MWDS-2016: THE SLOW DISSOLUTION RATE FOR PLUTONIUM NITRATE INTAKES AT THE MAYAK FACILITY

The slow dissolution rate of material deposited in the lung plays a key role in determining the eventual radiation dose received by the lung. It is therefore of great importance to establish a reliable value for this parameter, to incorporate into the latest Mayak Worker Dosimetry System (MWDS-2016). Disparate values have been obtained for the slow dissolution rate of plutonium nitrate. A volunteer study performed by Public Health England (PHE) and an analysis of United States Transuranium and Uranium Registries (USTUR) case 0269 have yielded slow dissolution rates in the region of 10-40 × 10-4 d-1. However, autopsies performed on 20 Mayak workers, exposed predominantly to nitrates, have resulted in estimates of slow dissolution rates of around 2.4 × 10-4 d-1. Three hypotheses have been proposed to explain this discrepancy: (1) a slower dissolution rate in the interstitium, (2) a third exponential component in the dissolution function and (3) a small component of oxide in the aerosol to which Mayak 'nitrate' workers were exposed. This paper describes tests of these competing hypotheses. Bayesian methods have been applied to the following datasets: PHE volunteer data; Beagle dog data; USTUR cases and Mayak worker data. It is concluded that a mixture of oxide and nitrate material, with the oxide forming ~14% of the intake, best describes the Mayak dissolution rate, without introducing values for other parameters which conflict with other studies.

The Mayak Worker Dosimetry System (MWDS 2013): Soluble Plutonium Retention in the Lungs of An Occupationally Exposed USTUR Case

For the first time, plutonium retention in human upper airways was investigated based on the dosimetric structure of the human respiratory tract proposed by the International Commission on Radiological Protection (ICRP). This paper describes analytical work methodology, case selection criteria, and summarizes findings on soluble (ICRP 68 Type M material) plutonium distribution in the lungs of a former nuclear worker occupationally exposed to plutonium nitrate [239Pu(NO3)4]. Thirty-eight years post-intake, plutonium was found to be uniformly distributed between bronchial (BB), bronchiolar (bb) and alveolar-interstitial (AI) dosimetric compartments as well as between the left and right lungs. 239+240Pu and 238Pu total body activity was estimated to be 2333 ± 23 and 42.1 ± 0.7 Bq, respectively. The results of this work provide key information on the extent of plutonium binding in the upper airways of the human respiratory tract.

The Mayak Worker Dosimetry System (MWDS-2013): Implementation of the Dose Calculations

The calculation of internal doses for the Mayak Worker Dosimetry System (MWDS-2013) involved extensive computational resources due to the complexity and sheer number of calculations required. The required output consisted of a set of 1000 hyper-realizations: each hyper-realization consists of a set (1 for each worker) of probability distributions of organ doses. This report describes the hardware components and computational approaches required to make the calculation tractable. Together with the software, this system is referred to here as the 'PANDORA system'. It is based on a commercial SQL server database in a series of six work stations. A complete run of the entire Mayak worker cohort entailed a huge amount of calculations in PANDORA and due to the relatively slow speed of writing the data into the SQL server, each run took about 47 days. Quality control was monitored by comparing doses calculated in PANDORA with those in a specially modified version of the commercial software 'IMBA Professional Plus'. Suggestions are also made for increasing calculation and storage efficiency for future dosimetry calculations using PANDORA.

The Mayak Worker Dosimetry System (MWDS-2013): How to Reduce Hyper-Realisations to Realisations

Two important aspects in which the MWDS-2013 output (absorbed dose to organs calculated in each calendar year) differs from previous data bases (MWDS-2008 and DOSES-2005) are that they have been designed to (a) deal explicitly with uncertainties in model parameters, and (b) differentiate parameters that are considered to be shared (unknown, but having the same value for all workers) and unshared (unknown, but having different values between workers). A multiple-realisation approach is used to preserve information on the effects of shared and unshared parameters both for internal and external doses. Previously, a single realisation (a set of organ doses: one for each worker in the cohort) was calculated using the best estimates of parameter values only. In MWDS-2013, a set of 1000 realisations is produced, to reflect the uncertainty in assumed model parameters: each realisation using a different set of parameter values. Within each realisation, shared parameter values are fixed throughout the cohort, while unshared parameters are allowed to vary between workers. One problem is that because the calculation of organ dose is Bayesian, the estimate for each organ dose is not just a single value, but is itself a distribution (hyper-dose). Technically, it is the probability density of dose given the sampled set of parameter values and given the data for that worker. Thus, in our case, the realisations consist not of single doses, but distributions of doses. The term hyper-realisation is used to differentiate this from the more conventional realisation. Although the multiple hyper-realisation in principle contains all of the necessary information on parameter uncertainty, including shared and unshared parameters, in order to make preliminary epidemiological analyses tractable, and also for consistency with the external doses, it was required to convert the hyper-realisations to realisations. The aim of this paper is to discuss the different approaches that were considered to do this, and to define the method that was eventually chosen. Single spot (point) estimates of dose (for each worker) were also calculated to support the epidemiological analysis. The different methods for obtaining these and the implications are also discussed.

The Mayak Worker Dosimetry System (MWDS-2013): A Bayesian Analysis to Quantify Pulmonary Binding of Plutonium in Lungs Using Historic Beagle Dog Data

The revised human respiratory tract model, published in Part 1 of the International Commission on Radiological Protection's (ICRP) report on Occupational Intakes of Radionuclides (OIR), includes a bound fraction, fb, to represent radionuclides that have become chemically bound in the lungs following dissolution of particulates in lung fluid. Bound radionuclides are not subject to particle transport clearance but can be absorbed to blood at a rate, sb. The occurrence of long-term binding of plutonium can greatly increase lung doses, particularly if it occurs in the bronchial and bronchiolar regions. However, there has been little evidence that currently supports the existence of a long-term bound state for plutonium. The present work describes the analysis of measurements of lung data obtained from a life span study of Beagle dogs that were exposed by inhalation to different concentrations of plutonium-239 (239Pu) nitrate aerosol at Pacific Northwest Laboratories, USA. The data have been analysed to assess whether a bound state was required to explain the data. A Bayesian approach was adopted for the analysis that accounts for uncertainties in model parameter values, including uncertainties in the rates of particle transport clearance. Furthermore, it performs the analysis using two different modelling hypotheses: a model based on the current ICRP human respiratory tract model and its treatment of alveolar particle transport clearance; and a model of particle transport clearance that is based on the updated model developed by ICRP to calculate dose coefficients for the OIR. The current model better represents clearance in dogs at early times (up to 1 year following intake) and the latter better represents retention at greater times (>5 years following intake). The results indicate that a long-term bound fraction of between 0.16 and 1.1%, with a mean value of between 0.24 and 0.8% (depending on the model) is required to explain the data.

THE MAYAK WORKER DOSIMETRY SYSTEM (MWDS-2013) FOR INTERNALLY DEPOSITED PLUTONIUM: AN OVERVIEW

The Mayak Worker Dosimetry System (MWDS-2013) is a system for interpreting measurement data from Mayak workers from both internal and external sources. This paper is concerned with the calculation of annual organ doses for Mayak workers exposed to plutonium aerosols, where the measurement data consists mainly of activity of plutonium in urine samples. The system utilises the latest biokinetic and dosimetric models, and unlike its predecessors, takes explicit account of uncertainties in both the measurement data and model parameters. The aim of this paper is to describe the complete MWDS-2013 system (including model parameter values and their uncertainties) and the methodology used (including all the relevant equations) and the assumptions made. Where necessary, Supplementary papers which justify specific assumptions are cited.

THE MAYAK WORKER DOSIMETRY SYSTEM-2013 (MWDS-2013): PHASE II-QUALITY ASSURANCE OF ORGAN DOSE CALCULATIONS

In order to check developed software tools, it was necessary to compare estimates of statistical characteristics of annual absorbed plutonium internal doses obtained by PANDORA and IMBA software with the same original data. The results were compared from dose calculations of five cases with different initial data on plutonium inhalation intake, lifetime measurements of plutonium activity in daily urine and post-mortem measurements in lungs, lung lymph nodes, liver and skeleton. Estimates of geometric mean and geometric standard deviation of annual regionally weighted lung dose and bone surface dose were compared. Satisfactory agreements of the estimates of statistical characteristics of annual doses to two critical organs for the selected cases were shown. One hundred individual hyper-realizations (forward model evaluations) are sufficient to calculate MWDS-2013 if only measurements of plutonium activity in daily urine are used, and 2000 individual hyper-realizations if both urine and autopsy measurement results are used.

The Mayak Worker Dosimetry System (Mwds-2013): Plutonium Dissolution in The Lungs-An Analysis of Mayak Workers

Lung doses resulting from inhalation of plutonium aerosols are highly dependent on the assumed rate of particle clearance, which occurs by two competing processes: (1) particle transport clearance to the alimentary tract and to the thoracic lymph nodes and (2) clearance to systemic tissues, which occurs by dissolution of particles in lung fluid followed by uptake to blood, which is a process collectively known as absorption. Unbiased and accurate estimates of the values of lung absorption parameters are required to obtain reliable estimates of lung dose, particularly those inferred from urine bioassay. Parameter values governing the rate of absorption are best estimated from data, such as autopsy measurements of plutonium in the lungs and systemic tissues, which directly relate to the exposed workers of interest. However, because the mathematical models that determine clearance from the lungs and systemic tissues are complex and consist of many parameters, estimates of model parameter values are subject to significant uncertainties. With this in mind, this paper uses a Bayesian approach to estimate one of the most important dissolution parameters: the slow rate of dissolution. This is estimated for both plutonium nitrate and plutonium oxide bearing aerosols in the lungs of former workers of the Mayak Production Association. A value of 2.6 × 10-4 d-1 is estimated for plutonium nitrates, and 4.7 × 10-5 d-1 for plutonium oxides.

THE MAYAK WORKER DOSIMETRY SYSTEM (MWDS 2013): A RE-ANALYSIS OF USTUR CASE 0269 TO DETERMINE WHETHER PLUTONIUM BINDS TO THE LUNGS

Radionuclides in ionic form can become chemically bound in the airways of the lungs following dissolution of inhaled particulates in lung fluid. The presence of long-term binding can greatly increase lung doses from inhaled plutonium, particularly if it occurs in the bronchial and bronchiolar regions. However, the only published evidence that plutonium binding occurs in humans comes from an analysis of the autopsy and bioassay data of United States Transuranium and Uranium Registries Case 0269, a plutonium worker who experienced a very high (58 kBq) acute inhalation of plutonium nitrate. This analysis suggested a bound fraction of around 8 %, inferred from an unexpectedly low ratio of estimated total thoracic lymph node activity:total lung activity, at the time of death. However, there are some limitations with this study, the most significant being that measurements of the regional distribution of plutonium activity in the lungs, which provide more direct evidence of binding, were not available when the analysis was performed. The present work describes the analysis of new data, which includes measurements of plutonium activity in the alveolar-interstitial (AI) region, bronchial (BB) and bronchiolar (bb) regions, and extra-thoracic (ET) regions, at the time of death. A Bayesian approach is used that accounts for uncertainties in model parameter values, including particle transport clearance, which were not considered in the original analysis. The results indicate that a long-term bound fraction between 0.4 and 0.7 % is required to explain this data, largely because plutonium activity is present in the extra-thoracic (ET2), bronchial and bronchiolar airways at the time of death.

Assessing the reliability of dose coefficients for exposure to radioiodine by members of the public, accounting for dosimetric and risk model uncertainties

Assessments of risk to a specific population group resulting from internal exposure to a particular radionuclide can be used to assess the reliability of the appropriate International Commission on Radiological Protection (ICRP) dose coefficients used as a radiation protection device for the specified exposure pathway. An estimate of the uncertainty on the associated risk is important for informing judgments on reliability; a derived uncertainty factor, UF, is an estimate of the 95% probable geometric difference between the best risk estimate and the nominal risk and is a useful tool for making this assessment. This paper describes the application of parameter uncertainty analysis to quantify uncertainties resulting from internal exposures to radioiodine by members of the public, specifically 1, 10 and 20-year old females from the population of England and Wales. Best estimates of thyroid cancer incidence risk (lifetime attributable risk) are calculated for ingestion or inhalation of ¹²⁹I and ¹³¹I, accounting for uncertainties in biokinetic model and cancer risk model parameter values. These estimates are compared with the equivalent ICRP derived nominal age-, sex- and population-averaged estimates of excess thyroid cancer incidence to obtain UFs. Derived UF values for ingestion or inhalation of ¹³¹I for 1 year, 10-year and 20-year olds are around 28, 12 and 6, respectively, when compared with ICRP Publication 103 nominal values, and 9, 7 and 14, respectively, when compared with ICRP Publication 60 values. Broadly similar results were obtained for ¹²⁹I. The uncertainties on risk estimates are largely determined by uncertainties on risk model parameters rather than uncertainties on biokinetic model parameters. An examination of the sensitivity of the results to the risk models and populations used in the calculations show variations in the central estimates of risk of a factor of around 2-3. It is assumed that the direct proportionality of excess thyroid cancer risk and dose observed at low to moderate acute doses and incorporated in the risk models also applies to very small doses received at very low dose rates; the uncertainty in this assumption is considerable, but largely unquantifiable. The UF values illustrate the need for an informed approach to the use of ICRP dose and risk coefficients.

A Bayesian analysis of plutonium exposures in Sellafield workers

The joint Russian (Mayak Production Association) and British (Sellafield) plutonium worker epidemiological analysis, undertaken as part of the European Union Framework Programme 7 (FP7) SOLO project, aims to investigate potential associations between cancer incidence and occupational exposures to plutonium using estimates of organ/tissue doses. The dose reconstruction protocol derived for the study makes best use of the most recent biokinetic models derived by the International Commission on Radiological Protection (ICRP) including a recent update to the human respiratory tract model (HRTM). This protocol was used to derive the final point estimates of absorbed doses for the study. Although uncertainties on the dose estimates were not included in the final epidemiological analysis, a separate Bayesian analysis has been performed for each of the 11 808 Sellafield plutonium workers included in the study in order to assess: A. The reliability of the point estimates provided to the epidemiologists and B. The magnitude of the uncertainty on dose estimates. This analysis, which accounts for uncertainties in biokinetic model parameters, intakes and measurement uncertainties, is described in the present paper. The results show that there is excellent agreement between the point estimates of dose and posterior mean values of dose. However, it is also evident that there are significant uncertainties associated with these dose estimates: the geometric range of the 97.5%:2.5% posterior values are a factor of 100 for lung dose, 30 for doses to liver and red bone marrow, and 40 for intakes: these uncertainties are not reflected in estimates of risk when point doses are used to assess them. It is also shown that better estimates of certain key HRTM absorption parameters could significantly reduce the uncertainties on lung dose in future studies.

Parameter uncertainty analysis of a biokinetic model of caesium

Parameter uncertainties for the biokinetic model of caesium (Cs) developed by Leggett et al. were inventoried and evaluated. The methods of parameter uncertainty analysis were used to assess the uncertainties of model predictions with the assumptions of model parameter uncertainties and distributions. Furthermore, the importance of individual model parameters was assessed by means of sensitivity analysis. The calculated uncertainties of model predictions were compared with human data of Cs measured in blood and in the whole body. It was found that propagating the derived uncertainties in model parameter values reproduced the range of bioassay data observed in human subjects at different times after intake. The maximum ranges, expressed as uncertainty factors (UFs) (defined as a square root of ratio between 97.5th and 2.5th percentiles) of blood clearance, whole-body retention and urinary excretion of Cs predicted at earlier time after intake were, respectively: 1.5, 1.0 and 2.5 at the first day; 1.8, 1.1 and 2.4 at Day 10 and 1.8, 2.0 and 1.8 at Day 100; for the late times (1000 d) after intake, the UFs were increased to 43, 24 and 31, respectively. The model parameters of transfer rates between kidneys and blood, muscle and blood and the rate of transfer from kidneys to urinary bladder content are most influential to the blood clearance and to the whole-body retention of Cs. For the urinary excretion, the parameters of transfer rates from urinary bladder content to urine and from kidneys to urinary bladder content impact mostly. The implication and effect on the estimated equivalent and effective doses of the larger uncertainty of 43 in whole-body retention in the later time, say, after Day 500 will be explored in a successive work in the framework of EURADOS.

An intake prior for the Bayesian analysis of plutonium and uranium exposures in an epidemiology study

In Bayesian inference, the initial knowledge regarding the value of a parameter, before additional data are considered, is represented as a prior probability distribution. This paper describes the derivation of a prior distribution of intake that was used for the Bayesian analysis of plutonium and uranium worker doses in a recent epidemiology study. The chosen distribution is log-normal with a geometric standard deviation of 6 and a median value that is derived for each worker based on the duration of the work history and the number of reported acute intakes. The median value is a function of the work history and a constant related to activity in air concentration, M, which is derived separately for uranium and plutonium. The value of M is based primarily on measurements of plutonium and uranium in air derived from historical personal air sampler (PAS) data. However, there is significant uncertainty on the value of M that results from paucity of PAS data and from extrapolating these measurements to actual intakes. This paper compares posterior and prior distributions of intake and investigates the sensitivity of the Bayesian analyses to the assumed value of M. It is found that varying M by a factor of 10 results in a much smaller factor of 2 variation in mean intake and lung dose for both plutonium and uranium. It is concluded that if a log-normal distribution is considered to adequately represent worker intakes, then the Bayesian posterior distribution of dose is relatively insensitive to the value assumed of M.

An assessment of the reliability of dose coefficients for intakes of radionuclides by members of the public

This paper summarises work undertaken on behalf of the Environment Agency for England to quantify uncertainties resulting from internal exposures to a number of radionuclides considered significant because of their anthropogenic origin, namely: (238)U, (226)Ra, (239)Pu, (241)Am, (137)Cs, (90)Sr, (131)I, (129)I and (3)H. Uncertainties in the biokinetic models that are used to calculate the retention and excretion of radionuclides are derived in order to calculate distributions of effective dose per unit intake following their inhalation or ingestion by members of the UK public. The central values and ranges of the distributions are used to inform the derivation of uncertainty factors (UFs) for the different dose coefficients, which can be used to assess reliability. These represent uncertainties inherent in the structures of the biokinetic models and their parameter values. The inferred UF values are typically around 2-3 for ingestion and 2-6 for inhalation for all age groups, and are comparable to UF values inferred from published studies. It is instructive to consider these ranges alongside the likely levels of exposure that are expected from the radionuclides considered (the microsievert range) and the dose limit of planned exposures for members of the public (1000 μSv).

Assessing the reliability of dose coefficients for ingestion and inhalation of 226Ra and 90Sr by members of the public

Assessments of risk to a population group resulting from internal exposure to a particular radionuclide can be used to assess the reliability of the appropriate International Commission on Radiological Protection (ICRP) dose coefficient, E(50), used as a radiation protection device for the specified exposure pathway. An estimate of the uncertainty on the risk is important for informing judgements on reliability. This paper describes the application of parameter uncertainty analysis to quantify uncertainties resulting from internal exposures to radioisotopes of the alkaline earth metals, (90)Sr and (226)Ra, by members of the UK public. The study derives uncertainties in biokinetic model parameter values to calculate the distributions of the effective dose per unit intake using the ICRP Publication 60 formalism. The distributions are used to infer the uncertainty on the mean effective dose per unit intake to inform the derivation of uncertainty factors (UF) for the appropriate ICRP Publication 72 dose coefficients. Here, a UF indicates a 95 % probability that the best estimate of risk per unit intake is within a factor, UF, of the nominal risk associated with the appropriate ICRP dose coefficient, E(50), with respect to uncertainties in the biokinetic model parameter values. Ingestion: it is assumed that exposure occurs through the ingestion of radionuclides present in food and water. The results for both radionuclides suggest a UF of within 3 for all age groups, with median values close to the ICRP values. Inhalation: it is assumed that environmental exposure to radium occurs primarily due to insoluble forms present in fly ash discharged from coal-fired power stations; for strontium, exposure is assumed to occur due to residual aerosols produced as a result of atmospheric nuclear testing and nuclear reactor accidents. The results suggest a UF of around 3 and 6 for inhalation of (90)Sr and (226)Ra, respectively, by members of the public.

The reliability of dose coefficients for inhalation and ingestion of uranium by members of the public

The best estimate of risk to a population group resulting from internal exposure to a particular radionuclide can be used to assess the reliability of the appropriate International Commission on Radiological Protection (ICRP) dose coefficient (E⁵⁰) for the specified exposure pathway. An estimate of the uncertainty on the risk is important for reliability decisions. This paper describes the application of parameter uncertainty analysis to quantify uncertainties resulting from internal exposures to uranium (as (²³⁸U) by members of the public. The study derives uncertainties in biokinetic model parameter values to calculate the distributions of the effective dose per unit intake using the ICRP Publication 60 formalism. The central values and ranges of the distributions are used to infer the uncertainty on the mean effective dose per unit intake to inform the derivation of uncertainty factors (UF) for the dose coefficients. Here, a UF is a conditional probability statement that the value of the best estimate of risk per unit intake has a 95 % probability of being within a factor, UF, of the nominal risk associated with the appropriate ICRP dose coefficient, E⁵⁰, with respect to uncertainties in the biokinetic model parameter values. Ingestion: it is assumed that exposure occurs through the ingestion of uranium present in food and water. The results suggest a UF of within 3 for all age groups, with median values close to the ICRP values. Inhalation: it is assumed that environmental exposure to uranium occurs via inhalation of a mixture of chemical forms. The results suggest a UF of around 2 for inhalation of uranium by members of the public, with median values close to the ICRP values.

A Bayesian analysis of uncertainties on lung doses resulting from occupational exposures to uranium

In a recent epidemiological study, Bayesian estimates of lung doses were calculated in order to determine a possible association between lung dose and lung cancer incidence resulting from occupational exposures to uranium. These calculations, which produce probability distributions of doses, used the human respiratory tract model (HRTM) published by the International Commission on Radiological Protection (ICRP) with a revised particle transport clearance model. In addition to the Bayesian analyses, point estimates (PEs) of doses were also provided for that study using the existing HRTM as it is described in ICRP Publication 66. The PEs are to be used in a preliminary analysis of risk. To explain the differences between the PEs and Bayesian analysis, in this paper the methodology was applied to former UK nuclear workers who constituted a subset of the study cohort. The resulting probability distributions of lung doses calculated using the Bayesian methodology were compared with the PEs obtained for each worker. Mean posterior lung doses were on average 8-fold higher than PEs and the uncertainties on doses varied over a wide range, being greater than two orders of magnitude for some lung tissues. It is shown that it is the prior distributions of the parameters describing absorption from the lungs to blood that are responsible for the large difference between posterior mean doses and PEs. Furthermore, it is the large prior uncertainties on these parameters that are mainly responsible for the large uncertainties on lung doses. It is concluded that accurate determination of the chemical form of inhaled uranium, as well as the absorption parameter values for these materials, is important for obtaining unbiased estimates of lung doses from occupational exposures to uranium for epidemiological studies. Finally, it should be noted that the inferences regarding the PEs described here apply only to the assessments of cases provided for the epidemiological study, where central estimates of dose were sought. Approved dosimetry service assessments of exposures are unlikely to yield significant underestimates, as pessimistic assumptions of lung solubility would almost always be used.

A method for calculating Bayesian uncertainties on internal doses resulting from complex occupational exposures

Estimating uncertainties on doses from bioassay data is of interest in epidemiology studies that estimate cancer risk from occupational exposures to radionuclides. Bayesian methods provide a logical framework to calculate these uncertainties. However, occupational exposures often consist of many intakes, and this can make the Bayesian calculation computationally intractable. This paper describes a novel strategy for increasing the computational speed of the calculation by simplifying the intake pattern to a single composite intake, termed as complex intake regime (CIR). In order to assess whether this approximation is accurate and fast enough for practical purposes, the method is implemented by the Weighted Likelihood Monte Carlo Sampling (WeLMoS) method and evaluated by comparing its performance with a Markov Chain Monte Carlo (MCMC) method. The MCMC method gives the full solution (all intakes are independent), but is very computationally intensive to apply routinely. Posterior distributions of model parameter values, intakes and doses are calculated for a representative sample of plutonium workers from the United Kingdom Atomic Energy cohort using the WeLMoS method with the CIR and the MCMC method. The distributions are in good agreement: posterior means and Q(0.025) and Q(0.975) quantiles are typically within 20 %. Furthermore, the WeLMoS method using the CIR converges quickly: a typical case history takes around 10-20 min on a fast workstation, whereas the MCMC method took around 12-72 hr. The advantages and disadvantages of the method are discussed.

Uncertainty analysis of doses from ingestion of plutonium and americium

Uncertainty analyses have been performed on the biokinetic model for americium currently used by the International Commission on Radiological Protection (ICRP), and the model for plutonium recently derived by Leggett, considering acute intakes by ingestion by adult members of the public. The analyses calculated distributions of doses per unit intake. Those parameters having the greatest impact on prospective doses were identified by sensitivity analysis; the most important were the fraction absorbed from the alimentary tract, f(1), and rates of uptake from blood to bone surfaces. Probability distributions were selected based on the observed distribution of plutonium and americium in human subjects where possible; the distributions for f(1) reflected uncertainty on the average value of this parameter for non-specified plutonium and americium compounds ingested by adult members of the public. The calculated distributions of effective doses for ingested (239)Pu and (241)Am were well described by log-normal distributions, with doses varying by around a factor of 3 above and below the central values; the distributions contain the current ICRP Publication 67 dose coefficients for ingestion of (239)Pu and (241)Am by adult members of the public. Uncertainty on f(1) values had the greatest impact on doses, particularly effective dose. It is concluded that: (1) more precise data on f(1) values would have a greater effect in reducing uncertainties on doses from ingested (239)Pu and (241)Am, than reducing uncertainty on other model parameter values and (2) the results support the dose coefficients (Sv Bq(-1) intake) derived by ICRP for ingestion of (239)Pu and (241)Am by adult members of the public.

EURADOS coordinated action on research, quality assurance and training of internal dose assessments

EURADOS working group on 'Internal Dosimetry (WG7)' represents a frame to develop activities in the field of internal exposures as coordinated actions on quality assurance (QA), research and training. The main tasks to carry out are the update of the IDEAS Guidelines as a reference document for the internal dosimetry community, the implementation and QA of new ICRP biokinetic models, the assessment of uncertainties related to internal dosimetry models and their application, the development of physiology-based models for biokinetics of radionuclides, stable isotope studies, biokinetic modelling of diethylene triamine pentaacetic acid decorporation therapy and Monte-Carlo applications to in vivo assessment of intakes. The working group is entirely supported by EURADOS; links are established with institutions such as IAEA, US Transuranium and Uranium Registries (USA) and CEA (France) for joint collaboration actions.

Plutonium worker dosimetry

Epidemiological studies of the relationship between risk and internal exposure to plutonium are clearly reliant on the dose estimates used. The International Commission on Radiological Protection (ICRP) is currently reviewing the latest scientific information available on biokinetic models and dosimetry, and it is likely that a number of changes to the existing models will be recommended. The effect of certain changes, particularly to the ICRP model of the respiratory tract, has been investigated for inhaled forms of (239)Pu and uncertainties have also been assessed. Notable effects of possible changes to respiratory tract model assumptions are (1) a reduction in the absorbed dose to target cells in the airways, if changes under consideration are made to the slow clearing fraction and (2) a doubling of absorbed dose to the alveolar region for insoluble forms, if evidence of longer retention times is taken into account. An important factor influencing doses for moderately soluble forms of (239)Pu is the extent of binding of dissolved plutonium to lung tissues and assumptions regarding the extent of binding in the airways. Uncertainty analyses have been performed with prior distributions chosen for application in epidemiological studies. The resulting distributions for dose per unit intake were lognormal with geometric standard deviations of 2.3 and 2.6 for nitrates and oxides, respectively. The wide ranges were due largely to consideration of results for a range of experimental data for the solubility of different forms of nitrate and oxides. The medians of these distributions were a factor of three times higher than calculated using current default ICRP parameter values. For nitrates, this was due to the assumption of a bound fraction, and for oxides due mainly to the assumption of slower alveolar clearance. This study highlights areas where more research is needed to reduce biokinetic uncertainties, including more accurate determination of particle transport rates and long-term dissolution for plutonium compounds, a re-evaluation of long-term binding of dissolved plutonium, and further consideration of modeling for plutonium absorbed to blood from the lungs.

A Monte Carlo method for calculating Bayesian uncertainties in internal dosimetry

This paper presents a novel Monte Carlo method (WeLMoS, Weighted Likelihood Monte-Carlo sampling method) that has been developed to perform Bayesian analyses of monitoring data. The WeLMoS method randomly samples parameters from continuous prior probability distributions and then weights each vector by its likelihood (i.e. its goodness of fit to the measurement data). Furthermore, in order to quality assure the method, and assess its strengths and weaknesses, a second method (MCMC, Markov chain Monte Carlo) has also been developed. The MCMC method uses the Metropolis algorithm to sample directly from the posterior distribution of parameters. The methods are evaluated and compared using an artificially generated case involving an exposure to a plutonium nitrate aerosol. In addition to calculating the uncertainty on internal dose, the methods can also calculate the probability distribution of model parameter values given the observed data. In other words, the techniques provide a powerful tool to obtain the estimates of parameter values that best fit the data and the associated uncertainty on these estimates. Current applications of the methodology, including the determination of lung solubility parameters, from volunteer and cohort data, are also discussed.

Uncertainty analysis of doses from inhalation of depleted uranium

Measurements of uranium excreted in urine have been widely used to monitor possible exposures to depleted uranium (DU). This paper describes a comprehensive probabilistic uncertainty analysis of doses determined retrospectively from measurements of DU in urine. Parametric uncertainties in the International Commission on Radiological Protection (ICRP) Human Respiratory Tract Model (HRTM) and ICRP systemic model for uranium were considered in the analysis, together with uncertainties in an alternative model for particle removal from the lungs. Probability distributions were assigned to HRTM parameters based on uncertainties documented in ICRP Publication 66 and elsewhere, including the Capstone study of aerosols produced after DU penetrator impacts. Uncertainties in the uranium systemic model were restricted to transfer rates having the greatest effect on urinary excretion, and hence retrospective dose assessments, over the measurement times considered (10-10(4) d). The overall uncertainty on dose (the ratio of the upper and lower quantiles, q0.975/q0.025) was estimated to be about a factor of 50 at 10 days after intake and about a factor of 10 at 10(3)-10(4) d. The dose to the lung dominated the committed effective dose, with the lung absorption parameters, particularly the slow dissolution rate, ss, dominating the overall uncertainty. The median dose determined from a measurement of 1 ng DU, collected in urine in a 24-h period, varied from 0.1 microSv at 10 d to about 1 mSv at 10(4) d. Despite the large uncertainties, the upper q0.975 quantile for the assessed dose was below 1 mSv up to 5,000 d.

Internal dosimetry: towards harmonisation and coordination of research

The CONRAD Project is a Coordinated Network for Radiation Dosimetry funded by the European Commission 6th Framework Programme. The activities developed within CONRAD Work Package 5 ('Coordination of Research on Internal Dosimetry') have contributed to improve the harmonisation and reliability in the assessment of internal doses. The tasks carried out included a study of uncertainties and the refinement of the IDEAS Guidelines associated with the evaluation of doses after intakes of radionuclides. The implementation and quality assurance of new biokinetic models for dose assessment and the first attempt to develop a generic dosimetric model for DTPA therapy are important WP5 achievements. Applications of voxel phantoms and Monte Carlo simulations for the assessment of intakes from in vivo measurements were also considered. A Nuclear Emergency Monitoring Network (EUREMON) has been established for the interpretation of monitoring data after accidental or deliberate releases of radionuclides. Finally, WP5 group has worked on the update of the existing IDEAS bibliographic, internal contamination and case evaluation databases. A summary of CONRAD WP5 objectives and results is presented here.

Internal dose assessments: uncertainty studies and update of ideas guidelines and databases within CONRAD project

The work of Task Group 5.1 (uncertainty studies and revision of IDEAS guidelines) and Task Group 5.5 (update of IDEAS databases) of the CONRAD project is described. Scattering factor (SF) values (i.e. measurement uncertainties) have been calculated for different radionuclides and types of monitoring data using real data contained in the IDEAS Internal Contamination Database. Based upon this work and other published values, default SF values are suggested. Uncertainty studies have been carried out using both a Bayesian approach as well as a frequentist (classical) approach. The IDEAS guidelines have been revised in areas relating to the evaluation of an effective AMAD, guidance is given on evaluating wound cases with the NCRP wound model and suggestions made on the number and type of measurements required for dose assessment.

Evaluation of scattering factor values for internal dose assessment following the IDEAS guidelines: preliminary results

The IDEAS Guidelines for the assessment of internal doses from monitoring data suggest default measurement uncertainties (i.e. scattering factors, SFs) to be used for different types of monitoring data. However, these default values were mainly based upon expert judgement. In this paper, SF values have been calculated for different radionuclides and types of monitoring data using real data contained in the IDEAS Internal Contamination Database. Results are presented.

Avoiding biased estimates of dose when nothing is known about the time of intake

A common problem in internal dosimetry occurs in routine monitoring, when it is required to estimate an intake from a measurement made at the end of a monitoring interval, and the time of intake is unknown. ICRP suggests that it should be assumed that the intake occurred in the middle of the monitoring period. However, it has been shown that this will, in the long-term, lead to biased estimates of a worker's intake and dose. In order to overcome this biasing, the United States Department of Energy (USDOE) recommends a different method based on calculating the intakes for all possible intake times in the interval and then taking an arithmetic average. In a recent paper, it has been shown that both the ICRP and USDOE methods were biased and that the only unbiased estimator of the true intake was obtained by assuming a constant chronic intake throughout the monitoring interval. In all of the analyses carried out to date on this 'Constant Chronic' method, it was assumed that the measurements were exact. In this paper, the effects of assuming either normally or log-normally distributed measurement errors are explored, and the effect on the bias of the intake estimate is investigated.

Estimating uncertainty on internal dose assessments

The estimation of uncertainty on doses broadly falls into three categories. (1) Estimating the uncertainty on prospective doses. Here, the intake is known and the uncertainties in individual parameter values must be propagated through the calculated dose. (2) Estimating the error or uncertainty on dose assessments made from single measurements. Here, intake, model parameter and measurement uncertainties are propagated into the measurement, but default ICRP parameter values are used to estimate the intake and dose from the measurement. (3)Estimating the probability distribution of an individual's dose from a set of monitoring data. Here, Bayesian inference methods must be used to estimate the uncertainty on the estimated dose. A computer code is being developed that performs all three types of uncertainty analysis using Monte Carlo simulation. The software samples biokinetic parameters from probability density functions and then calculates doses from these parameters by calling the dosimetry code IMBA Professional Plus. A description of the methodology, together with an example application of the software, is included in this paper.

Coordination of research on internal dosimetry in Europe: the CONRAD project

The EUropean RAdiation DOSimetry Group (EURADOS) initiated in 2005 the CONRAD Project, a Coordinated Network for Radiation Dosimetry funded by the European Commission (EC), within the 6th Framework Programme (FP). The main purpose of CONRAD is to generate a European Network in the field of Radiation Dosimetry and to promote both research activities and dissemination of knowledge. The objective of CONRAD Work Package 5 (WP5) is the coordination of research on assessment and evaluation of internal exposures. Nineteen institutes from 14 countries participate in this action. Some of the activities to be developed are continuations of former European projects supported by the EC in the 5th FP (OMINEX and IDEAS). Other tasks are linked with ICRP activities, and there are new actions never considered before. A collaboration is established with CONRAD Work Package 4, dealing with Computational Dosimetry, to organise an intercomparison on Monte Carlo modelling for in vivo measurements of (241)Am deposited in a knee phantom. Preliminary results associated with CONRAD WP5 tasks are presented here.

IMBA Professional Plus: a flexible approach to internal dosimetry

IMBA (Integrated Modules for Bioassay Analysis) is a suite of software modules that implement the current ICRP biokinetic and dosimetric models for estimation of intakes and doses. The IMBA modules have gone through extensive quality assurance, and are now used for routine formal dose assessment by Approved Dosimetry Services throughout the UK. HPA has continued to develop the IMBA modules. In addition, several projects, sponsored by organisations both in the USA and in Canada, have resulted in the development of customised user-friendly interfaces (IMBA Expert 'editions'). These enable users not only to use the standard ICRP models, but also to change many of the parameter values from ICRP defaults, and to apply sophisticated data handling techniques to internal dose calculations. These include: fitting measurement data with the maximum likelihood method; using multiple chronic and acute intakes; and dealing with different data types, such as urine, faces and whole body simultaneously. These interfaces were improved further as a result of user-feedback, and a general 'off-the-shelf' product, IMBA Professional, was developed and made available in January 2004. A new version, IMBA Professional Plus, was released in April 2005, which is both faster and more powerful than previous software. The aim of this paper is to describe the capabilities of IMBA Professional Plus, and the mathematical methods used.

IDEAS internal contamination database: a compilation of published internal contamination cases. A tool for the internal dosimetry community

In the scope of the IDEAS project to develop General Guidelines for the Assessment of Internal Dose from Monitoring data, two databases were compiled. The IDEAS Bibliography database contains references dealing with problems related to cases of internal contamination. The IDEAS Internal Contamination Database now contains more than 200 cases of internal contamination. In the near future, the IDEAS Internal Contamination database will be made available to the internal dosimetry community. The database has several potential applications, including: training, testing biokinetic models, testing software for calculating intakes and doses from bioassay data, comparison of data from a new accidental intake with that from previous exposures to similar materials. The database is by no means complete, and this presentation is also an appeal for internal contamination cases to extend and update it.

The autocorrelation coefficient as a tool for assessing goodness of fit between bioassay predictions and measurement data

Project IDEAS has produced guidelines for internal dose assessment. An integral part of this process is assessing the goodness of fit of biokinetic models to bioassay data. It is recommended that a fit should only be accepted if (a) it is close enough to the data not to be rejected by a chi2 test and (b) if it looks acceptable to 'the eye'. The latter criterion was added to enable the assessor to reject fits which seemed to display some sort of systematic bias. However, there are problems with both of these tests: (a) the chi2 test is dependent on the assumed uncertainties which are often unknown, (b) 'by eye' assessment is subjective. In this paper, another statistic, the autocorrelation coefficient of the residuals, rho, is investigated. The main advantages of the rho statistic are that it is objective, very sensitive to biasing and independent of the assumed errors.

Obtaining an unbiased estimate of intake in routine monitoring when the time of intake is unknown

A common problem in internal dosimetry occurs in routine monitoring, when it is required to estimate an intake from a measurement made at the end of a monitoring interval, and the time of intake is unknown. ICRP suggests that, in these cases, it should be assumed that the intake occurred in the middle of the monitoring period. However, it has been shown that this will, in the long term, lead to biased estimates of a worker's intake and dose. In order to overcome this biasing, the United States Department of Energy (USDOE) recommends a different method based on calculating the intakes for all possible intake-times in the interval, and then taking an arithmetic average. In this paper, it is shown that both the ICRP and USDOE methods are biased. An alternative method is suggested, which assumes a constant chronic intake throughout the monitoring interval. Monte Carlo simulations are used to estimate the magnitude of bias for two realistic monitoring programmes using all three methods. It is shown that the proposed method is unbiased and also yields estimates of intake that are generally closer to the actual intake, than the other two. The Monte Carlo conclusions are backed up by a theoretical analysis of bias. Finally, the source of bias in the apparently intuitive approach of the USDOE method is revealed by viewing the problem from a Bayesian perspective.

Uncertainties in doses from intakes of radionuclides assessed from monitoring measurements

The evaluation of uncertainties in doses from intakes of radionuclides is one of the most difficult problems in internal dosimetry. In this paper, the process of assessing internal doses from monitoring measurements is reviewed and the major sources of uncertainty are discussed. Methods developed independently at HPA and at IRSN for the determination of uncertainties in internal doses assessed from monitoring measurements are described. Both use a Monte Carlo simulation approach. Results are described for three illustrative examples. An alternative method developed at the Los Alamos National Laboratory that uses Bayesian statistical methods is also described briefly.

IMBA Expert: internal dosimetry made simple

In 1997, a collaboration between British Nuclear Fuels plc (BNFL), Westlakes Research Institute and NRPB started, with the aim of producing IMBA (Integrated Modules for Bioassay Analysis), a suite of software modules that implement the new ICRP models for estimation of intakes and doses. This was partly in response to new UK regulations, and partly due to the requirement for a unified approach in estimating intakes and doses from bioassay measurements within the UK. Over the past 5 years, the IMBA modules have been developed further, have gone through extensive quality assurance, and are now used for routine dose assessment by approved dosimetry services throughout the UK. More recently, interest in the IMBA methodology has been shown by the United States Department of Energy (USDOE), and in 2001 an ambitious project to develop a software package (IMBA Expert USDOE Edition) which would meet the requirements of all of the major USDOE sites began. Interest in IMBA Expert is now being expressed in many other countries. The aim of this paper is to outline the origin and evolution of the IMBA modules (the past); to describe the full capabilities of the current IMBA Expert system (the present) and to indicate possible future directions in terms of capabilities and availability (the future).