double coincidence ratio (TDCR) method is a liquid scintillation primary method for the absolute activity measurement of pure β− and pure electron capture emitters. This method requires specific three-photomultiplier liquid scintillation counters. The aim of the present work is to assess the TDCR method performance for routine tritium analysis in urine using an HIDEX 300 SL, the only three-photomultiplier liquid scintillation counter designed for routine laboratories. The physical parameters and the semi-empirical Birks parameter (kB) of the prepared liquid scintillation source were firstly determined. TDCR model equations solving and detection efficiencies calculations for measured samples were performed by TDCR07c computing program. Accuracy, uncertainties and detection limit of TDCR method were assessed through the tritium analysis of six intercomparison urine samples. The results demonstrate that the analytical performance of the TDCR method implemented on the HIDEX 300 SL is conform to the recommendations for the monitoring of workers exposed to tritium.

As part of the Centers for Disease Control and Prevention’s post-radiological/nuclear incident response mission, we developed rapid bioassay analytical methods to assess possible human exposure to radionuclides and internal contamination. Liquid scintillation counting (LSC) is a valuable analytical tool for the rapid detection and quantification of gross alpha/beta-emitting radionuclides in urine samples. A key characteristic of this type of bioassay method is its detection sensitivity for the priority threat radionuclides. We evaluated the limit of detection of selected LSC instruments to determine which instrument can be used when low-dose measurement is important.

Nuclear industry advancements and growing concerns about environmental contamination and terrorist activity have increased interest in quantifying radioisotopes in environmental and human samples. Increased presence in the environment, ease of entry into the food chain, nuclear medicine applications, and the possibility of radiological terrorism incidents can lead to human intake of these radionuclides (Radionuclides/Radiation Protection/US EPA (2020). https://www.epa.gov/radiation/radionuclides ; Radiation from the Earth (Terrestrial Radiation) (2015) Radiation and Your Health, Centers for Disease Control and Prevention. 7 December. https://www.cdc.gov/nceh/radiation/terrestrial.html ). A universal method to screen for and quantify individual radionuclides as well as both levels of alpha and beta emitters would address these concerns.

The paper presents the development of a rapid method for the direct determination of uranium using liquid scintillation analysis in uranium bearing materials with different uranium concentrations and with different impurity levels (U-ore, Uranium ore concentrate (UOC) and U-metal). Uranium extraction was carried out using hydrophobic ionic liquid Aliquat-336 thiosalicylate, ([A-336][TS]) from diluted samples in basic medium followed by liquid scintillation counting (LSC). Extraction efficiencies for uranium from aqueous medium was studied with very small volumes of ionic liquid under varying volume, pH and uranium concentration of the aqueous medium. Maximum extraction efficiency was achieved near pH 8–11. Uranium was successfully re-extracted from organic medium with nitric acid and electroplated onto stainless steel planchette for alpha spectrometry determination. Maximum re-extraction efficiency with 1 M HNO3 combined with electroplating efficiency observed was 80%. The methodology was applied for real samples. The uranium extraction efficiency using ionic liquid was about 85% for U-metal and UOC samples with comparatively low impurity level; whereas extraction efficiency for U-ore samples was found to be less than 40%. The technique can be applied for nuclear forensic applications as well as nuclear emergency scenarios for a quick initial assessment and isotopic analysis of uranium in the samples.

The elevated concentration of 210Po attached to aerosol particles inside a non-ferrous metal refinery was studied. The aerosols were collected on Petrianov or glass-fiber filters and subsequently subjected to radiochemical procedure consisting of acid leaching and concentration by micro co-precipitation with manganese oxide. For the activity determination alpha spectrometry and liquid scintillation counting (LSC) were applied. The obtained results showed that usually observed polonium concentration inside the plant was lower than 78 mBq m−3, however short spikes up to 7430 ± 710 mBq m−3 (related to the specific thermal process) were also observed.

The CO2 absorption and liquid scintillation counting (LSC) are methods used in radioactivity monitoring programs of nuclear facilities for 14C measurements due to high number of samples and relatively high expected level of 14C concentration. The paper describes the chemical sample preparation applied to a quality control material (IAEA-C3 Cellulose), in order to evaluate the reproducibility of CO2 absorption method for this type of material. Consequently, in the experiments we used two home-made scintillation cocktails, containing two amines, 2-methoxyethylamine (MEA) and 3-methoxypropyl amine (MPA), which detained CO2 as carbamates. Due to the fact that the material used in the dedicated experiments is cellulose (contained in all vegetable materials), the method can be considered as appropriate for 14C determination from biological and vegetable materials.

Rapid methods for determination of internal contamination play a key role during emergency situations in particular for incident response teams and affected populations. For this reason, these methods and their particular techniques must be fast, reliable, robust, simple and cost-efficient, as well as providing high-quality throughput. On the other hand, they are characterized by lower precision and they are often biased. The presented method is based on direct measurement of urine using powder scintillator YAP:Ce together with an alkaline medium as a replacement for traditional liquid scintillation cocktail. The overall efficiencies for trivalent actinides (Am-241 and Cm-244) were greater than 85% and 94%.

Lead-210 in drinking water can be rapidly determined by liquid scintillation counting (LSC) using a new sulfate precipitation method. In this method, 210Pb was first preconcentrated from water using iron hydroxide co-precipitation followed by sulfate precipitation to decontaminate most of non-alkaline earth elements. The Pb in the sulfate precipitate was then dissolved in strong alkaline solution, while alkaline earth elements (Sr, Ba, Ra) were separated as the sulfate/carbonate precipitate. To optimize the method, the influences of different acids and sulfate concentration on sulfate precipitation as well as the effects of pH, the added SO42−/CO32− concentrations and the Sr/Ba contents in the sulfate precipitate on the dissolution of PbSO4 were studied. In addition, X-ray fluorescence (XRF) analysis was applied for rapid determination of the chemical recovery and triple-to-double coincidence ratio (TDCR) quench correction curve for counting of 210Pb by LSC was established. The method was validated using 5 L of tap water samples and minimum detectable activity concentration (MDC) of ∼0.016 Bq·L−1 was achieved, which is sufficiently sensitive to meet the guidance level of 0.1 Bq·L−1 for 210Pb in drinking water as recommended by the World Health Organization (WHO).

For more accurate measurement of the potential alpha energy concentration (PAEC) of radon progeny by the liquid scintillation counting (LSC) method, its counting efficiency was studied in this work. A radon progeny standard source was used to provide different concentrations of radon progeny, and the progeny collected on the sampling filters were measured with a low background liquid scintillation counter. The result showed that the quenching effect could not be ignored, and the counting efficiency was estimated to be 86.12%. The co-comparison experiments showed that the PAECs of radon progeny measured by the LSC methods were in generally consistent with those of other three methods. It indicates that for more accurate measurements of PAEC of radon progeny by using the LSC method, the counting efficiency should be well determined.

A methodology for separation and purification of 210Po from uranium, thorium and daughters has been studied. Solvent extraction coupled with liquid scintillation analysis using HDEHP (di-(2-ethylhexyl) phosphoric acid) and TOPO (trioctylphosphine oxide) in toluene as extractive scintillator was optimized. About 95% of 210Po was extracted in the organic phase of TOPO as an extracting agent from an aqueous solution in 1 M HCl, compared to 1.6% extraction by HDEHP. The methodology was validated with uranium ore tailing sample, IAEA-385 sediment and IAEA-447 Moss-soil reference materials for application to environmental samples. A minimum detectable activity of 118 mBq kg−1 was achieved.

Thorium-227 is a potential therapeutic radionuclide for applications in targeted α-radioimmunotherapy for the treatment of various types of cancer. To provide nuclear medicine departments involved in Phase I clinical trials traceability to the SI unit of radioactivity (Bq), a standardisation of a radiochemically pure 227Th aqueous solution has been performed at the National Physical Laboratory. This was achieved via two primary liquid scintillation (LS) techniques −4π(LS)-γ digital coincidence counting (DCC) and 4π LS counting. These absolute techniques were supported by the indirect determination of the 227Th activity via the measurement of the ingrowth and decay rate of the decay progeny by both ionisations chambers and high purity germanium (HPGe) gamma-ray spectrometry. The results of the primary techniques were found to be consistent, both with each other (zeta score = 1.1) and to the decay progeny ingrowth measurements. An activity per unit mass of 20.726 (51) kBq g−1 was determined for the solution. A procedure has been developed that provided an effective separation of the 227Th from its decay progeny, which was shown by the effective time zero of the 227Th-223Ra nuclear chronometer measured by HPGe gamma-ray spectrometry.

This study was designed to optimize calcium-45 (45Ca) sample preparation method in a Liquid Scintillation Counter (LSC) for use with vegetables and soil samples. Different tests were conducted on f...

The liquid scintillation counting procedure for tritium quantification in environmental samples was optimized with respect to sample to cocktail volume ratio, fixing of region of interest of tritium spectra, and choice of counting vials for measurement of low-level activities of tritium in water samples. The optimum sample to cocktail ratio for Aqua-light scintillation cocktail based on the MDA values was found to be 8: 12 mL. The optimum window for tritium at optimum sample to cocktail ratio was found to be 2–12 keV based on figure of merit values, and the reduction in the background count rate was 31% compared to full window. The efficiency of the system at optimum sample to cocktail ratio was studied using two different quenching agents, carbon tetrachloride (CCl4) and nitromethane at different quench levels. It was found that quench effects were stronger for the standards with nitromethane compared to CCl4 at any specific volume. The tritium recovery factor for distillation technique, which is used for removing quenching agents present in water samples, was found to be 95% with the relative standard deviation of 0.34%. The conditions established in this paper are useful for measurement of low-level activities of tritium in the environmental samples.

The present work examines the validity of using barium as a carrier for radium-226 determination and optimizes the amount added during the radiochemical separation from rock samples. The activity concentration of Ra-226 is determined in rock samples by liquid scintillation spectrometer via its short-lived daughters after the radiochemical separation. The activity of Ra-226 was calculated based on the counts measured in a window set for Po-214 peak. The activity measured by Liquid Scintillation Counter, LSC using Pulse Shape Analysis, PSA. Ba-133 was used for chemical yield. The applied method was tested using a Certified Reference Material, CRM ore number-Dl-1a which shows a good agreement with the certified values. The applied method for measuring Ra-226 has been adapted to be used in rocks such as monazite. The average activity for Ra-226 in rock samples collected from the commercial grade Egyptian monazite was 13.85 ± 2.05 Bq/g with a detection limit 9.58 mBq/g. The optimum conditions have been established during the present work such as precision, cost and time preparation which makes the method useful for radio chemists and the decision makers.

89Sr and 90Sr are both fission products of high radiotoxicity, which can be released in significant amounts in the event of a nuclear accident. Radiostrontium isotopes will follow calcium all along the food chain and, after ingestion, accumulate in the bones. Therefore, it is imperative to be able to determine 89Sr and 90Sr in raw milk samples in case of an accidental situation to evaluate the dose given by both radionuclides to the population. Several methods exist for conducting 89Sr and 90Sr determination. However, most of them use at least one chromatographic step to purify strontium. This, unfortunately, increases the analytical time before the results can be released to the authorities. In addition, they often use liquid scintillation counting to determine the 89Sr and 90Sr activities, a method which can handle only one sample at a time. Here we propose using synthetic tunnel manganese oxides such as cryptomelane and todorokite and layered metal sulfides to selectively extract strontium from fresh milk and raw urine in a batch sorption method. We found that the method is very quick and yields very pure sources of (radio)-strontium, which can be counted in a proportional counter. Data (counts per minute) from the counter were fitted to a mathematical expression enabling the simultaneous determination of 89Sr and 90Sr. Because a proportional counter often has several drawers, it is typically possible to measure up to 16 samples at a time. Since cryptomelane is a binding phase easily synthesized in a large quantity, we anticipate that this technique could be an interesting alternative to conventional solid phase extraction chromatography methods.

Rapid determination of selected gross alpha and beta emitters in environmental matrices by solid-state scintillation technique is discussed. This method is based on sample treatment using microwave reactor and direct measurement of digested products using powder scintillator and alkaline solution as a substitute for traditional liquid scintillation cocktail. The selected group of radionuclides was chosen with respect to their use in nuclear industry, high radiotoxicity, and the possibility of potential misuse. The work aimed at verifying the connection of microwave decomposition using alkaline solution with solid-state scintillation using a powder scintillator YAP:Ce together with an alkaline medium.

In this paper, experiences of the last 20 years with the PERALS-technique are described. PERALS stands for photo electron-rejecting alpha liquid scintillation. This liquid scintillation technique was developed by Jack McDowell in the 1970s and is a powerful technique for the analyses of many natural alpha nuclides and also the beta nuclide 90Sr. The principle is based on a selective extraction of the radionuclide from the water phase by means of a complexing or ion pair reagent. The extractant contains also a cocktail suitable for scintillation counting. Therefore, the extract can be analyzed directly after the extraction step. After removing quenchers, such as oxygen, and the proper setting of a pulse shape discriminator, alpha pulses can be counted with a photomultiplier. This paper describes the development of robust analysis schemes for the determination of traces of polonium, thorium, uranium and other actinides in water samples (groundwater, rain water, river water, drinking water, mineral water, sea water). For radon and radium, the enrichment in the extract is poor. Therefore, PERALS methods are not suitable for trace analyses of these analytes. In addition, the extraction of the beta-emitter 90Sr with a PERALS cocktail is discussed, even though its beta spectrum is not analyzed with a PERALS counter. Results from the survey of drinking water and mineral water in Switzerland are presented for every radio element.

Comparison of two approaches for measurement of 93mNb was made to serve as a base for evaluation of new radiochemical procedures for precise sample preparation. Newly available materials for radiochemical separation of important nuclides will be used in these procedures. The 93mNb samples were irradiated using continuous energy neutron flux. The specific activities are measured using liquid scintillation counting and X-ray spectrometry. The results agree within the measurement error and may serve as base for evaluation of further activities in applying new materials for radiochemical separation of natural and technogenic radionuclides important for the environment.

Reliable determination of 226Ra content in drinking water, surface water and groundwater is required for radiological health-risk assessment of populations and radiation-dose calculations after ingestion and inhalation. This study aimed to determine 226Ra presence in the untreated water samples on a liquid scintillation counter via Cherenkov radiation detection. Cherenkov counting is a faster, simpler, less expensive technique than other commonly used methods for 226Ra determination. Step-by-step optimization of this technique on the Quantulus detector is presented in this paper. Improvement of detection limit/efficiency in the presence of sodium salicylate was investigated in this study. The main parameters of the method obtained were detection efficiency 15.87 (24)% and detection limit 0.415 Bq/L achieved for 1000 min of counting in 20 mL of sample volume. When 1 g of sodium salicylate was added, efficiency increased to 38.1 (5)%, with a reduction in the detection limit to 0.248 Bq/L for 500 min of counting. A satisfactory precision level of Cherenkov counting was obtained, the results deviating between 5% and 20% from reference values. The precision and accuracy of the Cherenkov counting technique were compared to liquid scintillation counting (EPA Method 913.0 for radon determination) and gamma spectrometry (the direct method for the untreated water samples on HPGe spectrometer). An overview of the advantages/disadvantages of each technique is elaborated in this paper.

W samples, previously irradiated with 20 MeV W ions at room temperature to a displacement-damage level of 0.23 displacements per atom (dpa) at the peak of displacements, were exposed to a deuterium-tritium (D2 ― DT ― T2) gas mixture with a tritium content of 6% at a total pressure of 1.2 kPa and temperatures of 773 K and 973 K for 3 h. The concentration of tritium retained in the displacement-damage zone of these W samples was determined by a method combining chemical etching and subsequent analysis of tritium in the etching solution using liquid scintillation counting (CE-LSC). (Some figures may appear in colour only in the online journal)

Effluents containing tritium (H) dispersed into the fresh water or marine environment from nuclear facilities can be taken up by biota. Aquatic and marine organisms are among the important pathways through which tritium can enter into the human body, and hence, assessment of the extent of pollution of these ecosystems is very important for radiation dose assessments. Tritium present in environmental matrices can be classified as tissue-free water tritium (TFWT) and organically bound tritium (OBT). Optimization of a method for the determination of OBT in fish, based on thermal oxidation of the sample, is discussed. Samples were subjected to thermal oxidation in a pyrolyser system, and the water produced from the combustion was analyzed by liquid scintillation spectrometry. Results show that a maximum of ~2 g of processed fish sample can be combusted efficiently in the pyrolyser. Using this method, a recovery of 84% was achieved, and minimum detectable activity (MDA) for the method was determined to be 8.5 Bq kg (sample weight = 2 g, counting time = 30,000 s, and detection efficiency = 20%).

In this paper, a method based on ultra-low level liquid scintillation analyzer for measuring uranium radioactivity in low-level radioactive wastewater is proposed. It can easily and quickly measure the radioactivity level of uranium in samples, and even distinguish the main isotopes of uranium. In low-level wastewater samples, if the uranium isotope has the same or similar composition and ratio to the natural uranium isotope, Comparing to the results in units of mass concentration given by laser-fluoremetry method, the liquid scintillation method can give the results in units of activity concentration, and these two results with different units can interconvert to each other. However, interconversion of results on above two measurement methods cannot be applied in low-level wastewater after uranium enrichment, because the proportion of isotopes with high specific activity in above wastewater, such as 234U and 235U, is significantly higher than that without enrichment. The measurement results in units of mass concentration will underestimate the uranium radioactivity level of the sample. The liquid scintillation method directly gives the results in units of radioactivity activity concentration, and it is more convenient to compare with relevant national standards to determine whether the emission standards are met. The lowest limit of detection of this method is 0.014 Bq L−1 within 600 min.

Detection and measurement of low-energy beta particles is commonly achieved by liquid scintillation counting, in particular for low-level tritium samples. When samples are contained in plastic scintillation vials for long-term storage, the tritium activity in the vials has been found to decrease faster than expected from its natural radioactive decay. Different explanations for this observation have attributed some of these tritium activity losses to diffusion of the sample, degradation of the LSC cocktail, and the potential long-term changes in quenching effects of the LSC cocktail. An alternative explanation may also be that the tritium organically binds to the carbon chains in the plastic bottle through direct H and H atom exchange. A study was designed and performed to test this latter hypothesis of H and H atom exchange in plastic. Deionized water was introduced in a plastic vial that previously contained tritiated water to assess any increase in tritium activity from the reverse atom exchange between the vial material and the deionized water. A greater loss in activity concentration is observed in plastic vials compared to glass vials as a function of storage time for the tritiated water. Furthermore, the tritium activity concentration in the deionized water increased when storage occurred in plastic vials, an effect that is not observed for storage in glass vials. The study results indicate that hydrogen atom exchange may possibly take place in plastic vials.