Exposure of German hunters and their family members to the radioactive nuclide 137Cs due to their eating habits.

The radioactive nuclide 137Cs that was released during the Chernobyl accident in 1986 is still present in the environment of wide parts of Middle Europe. It causes internal radiation exposure of people who incorporate this radionuclide when eating contaminated food, in particular boar meat and certain kinds of wild mushrooms. This study investigates the exposure of German hunters and their family members, which are supposed to consume these kinds of food more frequently than the general population. For this purpose, the body activity of 137Cs was measured in 58 hunters and their family members mostly from Southern Germany and it was compared with body activities from 1729 measurements of 251 members of the general population. Significantly higher 137Cs body activities were found in the hunters and their family members compared to the references after matching both groups for sociodemographic factors and the time of the measurement (factor of 2.4 between the geometric means, p-value ≪ 0.001). Within the hunters and their family members, an adjusted exponential regression showed significant associations of the activity and of mushroom picking (factor 2.6, 95% confidence interval 1.5-4.6) and of the frequency of having boar meat as part of the diet (factor 1.4 per meal per week, 95% confidence interval 1.1-1.8). No significant association was found for the 137Cs surface contamination at the hunting grounds. Within families, mostly similar results were found in terms of activity concentrations. For the hunter with the highest measured activity (1050 Bq), an annual contribution to the effective dose of 35 µSv was calculated, showing that also in hunters and their family members 137Cs is expected to cause only small contributions to the total effective dose from environmental exposure.

Thyroid monitoring of adults and children after reactor accidents with a new dose rate measurement device.

A specialized dose rate measurement device that was designed for monitoring the thyroid dose of children and adults after reactor accidents was tested. In measurements with neck phantoms and a human patient, the device was found to be capable of measuring the required low dose rates, even within increased ambient radiation. It is suitable for the application in emergency care centers.

A puzzling case of contamination with 241Am.

Two people were exposed to and contaminated with 241Am. In vivo determinations of the incorporated 241Am were performed using a whole-body counter and two partial-body counters for the skull and lung, respectively. Additionally, urine samples were analysed to estimate the systemic activity removed from the body. To improve the geometry of the skull measurements, an optimised detector configuration was used, a calibration with three physical phantoms of the human head was conducted, and the morphological variability between the individuals was also considered. The results of the measurements indicate that activity is not deposited in the deep tissues, rather in the skin tissues close to the body surface. Unfortunately, the many open questions relating to the actual circumstances during and after the incident make the interpretation of this case difficult if at all possible.

GHSI EMERGENCY RADIONUCLIDE BIOASSAY LABORATORY NETWORK: SUMMARY OF A RECENT EXERCISE.

The Global Health Security Initiative (GHSI) established a laboratory network within the GHSI community to develop their collective surge capacity for radionuclide bioassay in response to a radiological or nuclear emergency. A recent exercise was conducted to test the participating laboratories for their capabilities in screening and in vitro assay of biological samples, performing internal dose assessment and providing advice on medical intervention, if necessary, using a urine sample spiked with a single radionuclide, 241Am. The laboratories were required to submit their reports according to the exercise schedule and using pre-formatted templates. Generally, the participating laboratories were found to be capable with respect to rapidly screening samples for radionuclide contamination, measuring the radionuclide in the samples, assessing the intake and radiation dose, and providing advice on medical intervention. However, gaps in bioassay measurement and dose assessment have been identified. The network may take steps to ensure that procedures and practices within this network be harmonised and a follow-up exercise be organised on a larger scale, with potential participation of laboratories from the networks coordinated by the International Atomic Energy Agency and the World Health Organization.

AN INTERLABORATORY COMPARISON ON THE DETERMINATION OF 241Am, 244Cm AND 252Cf IN URINE.

An intercomparison exercise on the determination of (241)Am, (244)Cm and (252)Cf in urine was performed. Since it was designed with regard to emergency preparedness, the detection limit for each nuclide was set to 0.1 Bq per 24-h urine sample. Most of the participating laboratories were established bioassay laboratories. However, some laboratories that routinely determine (241)Am only in environmental samples were also invited in order to explore their potential for emergency bioassay analysis. Another aspect of the intercomparison was to investigate the performance of all laboratories concerning the chemical yields of the (243)Am tracer in comparison with (244)Cm and (252)Cf. In summary, both types of laboratories showed good results. There was a negative bias for the results of (244)Cm and (252)Cf, which can be explained by slightly different radiochemical behaviours of americium, curium and californium and which is in agreement with results reported in the literature.

Preparation of Ca-alginate biopolymer beads and investigation of their decorporation characteristics for 85Sr, 238U and 234Th by in vitro experiments.

The aim of this work was to investigate whether Ca-alginate biopolymer beads (CaABBs) can be used to reduce the bioavailability of radionuclides in the gastrointestinal tract of humans. The uptake of strontium, uranium and thorium from a simulated gastrointestinal system was studied by in vitro techniques using CaABBs. This agent was prepared from Na-alginate through cross-linking with divalent calcium ions according to the egg-box model. The effects of process variables such as pH of the gastrointestinal juice, incubation time and solid-to-solution ratio for the removal of radionuclides from the gastrointestinal juice were investigated. The results suggest that CaABBs are a potent material for reducing the bioavailability of radionuclides with a high uptake efficiency in the gastrointestinal tract.

Solubility of uranium and thorium from a healing earth in synthetic gut fluids: a case study for use in dose assessments.

The aim of this case study was to estimate the bioaccessibility of uranium ((238)U) and thorium ((232)Th) from a healing earth by analysing the solubility of these radionuclides in synthetic gastric and intestinal fluids. An easy applicable in vitro test system was used to investigate the fractional mobilization of the soil contaminants being potentially available for absorption under human in vivo conditions. These findings provided the basis for a prospective dose assessment. The solubility experiments were performed using two different in vitro digestion methods. The concentrations of (238)U and (232)Th in the solutions extracted from the soil were measured by inductively coupled plasma mass spectrometry (ICP-MS). The dissolved fractions in the synthetic gastrointestinal fluid ranged in average from 10.3% to 13.8% for (238)U and from 0.3% to 1.6% for (232)Th, respectively, depending on the digestion method. Subsequently, the committed effective doses from intake of (238)U and (232)Th after ingestion of the healing earth during 1 year were evaluated for adult persons. Thereby ingestion dose coefficients calculated as a function of bioaccessibility were used. The dose assessments ranged between 4.3 × 10(-7)-1.9 × 10(-6) Sv y(-1) for (238)U and 5.6 × 10(-7)-3.3 × 10(-6) Sv y(-1) for (232)Th, respectively. On the basis of the assumptions and estimations made, the present work indicates a relatively low radiation risk due to (238)U and (232)Th after internal exposure of the healing earth.

Genome-wide analysis of caesium and strontium accumulation in Saccharomyces cerevisiae.

(137)Cs and (90)Sr contribute to significant and long-lasting contamination of the environment with radionuclides. Due to their relatively high biological availability, they are transferred rapidly into biotic systems and may enter the food chain. In this study, we analysed 4862 haploid yeast knockout strains of Saccharomyces cerevisiae to identify genes involved in caesium (Cs(+)) and/or strontium (Sr(2+)) accumulation. According to this analysis, 212 mutant strains were associated with reproducible altered Cs(+) and/or Sr(2+) accumulation. These mutants were deficient for a wide range of cellular processes. Among those, the vacuolar function and biogenesis turned out to be crucial for both Cs(+) and Sr(2+) accumulation. Disruption of the vacuole diminished Cs(+) accumulation, whereas Sr(2+) enrichment was enhanced. Further analysis with a subset of the identified candidates were undertaken comparing the accumulation of Cs(+) and Sr(2+) with their essential counterparts potassium (K(+)) and calcium (Ca(2+)). Sr(2+) and Ca(2+) accumulation was highly correlated in yeast excluding the possibility of a differential regulation or uptake mechanisms. In direct contrast, the respective results suggest that Cs(+) uptake is at least partially dependent on mechanisms distinct from K(+) uptake. Single candidates (e.g. KHA1) are presented which might be specifically responsible for Cs(+) homeostasis.

Radiation dose assessment of exposure to depleted uranium.

Depleted uranium (DU) is claimed to contribute to human health problems, known as the Gulf War Syndrome and the Balkan Syndrome. Quantitative radiation dose is required to estimate the health risk of DU materials. The influences of the solubility parameters in the human alimentary tract and the respiratory tract systems and the aerosol particles size on the radiation dose of DU materials were evaluated. The dose conversion factor of daily urinary excretion of DU is provided. The retention and excretion of DU in the human body after a contamination at a wound site were predicted. Dose coefficients of DU after ingestion and inhalation were calculated using the solubility parameters of the DU corrosion products in simulated gastric and simulated lung fluid, which were determined in the Helmholtz Zentrum München. (238)U is the main radiation dose contributor per 1 Bq of DU materials. The dose coefficients of DU materials were estimated to be 3.5 x 10(-8) and 2.1 x 10(-6) Sv Bq(-1) after ingestion and inhalation for members of the public. The ingestion dose coefficient of DU materials is about 75% of the natural uranium value. The inhalation dose coefficient of DU material is in between those for Type M and Type S according to the category for inhaled materials defined by the International Commission on Radiological Protection. Radiation dose possibly received from DU materials can directly be estimated by using the dose conversion factor provided in this study, if daily urinary excretion of DU is measured.

Investigations on the solubility of corrosion products on depleted uranium projectiles by simulated body fluids and the consequences on dose assessment.

Ingestion and inhalation of corrosion products covering weathered penetrators made of depleted uranium (DU) represent potential radiological exposure pathways. In order to study the bioavailability of these corrosion products, their solubility was determined using simulated gastric and pulmonary juices. About 75 and 36% of the uranium in the corrosion products were found to be soluble in simulated gastric and pulmonary juices, respectively. The effective dose coefficient for adults after ingestion was calculated to be 0.61 muSv mg(-1) DU. This compares to an effective dose coefficient for an adult of 0.71 muSv mg(-1) for DU materials given by the World Health Organization (WHO). The effective dose coefficient for inhalation was calculated to be 3.7 x 10(-6 )Sv Bq(-1) for workers and 5.3 x 10(-6 )Sv Bq(-1) for members of the public, respectively, which is between those of particles of Types M and S as defined by the International Commission on Radiological Protection (ICRP). The speciation of the corrosion products was investigated by time-of-flight secondary ion mass spectrometry (TOF-SIMS). The mean oxidation state of uranium was found to be 4.6, which suggests that the uranium in the corrosion products consists of a mixture of U(IV) and U(VI) species.

Internal dose assessment of 210Po using biokinetic modeling and urinary excretion measurement.

The mysterious death of Mr. Alexander Litvinenko who was most possibly poisoned by Polonium-210 ((210)Po) in November 2006 in London attracted the attention of the public to the kinetics, dosimetry and the risk of this high radiotoxic isotope in the human body. In the present paper, the urinary excretion of seven persons who were possibly exposed to traces of (210)Po was monitored. The values measured in the GSF Radioanalytical Laboratory are in the range of natural background concentration. To assess the effective dose received by those persons, the time-dependence of the organ equivalent dose and the effective dose after acute ingestion and inhalation of (210)Po were calculated using the biokinetic model for polonium (Po) recommended by the International Commission on Radiological Protection (ICRP) and the one recently published by Leggett and Eckerman (L&E). The daily urinary excretion to effective dose conversion factors for ingestion and inhalation were evaluated based on the ICRP and L&E models for members of the public. The ingestion (inhalation) effective dose per unit intake integrated over one day is 1.7 x 10(-8) (1.4 x 10(-7)) Sv Bq(-1), 2.0 x 10(-7) (9.6 x 10(-7)) Sv Bq(-1) over 10 days, 5.2 x 10(-7) (2.0 x 10(-6)) Sv Bq(-1) over 30 days and 1.0 x 10(-6) (3.0 x 10(-6)) Sv Bq(-1) over 100 days. The daily urinary excretions after acute ingestion (inhalation) of 1 Bq of (210)Po are 1.1 x 10(-3) (1.0 x 10(-4)) on day 1, 2.0 x 10(-3) (1.9 x 10(-4)) on day 10, 1.3 x 10(-3) (1.7 x 10(-4)) on day 30 and 3.6 x 10(-4) (8.3 x 10(-5)) Bq d(-1) on day 100, respectively. The resulting committed effective doses range from 2.1 x 10(-3) to 1.7 x 10(-2) mSv by an assumption of ingestion and from 5.5 x 10(-2) to 4.5 x 10(-1) mSv by inhalation. For the case of Mr. Litvinenko, the mean organ absorbed dose as a function of time was calculated using both the above stated models. The red bone marrow, the kidneys and the liver were considered as the critical organs. Assuming a value of lethal absorbed dose of 5 Gy to the bone marrow, 6 Gy to the kidneys and 8 Gy to the liver, the amount of (210)Po which Mr. Litvinenko might have ingested is therefore estimated to range from 27 to 1,408 MBq, i.e 0.2-8.5 microg, depending on the modality of intake and on different assumptions about blood absorption.