Studies of strontium biokinetics in humans. Part 2: uptake of strontium from aqueous solutions and labelled foodstuffs

The accidental release of radioactive isotopes of strontium, mainly (90)Sr, into the environment and its transfer into the biosphere results in an internal radiation exposure of the affected population. In order to obtain reliable estimates of the committed dose due to an intake of Sr radionuclides, reliable information on its metabolic behaviour inside the human body is needed, i.e. biokinetic data on fractional uptake from contaminated foodstuffs, distribution to and retention in different organs and tissues of the human body. Such information can be obtained by tracer kinetic investigations. The committed effective dose is dependent on the fractional intestinal absorption of the ingested activity (f(1) value). The International Commission on Radiological Protection in its publication ICRP 67 adopted an f(1) value of 0.3 for adults. This study is aimed at investigating if the value corresponds with the actual uptake from contaminated foodstuffs. Aqueous solutions and contaminated vegetables, i.e. cress and salad (lettuce) were used as test materials. For this purpose, the methodology for intrinsic labelling of foodstuffs described in part 1 was applied. For aqueous solutions, a mean f(1) value of 0.63+/-0.14 (mean +/- SD) was obtained by administering 1 mg of strontium. The uptake of Sr from cress intrinsically labelled with about 1 mg Sr almost corresponds to that from aqueous solutions (f(1)=0.62+/-0.10), but from lettuce it is reduced by a factor of 2 (f(1)=0.27+/-0.08).     

Studies of strontium biokinetics in humans

The radioactive isotopes of strontium, mainly (90)Sr, which are common fission products, may significantly contribute to the internal exposure of the population in case of their accidental release into the environment and transfer to the food chain. For (90)Sr, the internal radiation dose is significantly dependent on the fractional absorption of the ingested activity (f(1)-value). Human data on the absorption of dietary strontium and of soluble forms of the element give values ranging from about 0.15 to 0.45 (up to 1.0) for adults. The International Commission on Radiological Protection (ICRP) has adopted f(1)-values of 0.6 for children of less than 1 year of age, 0.4 for children between 1 and 15 years and 0.3 for adolescents above 15 years of age. This study was aimed at investigating how far these values correspond to the actual uptake of strontium from contaminated foodstuffs. A methodology is presented that has been developed for preparing foodstuffs intrinsically labelled with stable isotopes and that will be used in tracer kinetic investigations. The results show that cress and salad can be adequately labelled, i.e. a strontium concentration of 1.36+/-0.47 g per kg of cress (wet weight) and of 0.29+/-0.04 g per kg of salad (wet weight) may be obtained within 15 days and 24 days, respectively. For the biokinetic investigations on humans, applying stable isotopes of Sr as tracers, about 0.1-1 mg strontium is required per volunteer, i.e. a few grams of the edible parts of the labelled material are sufficient.     

Influence of human biokinetics of strontium on internal ingestion dose of <Superscript>90</Superscript>Sr and absorbed dose of <Superscript>89</Superscript>Sr to organs and metastases

The objective of the present work is to apply the plasma clearance parameters to strontium, previously determined in our laboratory, to improve the biokinetic and dosimetric models of strontium-90 (⁹⁰Sr) used in radiological protection; and also to apply this data for the estimation of the radiation doses from strontium-89 (⁸⁹Sr) after administration to patients for the treatment of the painful bone metastases. Plasma clearance and urinary excretion of stable strontium tracers of strontium-84 (⁸⁴Sr) and strontium-86 (⁸⁶Sr) were measured in GSF-National Research Center for Environment and Health (GSF) in 13 healthy German adult subjects after intravenous injection and oral administration. The biological half-life of strontium in plasma was evaluated from 49 plasma concentration data sets following intravenous injections. This value was used to determine the transfer rates from plasma to other organs and tissues. At the same time, the long-term retention of strontium in soft tissue and whole body was constrained to be consistent with measured values available. A physiological urinary path was integrated into the biokinetic model of strontium. Parameters were estimated using our own measured urinary excretion values. Retention and excretion of strontium were modeled using compartmental transfer rates published by the International Commission on Radiological Protection (ICRP), the SENES Oak Ridge Inc. (SENES), and the Urals Research Center for Radiation Medicine (TBM). The results were compared with values calculated by applying our GSF parameters (GSF). For the dose estimation of ⁸⁹Sr, a bone metastases model (GSF-M) was developed by adding a compartment, representing the metastases, into the strontium biokinetic model. The related parameters were evaluated based on measured data available in the literature. A set of biokinetic parameters was optimized to represent not only the early plasma kinetics of strontium but also the long-term retention measured in soft tissue and whole body. The ingestion dose coefficients of ⁹⁰Sr were computed and compared with different biokinetic model parameters. The ingestion dose coefficients were calculated as 2.8 × 10⁻⁸, 2.1 × 10⁻⁸, 2.5 × 10⁻⁸ and 3.8 × 10⁻⁸ Sv Bq⁻¹ for ICRP, SENES, TBM and GSF model parameters, respectively. Moreover, organ absorbed dose for the radiopharmaceutical of ⁸⁹Sr in bone metastases therapy was estimated based on the GSF and ICRP biokinetic model parameters. The effective doses were 3.3, 1.8 and 1.2 mSv MBq⁻¹ by GSF, GSF-M, and ICRP Publication 67 model parameters, respectively, compared to the value of 3.1 mSv MBq⁻¹ reported by ICRP Publication 80. The absorbed doses of red bone marrow and bone surface, 17 and 21 mGy MBq⁻¹ calculated by GSF parameters, and 7.1 and 8.8 mGy MBq⁻¹ by GSF-M parameters, are comparable to the clinical results of 3–19 mGy MBq⁻¹ for bone marrow and 16 mGy MBq⁻¹ for bone surface. Based on the GSF-M model, the absorbed dose of ⁸⁹Sr to metastases was estimated to be 434 mGy MBq⁻¹. The strontium clearance half-life of 0.25 h from the plasma obtained in the present study is obviously faster than the value of 1.1 h recommended by ICRP. There are no significant changes for ingestion dose coefficients of ⁹⁰Sr using different model parameters. A model including the metastases was particularly developed for dose estimation of ⁸⁹Sr treatment for the pain of bone metastases.     

The application and adaptation of ICRP internal dosimetry models to the calculation of bone marrow tissue doses from 90Sr for epidemiological studies of Techa River populations

The operation of the Mayak Production Association in the Southern Urals region of Russia, resulted in releases of large amounts of radioactive effluent into the Techa River during the period 1949-1956. The residents of the riverside communities were thus exposed to both external radiation, and internal radiation following ingestion of contaminated water and foodstuffs. One of the most important radionuclides for internal exposure was 90Sr. This paper gives a brief overview of the models provided by International Commission on Radiological Protection (ICRP), which are of interest for assessing internal doses from 90Sr. The application of these models to the calculation of red bone marrow doses for the fetus and infant from 90Sr intakes by the mother and the infant is illustrated by an example. A hypothetical individual born in 1951 is used as an example for dose calculations. The following doses due to intakes of 90Sr are taken into account: received in utero due to maternal intakes during pregnancy; received after birth from 90Sr accumulated by the fetus in utero; from intakes in breast milk; from intakes in the infant's diet after weaning. It is shown that doses to the fetus following maternal ingestion and subsequent transfer to the fetus via the placenta dominate the doses received for this particular individual for the first two years of life. Doses to the infant from intakes in breast milk are substantially lower but do make significant contributions to total doses in the first two years after birth. By about the age of two years residual 90Sr from placental transfer still contributes about the same dose as do intakes by the infant, but in later years doses from intakes by the infant dominate.     

Internal dose assessment of <Superscript>210</Superscript>Po using biokinetic modeling and urinary excretion measurement

The mysterious death of Mr. Alexander Litvinenko who was most possibly poisoned by Polonium-210 (²¹⁰Po) 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 ²¹⁰Po 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 ²¹⁰Po 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 × 10⁻⁸ (1.4 × 10⁻⁷) Sv Bq⁻¹, 2.0 × 10⁻⁷ (9.6 × 10⁻⁷) Sv Bq⁻¹ over 10 days, 5.2 × 10⁻⁷ (2.0 × 10⁻⁶) Sv Bq⁻¹ over 30 days and 1.0 × 10⁻⁶ (3.0 × 10⁻⁶) Sv Bq⁻¹ over 100 days. The daily urinary excretions after acute ingestion (inhalation) of 1 Bq of ²¹⁰Po are 1.1 × 10⁻³ (1.0 × 10⁻⁴) on day 1, 2.0 × 10⁻³ (1.9 × 10⁻⁴) on day 10, 1.3 × 10⁻³ (1.7 × 10⁻⁴) on day 30 and 3.6 × 10⁻⁴ (8.3 × 10⁻⁵) Bq d⁻¹ on day 100, respectively. The resulting committed effective doses range from 2.1 × 10⁻³ to 1.7 × 10⁻² mSv by an assumption of ingestion and from 5.5 × 10⁻² to 4.5 × 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 ²¹⁰Po which Mr. Litvinenko might have ingested is therefore estimated to range from 27 to 1,408 MBq, i.e 0.2–8.5 μg, depending on the modality of intake and on different assumptions about blood absorption.     

Somatic cell mutations at the glycophorin a locus in erythrocytes of radiation workers from the Sellafield nuclear facility

The glycophorin A (GPA) somatic mutation assay for N0 and NN mutant erythrocytes was performed on 245 current and 48 retired workers who had been occupationally exposed to radiation at the British Nuclear Fuels plc facility at Sellafield. A positive association with increasing age was found for current workers for both N0 and NN frequencies of 0.14 +/- 0.05 x 10(-6) (P = 0.012) and 0.25 +/- 0.07 x 10(-6) (P = 0.0003) per year, respectively. No association with age was found for the retired workers. In a comparison of ever-smokers with never-smokers, no difference was observed for N0 frequencies for current workers, but a significantly higher frequency was found for ever-smokers in the retired group (P = 0.001). NN mutant frequencies were slightly higher in ever-smokers than in never-smokers for both current and retired workers, but in neither case was the increase significant. In age-adjusted analyses for N0 mutant frequencies, a slight positive radiation dose response was found for current workers (1.6 +/- 3.8 x 10(-6) per Sv), for retired workers (2.9 +/- 2.5 x 10(-6) per Sv), and in the combined analysis (2.6 +/- 2.2 x 10(-6) per Sv), but in no case did this reach significance. Similar analyses for NN mutant frequencies revealed a positive dose response for current workers (4.7 +/- 4.6 x 10(-6) per Sv) and a negative response for retired workers (-2.4 +/- 3.6 x 10(-6) per Sv) that was maintained in the combined analysis (-1.4 +/- 2.8 x 10(-6) per Sv), but none of these slopes was significantly different from zero. The results suggest that the GPA mutation assay is insufficiently sensitive to be used as a biological marker of low-dose chronic exposure and provide further evidence that, in contrast to high acute radiation exposure, protracted exposure is much less effective at inducing somatic mutations in vivo.     

Reconstruction of radionuclide intakes for the residents of East Urals Radioactive Trace (1957–2011)

The East Urals Radioactive Trace (EURT) was formed after a chemical explosion in the radioactive waste-storage facility of the Mayak Production Association in 1957 (Southern Urals, Russia) and resulted in an activity dispersion of 7.4?×?10¹⁶ Bq into the atmosphere. Internal exposure due to ingestion of radionuclides with local foodstuffs was the main factor of public exposure at the EURT. The EURT cohort, combining residents of most contaminated settlements, was formed for epidemiological study at the Urals Research Center for Radiation Medicine, Russia (URCRM). For the purpose of improvement of radionuclide intake estimates for cohort members, the following data sets collected in URCRM were used: (1) Total β-activity and radiochemical measurements of ⁹⁰Sr in local foodstuffs over all of the period of interest (1958–2011; n?=?2200), which were used for relative ⁹⁰Sr intake estimations. (2) ⁹⁰Sr measurements in human bones and whole body (n?=?338); these data were used for average ⁹⁰Sr intake derivations using an age- and gender-dependent Sr-biokinetic model. Non-strontium radionuclide intakes were evaluated on the basis of ⁹⁰Sr intake data and the radionuclide composition of contaminated foodstuffs. Validation of radionuclide intakes during the first years after the accident was first carried out using measurements of the feces β-activity of EURT residents (n?=?148). The comparison of experimental and reconstructed values of feces β-activity shows good agreement. ⁹⁰Sr intakes for residents of settlements evacuated 7–14 days after the accident were also obtained from ⁹⁰Sr measurements in human bone and whole body. The results of radionuclide intake reconstruction will be used to estimate the internal doses for the members of the EURT cohort.     

Effects of low chronic doses of ionizing radiation on antioxidant enzymes and G6PDH activities in Stipa capillata (Poaceae)

Stipa capillata (Poaceae) seeds were harvested from a control area (displaying a gamma dose rate of 0.23 micro Sv h(-1)) (C plants) and from two contaminated areas (5.4 and 25 micro Sv h(-1)) on the Semipalatinsk nuclear test site (SNTS) in Kazakhstan. The plants were grown for 124 d in a greenhouse under controlled conditions and exposed to three different treatments: (0) control; (E) external gamma irradiation delivered by a sealed 137Cs source with a dose rate of 66 micro Sv h(-1); (E+I) E treatment combined with internal beta irradiation due to contamination by 134Cs and 85Sr via root uptake from the soil. The root uptake led to a contamination of 100 Bq g(-1) for 85Sr and 5 Bq g(-1) for 134Cs (of plant dry weight) as measured at harvest. The activity of SOD, APX, GR, POD, CAT, G6PDH, and MDHAR enzymes was measured in leaves. Under (0) treatment, all enzymes showed similar activities, except POD, which had higher activity in plants originating from contaminated areas. Treatment (E) induced an enhancement of POD, CAT, GR, SOD, and G6PDH activities in plants originating from contaminated areas. Only control plants showed any stimulation of APX activity. Treatment (E+I) had no significant effect on APX, GR, CAT, and POD activities, but MDHAR activity was significantly reduced while SOD and G6PDH activities were significantly increased. The increase occurred in plants from all origins for SOD, with a greater magnitude as a function of their origin, and it occurred only in plants from the more contaminated populations for G6PDH. This suggests that exposure to a low dose rate of ionizing radiation for almost a half century in the original environment of Stipa has led to natural selection of the most adapted genotypes characterized by an efficient induction of anti-oxidant enzyme activities, especially SOD and G6PDH, involved in plant protection against reactive oxygen species.     

Estimating Total Effective Dose Equivalents from Terrorist Use of Radiological Dispersion Devices

Radiological dispersion devices (RDDs), commonly called “dirty bombs,” utilize a conventional explosive to deliberately disperse non-fissile material as an aerosol. This analysis models total effective dose equivalent (Sv) at various locations down-wind from the detonation site subsequent to terrorists detonating a ²⁴¹Am, ¹³⁷Cs, ⁶⁰Co, ¹⁹²Ir, or⁹⁰Sr RDD. A source term for each isotope equaling 3.7 × 10¹³ Bq with an instantaneous release by either high explosives or low explosives at street level is assumed in order to evaluate total effective dose equivalent (TEDE) under various meteorological scenarios for intentional releases of non-fissile materials by terrorists. The inhalation pathway on average contributes most to TEDE. The inhalation pathway accounts for 96% (0.22 Sv) of the mean exposure estimate of 0.2321 Sv and occurs over an extremely short time frame (i.e., a few minutes). Ground shine, on average, contributes the second most to TEDE estimates accounting for approximately 4% (0.009 Sv) of the estimate. A cautionary note with regard to ground shine is warranted, however, because Hotspot estimates for this pathway are based on the assumption that a person is exposed for 4 days (96 hours). The TEDE for submersion (i.e., passing through the plume without inhaling particles) is negligible for the scenarios evaluated contributing less than 1% (5.2 × 10^?6 Sv) to the TEDE estimate averaged across all 140 model runs (5 nuclides × 2 rainfall scenarios × 2 explosive scenarios × 7 wind and atmospheric stability scenarios). The TEDE value for ²⁴¹Am from inhalation is much greater, on average, than the inhalation TEDE value for ⁶⁰Co, ¹³⁷Cs, ¹⁹²Ir, or ⁹⁰Sr. This underscores the potentially high risk to human health posed by exposure to ²⁴¹Am. Ground shine is the primary exposure pathway for ⁶⁰Co and ¹³⁷Cs due to the energetic and penetrating gamma rays those radionuclides emit. ¹⁹²Ir and ⁹⁰Sr have relatively low mean TEDE values for all of the pathways examined.     

Solid cancer risk coefficient for fast neutrons in terms of effective dose

Cancer mortality risk coefficients for neutrons have recently been assessed by a procedure that postulates for the neutrons a linear dose dependence, invokes the excess risk of the A-bomb survivors at a gamma-ray dose D(1) of 1 Gy, and assumes a neutron RBE as a function of D(1) between 20 and 50. The excess relative risk (ERR) of 0.008/mGy has been obtained for R(1) = 20 and 0.016/mGy for R(1) = 50. To compare these results to the current ICRP nominal risk coefficient for solid cancer mortality (0.045/Sv for a population of all ages; 0.036/Sv for a working population), the ERR is translated into lifetime attributable risk and is then related to effective dose. The conversion is not trivial, because the neutron effective dose has been defined by ICRP not as a weighted genuine neutron dose (neutron kerma), but as a weighted dose that includes the dose from gamma rays that are induced by neutrons in the body. If this is accounted for, the solid cancer mortality risk for a working population is found to agree with the ICRP nominal risk coefficient for neutrons in their most effective energy range, 0.2 MeV to 0.5 MeV. In radiation protection practice, there is an added level of safety, because the effective dose, E, is-for monitoring purposes-assessed in terms of the operational quantity H*, which overestimates E substantially for neutrons between 0.01 MeV and 2 MeV.     

Epidemiological studies of cataract risk at low to moderate radiation doses: (not) seeing is believing

The prevailing belief for some decades has been that human radiation-related cataract occurs only after relatively high doses; for instance, the ICRP estimates that brief exposures of at least 0.5-2 Sv are required to cause detectable lens opacities and 5 Sv for vision-impairing cataracts. For protracted exposures, the ICRP estimates the corresponding dose thresholds as 5 Sv and 8 Sv, respectively. However, several studies, especially in the last decade, indicate that radiation-associated opacities occur at much lower doses. Several studies suggest that medical or environmental radiation exposure to the lens confers risk of opacities at doses well under 1 Sv. Among Japanese A-bomb survivors, risks for cataracts necessitating lens surgery were seen at doses under 1 Gy. The confidence interval on the A-bomb dose threshold for cataract surgery prevalence indicated that the data are compatible with a dose threshold ranging from none up to only 0.8 Gy, similar to the dose threshold for minor opacities seen among Chernobyl clean-up workers with primarily protracted exposures. Findings from various studies indicate that radiation risk estimates are probably not due to confounding by other cataract risk factors and that risk is seen after both childhood and adult exposures. The recent data are instigating reassessments of guidelines by various radiation protection bodies regarding permissible levels of radiation to the eye. Among the future epidemiological research directions, the most important research need is for adequate studies of vision-impairing cataract after protracted radiation exposure.     

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 μSv mg⁻¹ DU. This compares to an effective dose coefficient for an adult of 0.71 μSv mg⁻¹ for DU materials given by the World Health Organization (WHO). The effective dose coefficient for inhalation was calculated to be 3.7 × 10⁻⁶ Sv Bq⁻¹ for workers and 5.3 × 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.     

Flexible dose-response models for Japanese atomic bomb survivor data: Bayesian estimation and prediction of cancer risk

Generalised absolute risk models were fitted to the latest Japanese atomic bomb survivor cancer incidence data using Bayesian Markov Chain Monte Carlo methods, taking account of random errors in the DS86 dose estimates. The resulting uncertainty distributions in the relative risk model parameters were used to derive uncertainties in population cancer risks for a current UK population. Because of evidence for irregularities in the low-dose dose response, flexible dose-response models were used, consisting of a linear-quadratic-exponential model, used to model the high-dose part of the dose response, together with piecewise-linear adjustments for the two lowest dose groups. Following an assumed administered dose of 0.001 Sv, lifetime leukaemia radiation-induced incidence risks were estimated to be 1.11 x 10(-2) Sv(-1) (95% Bayesian CI -0.61, 2.38) using this model. Following an assumed administered dose of 0.001 Sv, lifetime solid cancer radiation-induced incidence risks were calculated to be 7.28 x 10(-2) Sv(-1) (95% Bayesian CI -10.63, 22.10) using this model. Overall, cancer incidence risks predicted by Bayesian Markov Chain Monte Carlo methods are similar to those derived by classical likelihood-based methods and which form the basis of established estimates of radiation-induced cancer risk.     

Radioecological research of the Irtysh-River's and Ob-River's water, base and flood depositions

The radioecological research of Irtysh-river and Ob-river was held. The content of 137Cs in Irtysh water was compounded 0.62-1.23 Bq/m3, in Ob-- 0.24-0.27 Bq/m3, and the one of 90Sr in Irtysh-- 10-20 Bq/m3, and in Ob-- 5-10 Bq/m3, that is much lower than the permissible sanitary-hygienic norms for the population. The 137Cs stores density on Irtysh-river input lease was compounded 2.7 kBq/m2, is almost in 11 times slashed downstream and is peer 245 kBq/m2 before the Irtysh-river lockin. The 90Sr stores density also was slashed in surveyed leases with 212 down to 106 Bq/m2. Two variants of integrated stores of 137Cs and of 90Sr in flood of the Irtysh-river was held. The balance calculation of annual radionuclides sinks confirms the dominant amount of 137Cs and of 90Sr in downstream Ob-river leases acts now on the Ob's sleeve, instead of from the Irtysh-river as it was supposed earlier. The 137Cs medial annual inflow from the Ob's sleeve almost is in 2 times, and the 90Sr inflow is in 2.3 times more, than are acts from Irtysh-river sleeve.     

Track structures and dose distributions from decays of (131)I and (125)I in and around water spheres simulating micrometastases of differentiated thyroid cancer

The disintegration of the radionuclides (131)I and (125)I and the subsequent charged-particle tracks left behind in water (as a model substance for a biological cell) are simulated by the Monte Carlo track structure simulation code PARTRAC, using new inelastic electron scattering cross sections for condensed water. Every photon and electron emitted was followed in detail, event by event, down to 10 eV. From the spatial information on the track structures, absorbed dose distributions per (131)I and (125)I decay were calculated in and around water spheres simulating micrometastases as well as in the tissue surrounding such metastases. These radionuclides were assumed to be distributed uniformly inside spheres of different diameters (0.01, 0.03, 0.1, 0.3, 1.0 and 3.0 mm). The respective electron degradation spectra, the nearest-neighbor distance distributions between inelastic events, and the distance distributions for all activations for both iodine radionuclides were calculated. The absorbed fractions of the initial electron energies, absorbed doses and energy depositions, and single-event distributions, F(1)(epsilon), inside the six water spheres described above and in the surrounding tissue were also calculated. The absorbed doses per decay inside the six water spheres, i.e., the calculated S values (listed from 0.01 to 3.0 mm), were 6.8 x 10(-4), 7.2 x 10(-5), 5.5 x 10(-6), 4.9 x 10(-7), 3.1 x 10(-8) and 1.8 x 10(-9) Gy Bq(-1) s(-1) for (131)I, and 3.4 x 10(-3), 1.7 x 10(-4), 5.1 x 10(-6), 2.0 x 10(-7), 5.6 x 10(-9) and 2.2 x 10(-10) Gy Bq(-1) s(-1) for (125)I. It is concluded that, in the treatment of thyroid cancer, the geometrical track structure properties of (125)I might be superior to those of (131)I in micrometastases with diameters less than 0.1 mm; however, in this medical context, many other factors also have to be considered.     

Reconstruction of 90Sr intake for breast-fed infants in the Techa riverside settlements

The Techa River (Southern Urals, Russia) was contaminated as a result of radioactive releases by the Mayak plutonium production facility during 1949-1956. The persons born after the onset of the contamination have been identified as the "Techa River Offspring Cohort" (TROC). The TROC has the potential to provide direct data on health effects in progeny that resulted from exposure of a general parent population to chronic radiation. The purpose of the present investigation is the estimation of (90)Sr intake from breast milk and river water in the period from birth to 6 months of life, necessary for an infant dose calculation. The investigation is based on all available data concerning radioactive contamination due to global fallouts and Mayak releases in the Southern Urals where extensive radiometric and radiochemical investigations of human tissues and environmental samples were conducted during the second half of the twentieth century. The strontium transfer factor from mother's daily diet to breast milk was estimated as 0.05 (0.01-0.13) d L(-1). Based on this transfer factor and data on (90)Sr water contamination, the average total (90)Sr intake for an infant born in the middle Techa River region was found to be equal to 60-80 kBq in 1950-1951. For the same period, calculations of (90)Sr intake using ICRP models gave values of 70-100 kBq. From 1952 onwards, the differences in intakes calculated using the two approaches increased, reaching a factor of 2-3 in 1953. The Techa River data provide the basis for improving and adapting the ICRP models for application to Techa River-specific population.     

Reconstruction and forecast of doses due to ingestion of <Superscript>137</Superscript>Cs and <Superscript>90</Superscript>Sr after the Chernobyl accident

The assessment doses due to ingestion of ¹³⁷Cs and ⁹⁰Sr for the population suffering from the Chernobyl accident was performed on the basis of the new mechanistic ecological model for assessment of radiological consequences of agricultural lands contamination (EMARC). The EMARC model allows estimation of internal doses based on ecological factors influencing the contamination of foodstuff, for the post-accidental years in the countries of the former Soviet Union. The EMARC model allows estimation of all quantities required in radiation hygiene practice. For example, the proposed analytical method may be used for both retrospective dose reconstruction and prospective estimates of annual dose and integrated “life-time” dose, for different age intervals. According to the EMARC model, estimated reference “life-time” doses for adults are between 7 and 269 μSv kBq⁻¹ m² for ¹³⁷Cs, and between 25 and 235 μSv kBq⁻¹ m² for ⁹⁰Sr. Maximal doses were estimated for persons who were 3, 9 and 11 years old, at the time of the accident and these doses exceed those for adults by a factors of 1, 5 for ⁹⁰Sr, and 1.4 for ¹³⁷Cs.     

Doses from external irradiation and ingestion of 134Cs, 137Cs and 90Sr of the population of Belarus accumulated over 35 years after the Chernobyl accident

This study considers the exposure of the population of the most contaminated Gomel and Mogilev Oblasts in Belarus to prolonged sources of irradiation resulting from the Chernobyl accident. Dose reconstruction methods were developed and applied in this study to estimate the red bone-marrow doses (RBMs) from (i) external irradiation from gamma-emitting radionuclides deposited on the ground and (ii) ¹³⁴Cs, ¹³⁷Cs and ⁹⁰Sr ingestion with locally produced foodstuffs. The mean population-weighted RBM doses accumulated during 35 years after the Chernobyl accident were 12 and 5.7 mGy for adult residents in Gomel and Mogilev Oblasts, respectively, while doses for youngest age groups were 20–40% lower. The highest mean area-specific RBM doses for adults accumulated in 1986–2021 were 63, 56 and 46 mGy in Narovlya, Vetka and Korma raions in Gomel Oblast, respectively. For most areas, external irradiation was the predominant pathway of exposure (60–70% from the total dose), except for areas with an extremely high aggregated ¹³⁷Cs soil to cow’s milk transfer coefficient (≥?5.0 Bq L^?1 per kBq m^?2), where the contribution of ¹³⁴Cs and ¹³⁷Cs ingestion to the total RBM dose was more than 70%. The contribution of ⁹⁰Sr intake to the total RBM dose did not exceed 4% for adults and 10% for newborns in most raion in Gomel and Mogilev Oblasts. The validity of the doses estimated in this study was assessed by comparison with doses obtained from measurements by thermoluminescence dosimeters and whole-body counters done in 1987–2015. The methodology developed in this study can be used to calculate doses to target organs other than RBM such as thyroid and breast doses. The age-dependent and population-weighted doses estimated in this study are useful for ecological epidemiological studies, for projection of radiation risk, and for justification of analytical epidemiological studies in populations exposed to Chernobyl fallout.

     

The effect of mass on the gastrointestinal absorption of plutonium and neptunium

Absorption and retention of plutonium were determined in mice after intragastric administration of either 6 X 10(-4) or 1.5 mg/kg in bicarbonate, citrate, or nitrate media. At the higher concentration, absorption of the citrate was greater than that of the nitrate; at the lower concentration, chemical form was not an important factor in absorption. Concentration and chemical form had much less influence on absorption by the neonatal (versus the adult) rat. The transfer factor (f1) for neonates was between one and two orders of magnitude higher than for adults. Absorption and retention of neptunium were determined in rats and/or mice after intragastric administration at doses ranging from 2.2 X 10(-7) to 43 mg/kg in nitrate solutions of pH 1.5. At the higher concentrations, absorption was 1.5 to 2.7%. For lower concentrations, absorption was 25 to 65 times less. In contrast to results obtained in adult animals, absorption of neptunium by neonates decreased with increasing dose. The data obtained in adult animals suggest that the f1 factor recommended by the ICRP for plutonium should be increased by a factor of 10, but the neptunium f1 factor, in contrast, should be decreased by a factor of 10.     

Spatial distributions and dose assessment of natural radionuclides in rocks and soils of some selected sites in southwestern Nigeria

The activity concentrations of ²³⁸U, ²³²Th, and ⁴⁰K in rock and soil samples collected from Ondo and Ekiti States in southwestern Nigeria were measured by using gamma-ray spectrometer with a high-purity germanium (HPGe) detector. The mean activity concentrations of ²³⁸U, ²³²Th, and ⁴⁰K in the rock were 25.53, 61.12, and 554.20 Bq kg^?1 respectively, while that of the soil were 8.27, 17.37, and 151.72 Bq kg^?1 respectively. Results showed that the activity concentrations of ²³⁸U, ²³²Th, and ⁴⁰K were higher in the rocks than the soils of the areas studied. To assess the radiological impact of some radionuclides on the population in the region, the annual effective dose equivalent (AEDE), annual gonadal equivalent dose (AGED), and excess lifetime cancer risk (ELCR) were calculated. The mean values of the indoor and outdoor AEDE, and AGED were 88.08 µSv y^?1, 352.34 µSv y^?1, and 508.40 mSv y^?1, respectively for the rock samples, and 25.31 µSv y^?1, 101.25 µSv y^?1, and 145.80 mSv y^?1 respectively for the soil samples. The mean values obtained for AEDE and AGED for the soil were below ICRP recommended limits of 1 and 300 mSv y^?1, respectively. AGED for the rocks was higher than the maximum permissible limit.