Absorbed dose in target cell nuclei and dose conversion coefficient of radon progeny in the human lung

To calculate the absorbed dose in the human lung due to inhaled radon progeny, ICRP focussed on the layers containing the target cells, i.e., the basal and secretory cells. Such an approach did not consider details of the sensitive cells in the layers. The present work uses the microdosimetric approach and determines the absorbed alpha-particle energy in non-spherical nuclei of target cells (basal and secretory cells). The absorbed energy for alpha particles emitted by radon progeny in the human respiratory tract was calculated in basal- and secretory-cell nuclei, assuming conical and ellipsoidal forms for these cells. Distributions of specific energy for different combinations of alpha-particle sources, energies and targets are calculated and shown. The dose conversion coefficient for radon progeny is reduced for about 2mSv/WLM when conical and ellipsoidal cell nuclei are considered instead of the layers. While changes in the geometry of secretory-cell nuclei do not have significant effects on their absorbed dose, changes from spherical to conical basal-cell nuclei have significantly reduced their absorbed dose from approximately 4 to approximately 3mGy/WLM. This is expected because basal cells are situated close to the end of the range of 6MeV alpha particles. This also underlines the significance of better and more precise information on targets in the T-B tree. A further change in the dose conversion coefficient can be achieved if a different weighting scheme is adopted for the doses for the cells. The results demonstrate the necessity for better information on the target cells for more accurate dosimetry for radon progeny.     

Methods for calculating dose conversion coefficients for terrestrial and aquatic biota

Plants and animals may be exposed to ionizing radiation from radionuclides in the environment. This paper describes the underlying data and assumptions to assess doses to biota due to internal and external exposure for a wide range of masses and shapes living in various habitats. A dosimetric module is implemented which is a user-friendly and flexible possibility to assess dose conversion coefficients for aquatic and terrestrial biota. The dose conversion coefficients have been derived for internal and various external exposure scenarios. The dosimetric model is linked to radionuclide decay and emission database, compatible with the ICRP Publication 38, thus providing a capability to compute dose conversion coefficients for any nuclide from the database and its daughter nuclides. The dosimetric module has been integrated into the ERICA Tool, but it can also be used as a stand-alone version.     

Radioactive particles in dose assessments

Radioactive particles present a novel exposure pathway for members of the public. For typical assessments of potential doses received by members of the public, habit surveys and environmental monitoring combine to allow the assessment to occur. In these circumstances it is believed that the probability of encounter/consumption is certain. The potential detriment is assessed through sampling the use of environmental monitoring data and dose coefficients such as that in ICRP 60 [ICRP, 1990. 1990 Recommendations of the international commission on radiological protection. Publication 60. Annals of the ICRP 21 (1-3)]. However, radioactive particles often represent a hazard that is difficult to quantify and where the probability of encounter is less than certain as are the potential effects on health. Normal assessment methodologies through sampling and analysis are not appropriate for assessing the impact of radioactive particles either prospectively or retrospectively. This paper details many of the issues that should be considered when undertaking an assessment of the risk to health posed by radioactive particles.     

In vitro dissolution studies of uranium bearing material in simulated lung fluid

Inhaled uranium (U) bearing material will partially dissolve in the fluid lining of the lung, followed by a combination of retention, re-distribution, and excretion of the U. The rate of dissolution influences the retention time at the site of deposition, and the extent to which the material is available for re-distribution to other tissues. The consequential radiation dose is dependent upon the material distribution in the body and the exposure time to various tissues. The International Commission on Radiological Protection, ICRP 66 [International Commission on Radiological Protection (ICRP), 1994. Human Respiratory Tract Model for Radiological Protection. ICRP Publication 66] recommends the use of experimentally determined solubility coefficients in dose modelling. Material specific absorption parameters allow for better dose estimation than using ICRP default values for F (fast), M (moderate) and S (slow) classifications of U compounds. In vitro dissolution tests were carried out on U material collected from two U mines located in Australia. A static technique was designed in which particle samples were sandwiched between two 0.1-mum pore size membrane filters. The filter sandwich was exposed to a solvent (simulated lung fluid) for selected time intervals, at controlled test conditions for temperature and pH. The collected solution was analysed for U concentration using ICP-MS. The resulting dissolution curves were fitted with a double or triple exponential equation to determine the dissolution coefficients.     

Doses of external exposure in Jordan house due to gamma-emitting natural radionuclides in building materials

The use of building materials containing naturally occurring radionuclides as ⁴⁰K, ²³²Th, and ²³⁸U and their progeny results in external exposures of the residents of such buildings. In the present study, indoor dose rates for a typical Jordan concrete room are calculated using Monte Carlo method. Uniform chemical composition of the walls, floor and ceiling as well as uniform mass concentrations of the radionuclides in walls, floor and ceiling are assumed.Using activity concentrations of natural radionuclides typical for the Jordan houses and assuming them to be in secular equilibrium with their progeny, the maximum annual effective doses are estimated to be 0.16, 0.12 and 0.22 mSv a⁻¹ for ⁴⁰K, ²³²Th- and ²³⁸U-series, respectively. In a total, the maximum annual effective indoor dose due to external γ-radiation is 0.50 mSv a⁻¹. Additionally, organ dose coefficients are calculated for all organs considered in ICRP Publication 74. Breast, skin and eye lenses have the maximum equivalent dose rate values due to indoor exposures caused by the natural radionuclides, while equivalent dose rates for uterus, colon (LLI) and small intestine are found to be the smallest. More specifically, organ dose rates (nSv a⁻¹ per Bq kg⁻¹) vary from 0.044 to 0.060 for ⁴⁰K, from 0.44 to 0.60 for radionuclides from ²³⁸U-series and from 0.60 to 0.81 for radionuclides from ²³²Th-series.The obtained organ and effective dose conversion coefficients can be conveniently used in practical dose assessment tasks for the rooms of similar geometry and varying activity concentrations and local-specific occupancy factors.     

Inhalation dose assessment of indoor radon progeny using biokinetic and dosimetric modeling and its application to Jordanian population

High indoor radon concentrations in Jordan result in internal exposures of the residents due to the inhalation of radon and its short-lived progeny. It is therefore important to quantify the annual effective dose and further the radiation risk to the radon exposure. This study describes the methodology and the biokinetic and dosimetric models used for calculation of the inhalation doses exposed to radon progeny. The regional depositions of aerosol particles in the human respiratory tract were firstly calculated. For the attached progeny, the activity median aerodynamic diameters of 50 nm, 230 nm and 2500 nm were chosen to represent the nucleation, accumulation and coarse modes of the aerosol particles, respectively. For the unattached progeny, the activity median thermodynamic diameter of 1 nm was chosen to represent the free progeny nuclide in the room air. The biokinetic models developed by the International Commission on Radiological Protection (ICRP) were used to calculate the nuclear transformations of radon progeny in the human body, and then the dosimetric model was applied to estimate the organ equivalent doses and the effective doses with the specific effective energies derived from the mathematical anthropomorphic phantoms. The dose conversion coefficient estimated in this study was 15 mSv WLM⁻¹ which was in the range of the values of 6-20 mSv WLM⁻¹ reported by other investigators. Implementing the average indoor radon concentration in Jordan, the annual effective doses were calculated to be 4.1 mSv y⁻¹ and 0.08 mSv y⁻¹ due to the inhalation of radon progeny and radon gas, respectively. The total annual effective dose estimated for Jordanian population was 4.2 mSv y⁻¹. This high annual effective dose calculated by the dosimetric approach using ICRP biokinetic and dosimetric models resulted in an increase of a factor of two in comparison to the value by epidemiological study. This phenomenon was presented by the ICRP in its new published statement on radon.     

Background exposure rates of terrestrial wildlife in England and Wales

It has been suggested that, when assessing radiation impacts on non-human biota, estimated dose rates due to anthropogenically released radionuclides should be put in context by comparison to dose rates from natural background radiation. In order to make these comparisons, we need data on the activity concentrations of naturally occurring radionuclides in environmental media and organisms of interest. This paper presents the results of a study to determine the exposure of terrestrial organisms in England and Wales to naturally occurring radionuclides, specifically (40)K, (238)U series and (232)Th series radionuclides. Whole-body activity concentrations for the reference animals and plants (RAPs) as proposed by the ICRP have been collated from literature review, data archives and a targeted sampling campaign. Data specifically for the proposed RAP are sparse. Soil activity concentrations have been derived from an extensive geochemical survey of the UK. Unweighted and weighted absorbed dose rates were estimated using the ERICA Tool. Mean total weighted whole-body absorbed dose rates estimated for the selected terrestrial organisms was in the range 6.9 x 10(-2) to 6.1 x 10(-1) microGy h(-1).     

Enhancement of population doses due to production of electricity from brown coal in Poland

Changes in natural radioactivity as a result of the operations of three Polish brown-coal-fired power plants have been assessed via calculations based on the results of radio-spectrometric measurements of coal and fly ash, information on local climatic conditions and data on atmospheric releases from the plants. Calculations were performed using a computer code derived from solution of the Pasquille equation. Transfer coefficients and conversion factors of absorbed radioactivity to effective dose equivalent (EDE) were selected using literature data. Values for EDE due to inhalation, ingestion and external gamma radiation from radionuclides deposited on the ground were calculated for the population inhabiting the most polllated area and also for the whole population. The individual maximum EDE for the most ‘radioactive’ power plant was estimated at 10·4 μSv y⁻¹. This EDE is less than 1% of the total natural radiation burden of 2 mSv y⁻¹. EDE estimates for the vicinities of the other two power plants are 0·5 and 4·2 μSv y⁻¹. The collective EDE resulting from the production of electrical energy from brown coal was calculated to be 104 man-Sv y⁻¹, i.e. 14 man-Sv per GW-year. It is concluded that some of the brown-coal-fired power plants in Poland should improve their fly ash control and that the manner in which existing transfer coefficients are reported in the literature does not lead to unequivocal results when applied to EDE calculations.     

Effect of low continuous 99Tc intake on its absorption and metabolism in young rats

A continuous daily oral dose of (99)TcO(4)(-) (5 Bq g(-1) day(-1)) was given to young rats for 56 days. In one group the intake was continued for 98 days, while in the other the intake was stopped to facilitate the determination of biological half lives in different organs and tissues. The absorbed fraction of (99)Tc increased from 0.7 to 0.9 during the experimental period (m=0.85 sigma=0.02). The whole body retained fraction (m=0.15 sigma=0.05) indicated a storage in the tissues. The liver and kidneys showed increasing concentrations of (99)Tc until a plateau was reached after 6 and 9 weeks, then a rapid decline after the intake was stopped (half life of 6 and 5 days respectively). In hair and thyroid tissue, the accumulation of (99)Tc increased linearly then declined very slowly after the intake was stopped and did not reach a plateau during the experiment. No significant amounts of (99)Tc were detected in the muscular tissues or in the walls of the digestive tract. After the contamination period, urinary and fecal excretions fell off very quickly (1 week), but significant, though low, quantities of urinary and fecal excretions were still observed 6 weeks after stopping intake.     

Bronchial dosimeters for radon progeny for Chinese males and females

Bronchial dosimeters have been designed for adult Chinese males and females for home and mine exposures, which can give the bronchial doses from radon progeny by direct measurements. The bronchial dosimeter for home exposures consists of five 400-mesh wire screens. With a sampling face velocity of 3.3 cm s(-1) for Chinese males and 2.7 cm s(-1) for Chinese females, the deposition pattern on the wire screens were found to satisfactorily match the variation of the dose conversion coefficients (in units of mSv WLM(-1)) with the size of radon progeny from 1 to 1000 nm. The bronchial dosimeter for mine exposures consists of four 250-mesh wire screens. With a sampling face velocity of 3.3 cm s(-1), the deposition pattern on the wire screens were found to satisfactorily match the variation of the dose conversion coefficients for both Chinese males and females. In this way, the bronchial dosimeters directly give the bronchial doses from the alpha counts recorded on the wire-screens.     

A study on natural radiation exposure in different realistic living rooms

In the first part of the paper, the factors affecting 222Rn properties in 25 different realistic living rooms (with low ventilation rates) of different houses in El-Minia City (Upper Egypt) have been studied; they included the activity concentration of 222Rn gas (C(o)), the unattached fraction (f(p)) of 218Po and 214Pb, the unattached potential alpha energy concentration (PAEC) and the equilibrium factor (F). The activity distributions of unattached 218Po and 214Pb as well as for the PAEC were determined. With a dosimetric model calculation [ICRP, 1994b. Human Respiratory Tract Model For Radiological Protection. Pergamon Press, Oxford. ICRP Publication 66] the total deposition fraction of unattached 218Po and 214Pb in human respiratory tract was evaluated to determine the total equivalent dose. An electrostatic precipitation method and a wire screen diffusion battery technique were both employed for the determination of 222Rn gas concentration and its unattached decay products, respectively. The mean activity concentration of 222Rn gas (C(o)) was found to be 110+/-20 Bq m(-3). The mean unattached activity concentrations of 218Po and 214Pb were found to be 0.6 and 0.35 Bq m(-3), respectively. A mean unattached fraction (f(p)) of 0.09+/-0.01 was obtained at a mean aerosol particle concentration (Z) of (2.9+/-0.23) x 10(3)cm(-3). The mean equilibrium factor (F) was determined to be 0.31+/-0.02. The mean PAEC of unattached 218Po and 214Pb was found to be 8.74+/-2.1 Bq m(-3). The activity distributions of 218Po and 214Pb show mean activity median diameters (AMD) of 1.5 and 1.85 nm with mean geometric standard deviations (SD) of 1.33 and 1.45, respectively. The mean activity distribution of the PAEC shows an AMD of 1.65 nm with a geometric standard deviation of 1.25. At a total deposition fraction of about 97% the total equivalent dose to the lung was determined to be about 133 microSv. The second part of this paper deals with a study of natural radionuclide contents of samples collected from the building materials of the rooms under investigations in part one. Analyses were performed in Marinelli beakers with a gamma multichannel analyzer equipped with a NaI(Tl) detector. The samples revealed the presence of the uranium-radium and thorium radioisotopes as well as 40K. Nine gamma-lines of the natural radioisotopes corresponding to 212Pb, 214Pb, 214Bi, 228Ac, 40K and 208Tl were detected and measured. The activity concentrations of 226Ra, 232Th and 40K were determined with mean activity concentrations of 58+/-19, 31+/-11 and 143+/-62 Bq kg(-1), respectively. These activities amount to a radium equivalent (Ra(eq)) of 113 Bq kg(-1) and to a mean value of external hazard index of 0.31.     

Choosing an alpha radiation weighting factor for doses to non-human biota

The risk to non-human biota from exposure to ionizing radiation is of current international interest. In calculating radiation doses to humans, it is common to multiply the absorbed dose by a factor to account for the relative biological effectiveness (RBE) of the radiation type. However, there is no international consensus on the appropriate value of such a factor for weighting doses to non-human biota. This paper summarizes our review of the literature on experimentally determined RBEs for internally deposited alpha-emitting radionuclides. The relevancy of each experimental result in selecting a radiation weighting factor for doses from alpha particles in biota was judged on the basis of criteria established a priori. We recommend a nominal alpha radiation weighting factor of 5 for population-relevant deterministic and stochastic endpoints, but to reflect the limitations in the experimental data, uncertainty ranges of 1-10 and 1-20 were selected for population-relevant deterministic and stochastic endpoints, respectively.     

Uncertainties of internal dose assessment for animals and plants due to non-homogeneously distributed radionuclides

The dose conversion coefficients (DCCs) for the assessment of internal absorbed dose rate in reference animals and plants have been generally calculated assuming a homogeneous distribution of radionuclides within the body. Realistic scenarios of internal exposure must account for some radionuclides which tend to concentrate in specific organs or tissues. To study the effect of such inhomogeneous distributions, internal DCCs have been calculated assuming both a central and an eccentric point source. The analysis of the results showed that uncertainties of the whole body DCC due to non-homogeneous radionuclide distribution are less than 30% for photons and electrons for all considered organisms. For electrons, the uncertainties are negligible below certain energies, dependent on the size of the organisms. Additionally, the organ doses due to the accumulation of the radionuclide in an organ are also described and organ/whole body doses ratios are estimated.     

Radon progeny dose conversion coefficients for Chinese males and females

The airway dimensions for Caucasian males have been scaled by multiplying by factors 0.95 and 0.88 to give those for Chinese males and females, respectively. Employing the most recent data on physical and biological parameters, the radiation doses to the basal and secretory cells due to alpha particles from 218Po and 214Po, homogeneously distributed in the mucous layer, have been calculated. The emission of alpha particles has been simulated by a Monte Carlo method. For both basal and secretory cells, the dose conversion coefficients (DCCs) for physical conditions of sleep, rest, light and heavy exercise, have been obtained for Chinese males and females for unattached progeny, and for attached progeny of diameters 0.02, 0.15, 0.25, 0.30 and 0.50 micron. For basal cells, the coefficients lie in the range 0.69-6.82 mGy/(Js/m3) or 8.7-86 mGy/WLM for unattached progeny and in the range 0.045-1.98 mGy/(Js/m3) or 0.57-25 mGy/WLM for attached progeny. The corresponding ranges for Caucasian males are 1.27-8.81 mGy/(Js/m3) or 16-111 mGy/WLM-1 and 0.05-2.30 mGy/(Js/m3) or 0.64-29 mGy/WLM. For secretory cells, the coefficients lie in the range 0.095-16.82 mGy/(Js/m3) (1.2-212 mGy/WLM) for unattached progeny and in the range 0.095-6.67 mGy/(Js/m3) (1.2-84 mGy/WLM) for attached progeny. The corresponding ranges for Caucasian males are 0.34-21.51 mGy/(Js/m3) (4.3-271 mGy/WLM) and 0.1-7.78 mGy/(Js/m3) (1.3-98 mGy/WLM). The overall DCCs calculated for a typical home environment are 0.59 and 0.52 mSv/(Js/m3) (7.4 and 6.5 mSv/WLM) for Chinese males and females, respectively, which are 80 and 70% of the value, 0.73 mSv/(Js/m3) (9.2 mSv/WLM), for Caucasian males.     

Contribution of (222)Rn-bearing water to the occupational exposure in thermal baths

The present study investigates the short- and long-term effects of radon ((222)Rn) released from water on the progeny exposure in a thermal spa. For the purposes of this work, the Polichnitos spa was used as a case study. The bathroom was supplied with water containing 110-210 kBq m(-3) of (222)Rn. The (222)Rn concentration in air and the short-lived (222)Rn progenies in attached and unattached form were monitored into the bathroom and the surrounding premises. The equilibrium factor (F-factor) and the unattached fraction were estimated. The results of this study show that water flow during bath filling is by far the dominant mechanism by which (222)Rn is released in the air of the bathroom. The progeny exposure was correlated linearly with the (222)Rn concentration in the entering water. The annual effective dose received by a worker was found to be below the lower limit value of 3 mSv recommended by ICRP 65. The dose limit was exceeded only for water containing more than 300 kBq m(-3).     

Bronchial Rn dose survey for residences

The bronchial dosimeter for Rn progeny proposed by Yu and Guan in 1998 was employed to survey the bronchial dose from Rn progeny in 30 residences in Hong Kong. An average bronchial deposition fraction of Rn progeny was obtained as 0.0334, which gave an average dose conversion factor (DCF) of 8.5 mSv WLM-1. The mean values of potential alpha energy concentration (PAEC) deposited in the tracheobronchial region (PAECT-B), total PAEC in air (PAECT), annual effective dose (E), concentration of Rn gas (RC) and annual dose conversion factor (ADCF) for all the residential sites combined were 0.11 +/- 0.05, 3.1 +/- 1.4 mWL, 1.2 +/- 0.5 mSv yr-1, 23 +/- 10 Bq m-3 and 0.055 +/- 0.020 (mSv yr-1 per Bqm-3), respectively, with air-conditioned sites (AC sites) and non-AC sites having significantly different mean ADCF values. The indoor relative humidity affected PAECT and RC with high confidence levels (> 95%).     

The influence of geographic location on population exposure to emissions from power plants throughout China

This analysis seeks to evaluate the influence of emission source location on population exposure in China to fine particles and sulfur dioxide. We use the concept of intake fraction, defined as the fraction of material or its precursor released from a source that is eventually inhaled or ingested by a population. We select 29 power-plant sites throughout China and estimate annual average intake fractions at each site, using identical source characteristics to isolate the influence of geographic location. In addition, we develop regression models to interpret the intake fraction values and allow for extrapolation to other sites. To model the concentration increase due to emissions from selected power plants, we used a detailed long-range atmospheric dispersion model, CALPUFF. Primary fine particles have the highest average intake fraction (1 x 0(-5)), followed by sulfur dioxide (5 x 10(-6)), sulfate from sulfur dioxide (4 x 10(-6)), and nitrate from nitrogen oxides (4 x 10(-6)). For all pollutants, the intake fractions span approximately an order of magnitude across sites. In the regression analysis, the independent variables are meteorological proxies (such as climate region and precipitation) and population at various distances from the source. We find that population terms can explain a substantial percentage of variability in the intake fraction for all pollutants (R(2) between 0.86 and 0.95 across pollutants), with a significant modifying influence of meteorological regime. Near-source population is more important for primary coarse particles while population at medium to long distance is more important for primary fine particles and secondary particles. A significant portion of intake fraction (especially for secondary particles and primary fine particles) occurs beyond 500 km of the source, emphasizing the need for detailed long-range dispersion modeling. These findings demonstrate that intake fractions for power plants in China can be estimated with reasonable precision and summarized using simple regression models. The results should be useful for informing future decisions about power-plant locations and controls.     

Results of the European Commission MARINA II study: part I--general information and effects of discharges by the nuclear industry

From the collated data relevant to discharges by the nuclear industry, it results that the input of beta activity (excluding Chernobyl fallout and tritium) into the OSPAR region decreased by a factor of 4 from 1986 to 1991, reaching by this date the same level as in the early 1950s. Over the same period the discharges of the alpha activity into the OSPAR region also decreased by a factor 3, the same trend has been seen also for tritium. Since 1986 the effective dose to members of the critical group in the vicinity of Sellafield and Cap de La Hague was consistently below the ICRP and EU limit of 1 mSv per year to members of the general public. The overall radiological impact from nuclear industry on the population of the European Union from the OSPAR area has decreased from 280 manSv y(-1) in 1978 to 14 manSv y(-1) in 2000.     

Natural radionuclide content in building materials and gamma dose rate in dwellings in Cuba

An extensive research project to investigate the radioactive properties of Cuban building materials was carried out because there is a lack of information on the radioactivity of such materials in Cuba. In the framework of this project 44 samples of commonly used raw materials and building products were collected in five Cuban provinces. The activity concentrations of natural radionuclides were determined by gamma ray spectrometry using a p-type coaxial high purity germanium detector and their mean values were in the ranges: 9-857Bqkg(-1) for (40)K; 6-57Bqkg(-1) for (226)Ra; and 1.2-22Bqkg(-1) for (232)Th. The radium equivalent activity in the 44 samples varied from 4Bqkg(-1) (wood) to 272Bqkg(-1) (brick). A high pressure ionisation chamber was used to measure the indoor absorbed dose rate in 543 dwellings and workplaces in five Cuban provinces. The average absorbed dose rates in air ranged from 43nGyh(-1) (Holguín) to 73nGyh(-1) (Camagüey) and the corresponding population-weighted annual effective dose due to external gamma radiation was estimated to be 145+/-40microSv. This value is 51% lower than the effective dose due to internal exposure from inhalation of decay products of (222)Rn and (220)Rn and it is 16% higher than the calculated value for the typical room geometry of a Cuban house.     

The ERICA Tool

The ERICA Tool is a computerised, flexible software system that has a structure based upon the ERICA Integrated Approach to assessing the radiological risk to biota. The Tool guides the user through the assessment process, recording information and decisions and allowing the necessary calculations to be performed to estimate risks to selected animals and plants. Tier 1 assessments are media concentration based and use pre-calculated environmental media concentration limits to estimate risk quotients. Tier 2 calculates dose rates but allows the user to examine and edit most of the parameters used in the calculation including concentration ratios, distribution coefficients, percentage dry weight soil or sediment, dose conversion coefficients, radiation weighting factors and occupancy factors. Tier 3 offers the same flexibility as Tier 2 but allows the option to run the assessment probabilistically if the underling parameter probability distribution functions are defined. Results from the Tool can be put into context using incorporated data on dose-effects relationships and background dose rates.