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International audience; This article aims at comparing reference methods for the assessment of cancer risk from exposure to genotoxic carcinogen chemical substances and to ionizing radiation. For chemicals, cancer potency is expressed as a toxicological reference value (TRV) based on the most sensitive type of cancer generally observed in animal experiment of oral or inhalation exposure. A dose-response curve is established by modelling experimental data adjusted to apply to human exposure. This leads to a point of departure from which the TRV is derived as the slope of a linear extrapolation to zero dose. Human lifetime cancer risk can then be assessed as the product of dose by TRV and it is generally considered to be acceptable below 10-5 for public in normal situation. Radiation exposure is assessed as effective dose corresponding to a weighted average of energy deposition in body organs. Cancer risk models were derived from the epidemiological follow-up of atomic bombing survivors. Considering a linear-no-threshold dose-risk relationship and average baseline risks, lifetime nominal risk coefficients were established for 13 types of cancers. Those are adjusted according to the severity of each cancer type and combined into an overall indicator denominated radiation detriment. Exposure to radiation is subject to dose limits proscribing unacceptable health detriment. The differences between chemical and radiological cancer risk assessments are discussed and concern data sources, extrapolation to low doses, definition of dose, considered health effects and level of conservatism. Despite these differences, opportunities exist to bring closer the two types of risk assessment.

International audience; Procedures using iodine-131 represent more than 90% of all therapies in nuclear medicine in Algeria. It is important to evaluate the long-term biological effects of iodine treatment on non-target organs to improve patient radiation protection. This experimental radiotoxicology study aims to determine the biokinetic models of iodine contamination. For this purpose, two Wistar rat models, with and without a thyroid, have been used to evaluate the biological half-life of iodine and then to perform a biodistribution study of iodine activity in 15 organs and tissues. For the most relevant organs, the respective absorbed doses have been calculated using RODES software. © 2019 Society for Radiological Protection. Published on behalf of SRP by IOP Publishing Limited. All rights reserved.

International audience; Nowadays, protection standards against ionizing radiation health effects are mainly derived from the results of the epidemiological study of atomic bomb survivors in Hiroshima and Nagasaki, who were exposed externally to radiation, at high dose rate. However, exposure to ionizing radiation in the general population and in workers generally occurs as a cumulation of low doses over years (therefore, at low dose rates), and as a result of external, but also internal, exposure. Epidemiological studies in radiation workers allow studying the long-term health effects of such exposures, and are precious to assess the adequacy of current radiation protection standards against cancer. In addition, they also allow for the study of new priority research questions in radiation protection, such as potential non-cancer effects of low dose radiation. Finally, they also provide a general picture of health in populations of radiation workers, which is useful for occupational and public health expertise. This article provides an overview of epidemiological research conducted in workers of the nuclear cycle by the French Institute for Radiological Protection and Nuclear Safety (IRSN), and discusses perspectives for its future developments. © EDP Sciences 2018; À ce jour, les normes internationales de radioprotection sont essentiellement basées sur les connaissances des effets sanitaires des rayonnements ionisants issues des études de cohortes de survivants des bombardements atomiques d’Hiroshima et de Nagasaki, qui ont subi des irradiations par voie externe à forts débits de doses. Or, les expositions aux rayonnements ionisants survenant en population générale et chez certains travailleurs correspondent généralement à de faibles doses, cumulées de manière prolongée dans le temps. Ces expositions peuvent avoir lieu par irradiation externe mais aussi par contamination interne. Les études épidémiologiques chez les travailleurs du nucléaire permettent notamment de mieux caractériser les effets de ces conditions d’expositions. Elles permettent ainsi d’évaluer l’adéquation des normes de radioprotection pour protéger les travailleurs, mais également la population générale adulte, contre les risques de cancer. Par ailleurs, elles permettent d’aborder de nouveaux questionnements de recherche en radioprotection, comme les potentiels effets non cancéreux. Enfin, ces études fournissent un bilan de santé général des populations de travailleurs du nucléaire. Cet article propose une vue d’ensemble des recherches épidémiologiques menées par l’Institut de Radioprotection et de Sûreté Nucléaire (IRSN) chez les travailleurs du cycle électronucléaire en France. Il dresse également quelques perspectives de développements possibles de ces recherches.

– The 2007 Recommendations of the International Commission on Radiological Protection (ICRP, 2007) introduced changes that affect the calculation of effective dose, and implied a revision of the dose coefficients for internal exposure, published previously in the Publication 30 series (ICRP, 1979, 1980, 1981, 1988) and Publication 68 (ICRP, 1994). In addition, new data are now available that support an update of the radionuclide-specific information given in Publications 54 and 78 (ICRP, 1988a, 1997b) for the design of monitoring programmes and retrospective assessment of occupational internal doses. Provision of new biokinetic models, dose coefficients, monitoring methods, and bioassay data was performed by Committee 2, Task Group 21 on Internal Dosimetry, and Task Group 4 on Dose Calculations. A new series, the Occupational Intakes of Radionuclides (OIR) series, will replace the Publication 30 series and Publications 54, 68, and 78. OIR Part 1 has been issued (ICRP, 2015), and describes the assessment of internal occupational exposure to radionuclides, biokinetic and dosimetric models, methods of individual and workplace monitoring, and general aspects of retrospective dose assessment. OIR Part 2 (ICRP, 2016), this current publication and upcoming publications in the OIR series (Parts 4 and 5) provide data on individual elements and their radioisotopes, including information on chemical forms encountered in the workplace; a list of principal radioisotopes and their physical half-lives and decay modes; the parameter values of the reference biokinetic model; and data on monitoring techniques for the radioisotopes encountered most commonly in workplaces. Reviews of data on inhalation, ingestion, and systemic biokinetics are also provided for most of the elements. Dosimetric data provided in the printed publications of the OIR series include tables of committed effective dose per intake (Sv Bq−1 intake) for inhalation and ingestion, tables of committed effective dose per content (Sv Bq−1 measurement) for inhalation, and graphs of retention and excretion data per Bq intake for inhalation. These data are provided for all absorption types and for the most common isotope(s) of each element. The electronic annex that accompanies the OIR series of publications contains a comprehensive set of committed effective and equivalent dose coefficients, committed effective dose per content functions, and reference bioassay functions. Data are provided for inhalation, ingestion, and direct input to blood. This third publication in the series provides the above data for the following elements: ruthenium (Ru), antimony (Sb), tellurium (Te), iodine (I), caesium (Cs), barium (Ba), iridium (Ir), lead (Pb), bismuth (Bi), polonium (Po), radon (Rn), radium (Ra), thorium (Th), and uranium (U).

ObjectivesThere is growing evidence of an association between low-dose external γ-radiation and circulatory system diseases (CSDs), yet sparse data exist about an association with chronic internal uranium exposure and the role of non-radiation risk factors. We conducted a nested case–control study of French AREVA NC Pierrelatte nuclear workers employed between 1960 and 2005 to estimate CSD risks adjusting for major CSD risk factors (smoking, blood pressure, body mass index, total cholesterol and glycaemia) and external γ-radiation dose.MethodsThe study included 102 cases of death from CSD and 416 controls individually matched on age, gender, birth cohort and socio-professional status. Information on CSD risk factors was collected from occupational medical records. Organ-specific absorbed doses were estimated using biomonitoring data, taking into account exposure regime and uranium physicochemical properties. External γ-radiation was measured by individual dosimeter badges. Analysis was conducted with conditional logistic regression.ResultsWorkers were exposed to very low radiation doses (mean γ-radiation dose 2 and lung uranium dose 1 mGy). A positive but imprecise association was observed (excess OR per mGy 0.2, 95% CI 0.004 to 0.5). Results obtained after adjustment suggest that uranium exposure might be an independent CSD risk factor.ConclusionsOur results suggest that a positive association might exist between internal uranium exposure and CSD mortality, not confounded by CSD risk factors. Future work should focus on numerous uncertainties associated with internal uranium dose estimation and on understanding biological pathway of CSD after protracted low-dose internal radiation exposure.

The 2007 Recommendations (ICRP, 2007) introduced changes that affect the calculation of effective dose, and implied a revision of the dose coefficients for internal exposure, published previously in the Publication 30 series (ICRP, 1979a,b, 1980a, 1981, 1988) and Publication 68 (ICRP, 1994b). In addition, new data are now available that support an update of the radionuclide-specific information given in Publications 54 and 78 (ICRP, 1989a, 1997) for the design of monitoring programmes and retrospective assessment of occupational internal doses. Provision of new biokinetic models, dose coefficients, monitoring methods, and bioassay data was performed by Committee 2 and its task groups. A new series, the Occupational Intakes of Radionuclides (OIR) series, will replace the Publication 30 series and Publications 54, 68, and 78. OIR Part 1 (ICRP, 2015) describes the assessment of internal occupational exposure to radionuclides, biokinetic and dosimetric models, methods of individual and workplace monitoring, and general aspects of retrospective dose assessment. OIR Part 2 (ICRP, 2016), OIR Part 3 (ICRP, 2017), this current publication, and the final publication in the OIR series (OIR Part 5) provide data on individual elements and their radioisotopes, including information on chemical forms encountered in the workplace; a list of principal radioisotopes and their physical half-lives and decay modes; the parameter values of the reference biokinetic models; and data on monitoring techniques for the radioisotopes most commonly encountered in workplaces. Reviews of data on inhalation, ingestion, and systemic biokinetics are also provided for most of the elements. Dosimetric data provided in the printed publications of the OIR series include tables of committed effective dose per intake (Sv per Bq intake) for inhalation and ingestion, tables of committed effective dose per content (Sv per Bq measurement) for inhalation, and graphs of retention and excretion data per Bq intake for inhalation. These data are provided for all absorption types and for the most common isotope(s) of each element. The online electronic files that accompany the OIR series of publications contains a comprehensive set of committed effective and equivalent dose coefficients, committed effective dose per content functions, and reference bioassay functions. Data are provided for inhalation, ingestion, and direct input to blood. This fourth publication in the OIR series provides the above data for the following elements: lanthanum (La), cerium (Ce), praseodymium (Pr), neodymium (Nd), promethium (Pm), samarium (Sm), europium (Eu), gadolinium (Gd), terbium (Tb), dysprosium (Dy), holmium (Ho), erbium (Er), thulium (Tm), ytterbium (Yb), lutetium (Lu), actinium (Ac), protactinium (Pa), neptunium (Np), plutonium (Pu), americium (Am), curium (Cm), berkelium (Bk), californium (Cf), einsteinium (Es), and fermium (Fm).

International audience; The biokinetic model for systemic americium (Am) currently recommended by the International Commission on Radiological Protection (ICRP) for application to occupational intake of Am is based on information available through the early 1990s. Much additional information on Am biokinetics has been developed in the past 25 y, including measurements of retention and excretion of 241Am in many workers with 241Am burdens and post mortem measurements of 241Am in tissues of some of those workers. The ICRP's current Am model is reasonably consistent with the updated information, with the main exception that the current model apparently overestimates 24-hour urinary Am as a fraction of skeletal or systemic Am at late times after intake. This paper provides an overview of current information on the systemic kinetics of Am in adult human subjects and laboratory animals and presents an updated biokinetic model for systemic Am that addresses the discrepancies between the current database and current ICRP systemic model for Am. This model is applied in Part 4 (to appear) of an ICRP series of reports on intake of radionuclides by workers called the OIR (Occupational Intake of Radionuclides) series. © 2019 Not subject to copyright in the USA. Contribution of Oak Ridge National Laboratory.

This document provides a comprehensive overview study on the physico-chemical speciation of radioiodine observed in the atmosphere after various emissions related to nuclear activities: nuclear weapon tests, accident and incident releases, and routine discharges. The study covers different types of nuclear facilities including medical isotope production facilities (MIPFs), reprocessing plants (RPs), and nuclear power plants (NPPs). Most attention is paid to 131I which has a major human health impact in the early stages of a nuclear emergency situation with regard to inhalation. Iodine-131 combines a high yield by neutron-induced nuclear fission of 235U (2.87%) or 239Pu (3.8%), high dose coefficients, and a radioactive half-life long enough to allow for spreading at global scales and entering the food chain but sufficiently short to produce a significant dose commitment when inhaled or ingested. Reliable dose assessment requires both detailed and valid information on the physico-chemical composition of 131I present in the air. Apart from reactor explosions and fires, which produce large amounts of particles and may therefore favor the presence of iodine in particulate form at short distance, other nuclear accident scenarios will lead fairly rapidly to a dominant gaseous radioiodine proportion in the atmosphere.

International audience; European Radiation Dosimetry Group (EURADOS) Working Group 7 is a network on internal dosimetry that brings together researchers from more than 60 institutions in 21 countries. The work of the group is organised into task groups that focus on different aspects, such as development and implementation of biokinetic models (e.g. for diethylenetriamine penta-acetic acid decorporation therapy), individual monitoring and the dose assessment process, Monte Carlo simulations for internal dosimetry, uncertainties in internal dosimetry, and internal microdosimetry. Several intercomparison exercises and training courses have been organised. The IDEAS guidelines, which describe – based on the International Commission on Radiological Protection’s (ICRP) biokinetic models and dose coefficients – a structured approach to the assessment of internal doses from monitoring data, are maintained and updated by the group. In addition, Technical Recommendations for Monitoring Individuals for Occupational Intakes of Radionuclides have been elaborated on behalf of the European Commission, DG-ENER (TECHREC Project, 2014–2016, coordinated by EURADOS). Quality assurance of the ICRP biokinetic models by calculation of retention and excretion functions for different scenarios has been performed and feedback was provided to ICRP. An uncertainty study of the recent caesium biokinetic model quantified the overall uncertainties, and identified the sensitive parameters of the model. A report with guidance on the application of ICRP biokinetic models and dose coefficients is being drafted at present. These and other examples of the group’s activities, which complement the work of ICRP, are presented. © 2018, The International Society for Prosthetics and Orthotics.

International audience; In the early phase of a nuclear reactor accident, in-vivo monitoring of impacted population would be highly useful to detect potential contamination during the passage of the cloud and to estimate the dose from inhalation of measured radionuclides. However, it would be important to take into account other exposure components (1) inhalation of unmeasured radionuclides and (2) external irradiation from the plume and from the radionuclides deposited on the soil. This article presents a methodology to calculate coefficients used to convert in-vivo measurement results directly into doses, not only from the measured radionuclides but from all sources of exposure according to model-based projected doses. This early interpretation of in-vivo measurements will provide an initial indication of individual exposure levels. As an illustration, the methodology is applied to two scenarios of accidents affecting a nuclear power plant a loss-of-coolant accident leading to core meltdown and a steam generator tube rupture accident. © The Author(s) 2019. Published by Oxford University Press. All rights reserved. For Permissions, please email journals.permissions@oup.com.

International audience; PURPOSE: The aim is to investigate associations between mortality and exposure to ionizing radiation in a cohort of uranium workers with potential for internal and external radiation exposures.METHODS: Workers employed for at least 6 months between 1958 and 2006 in five plants involved in the French nuclear fuel cycle were included and followed up between 1968 and 2013. Cause-specific standardized mortality ratios were calculated. Analyses of associations between individual cumulative radiation dose (both internal and external, lagged by 5-15 years) and mortality were conducted using Poisson regression.RESULTS: The cohort includes 4541 workers. The mean cumulative external dose was 11.12 mGy. Mean cumulative internal doses ranged, depending on modelling hypotheses, from 0.05 to 0.09 mGy (liver) and from 4.22 to 10.90 mGy (lung). At the end of the follow-up, 838 workers were deceased and 28 lost to follow-up. A healthy worker effect was observed. The risk of prostate and lung cancers mortality was significantly higher for workers exposed to cumulative external dose above 50 mGy compared to non-exposed, but these associations were based only on three cases and became non-significant, although of similar magnitude, after adjustment for smoking. Associations with internal dose showed no consistent pattern.CONCLUSIONS: For the first time, a study was conducted in a French cohort of uranium workers with a complete reconstruction of internal dose. Results are preliminary and must be interpreted with caution because of the limited cohort size and significant sources of uncertainty. Future steps of this study will overcome these limitations.

International audience; H Miloudi, M Locatelli, G Autret, D Balvay, A Desbrée, E Blanchardon, J M Bertho application of RODES software to experimental biokinetic data for dose assessment in mice and rats. In support of experimental studies of chronic, long-term contamination in rodents, voxel-based computer models were built representing adult mice and juvenile, adult and elderly rats of both sexes. RODES software was created to calculate absorbed radiation doses to organs with these specific anatomical models. Absorbed doses were then calculated starting from previously published biokinetic data. Whole body doses showed less than 5% differences between calculation with RODES and calculation with the ICRP Publication 108 model for long term exposure to 90Sr of mice. Similar results were obtained for long term exposure to 137Cs. Dose distribution for 90Sr internal contamination also showed that the dose to the skeleton is six fold more as compared to the whole body dose while radiation dose to other organs is less than the mean whole body dose. These results underline the importance of using specific anatomical models according to the age and the sex of experimental animals. © 2017 IOP Publishing Ltd.

Supplemental Digital Content is available in the text., Background: Carcinogenic risks of internal exposures to alpha-emitters (except radon) are poorly understood. Since exposure to alpha particles—particularly through inhalation—occurs in a range of settings, understanding consequent risks is a public health priority. We aimed to quantify dose–response relationships between lung dose from alpha-emitters and lung cancer in nuclear workers. Methods: We conducted a case–control study, nested within Belgian, French, and UK cohorts of uranium and plutonium workers. Cases were workers who died from lung cancer; one to three controls were matched to each. Lung doses from alpha-emitters were assessed using bioassay data. We estimated excess odds ratio (OR) of lung cancer per gray (Gy) of lung dose. Results: The study comprised 553 cases and 1,333 controls. Median positive total alpha lung dose was 2.42 mGy (mean: 8.13 mGy; maximum: 316 mGy); for plutonium the median was 1.27 mGy and for uranium 2.17 mGy. Excess OR/Gy (90% confidence interval)—adjusted for external radiation, socioeconomic status, and smoking—was 11 (2.6, 24) for total alpha dose, 50 (17, 106) for plutonium, and 5.3 (−1.9, 18) for uranium. Conclusions: We found strong evidence for associations between low doses from alpha-emitters and lung cancer risk. The excess OR/Gy was greater for plutonium than uranium, though confidence intervals overlap. Risk estimates were similar to those estimated previously in plutonium workers, and in uranium miners exposed to radon and its progeny. Expressed as risk/equivalent dose in sieverts (Sv), our estimates are somewhat larger than but consistent with those for atomic bomb survivors. See video abstract at, http://links.lww.com/EDE/B232.

International audience; In order to support animal experiments of chronic radionuclides intake with realistic dosimetry, voxel-based three-dimensional computer models of mice and rats of both sexes and three ages were built from magnetic resonance imaging. Radiation transport of mono-energetic photons of 11 energies and electrons of 7 energies was simulated with MCNPX 2.6c to assess specific absorbed fractions (SAFs) of energy emitted from 13 source regions and absorbed in 28 target regions. RODES software was developed to combine SAF with radiation emission spectra and user-supplied biokinetic data to calculate organ absorbed doses per nuclear transformation of radionuclides in source regions (S-factors) and for specific animal experiments with radionuclides. This article presents the design of RODES software including the simulation of the particles in the created rodent voxel phantoms. SAF and S-factor values were compared favourably with published results from similar studies. The results are discussed for rodents of different ages and sexes. © 2017 IOP Publishing Ltd.

International audience; Purpose: Technetium-99 m (Tc-99 m)-labelled microparticles, functionalized with fucoidan to present a high affinity for P-Selectin, or [ 99m Tc] MP-fucoidan, were developed as a novel SPECT radiotracer for abdominal aortic aneurysm (AAA). As a prerequisite step forwards a clinical trial, the biodistribution and dosimetry of these [ 99m Tc] MP-fucoidan microparticles were performed in rats in order to estimate the absorbed and effective dose in humans.Procedures: Microparticles with a maximum hydrodynamic diameter of 4 μm were obtained by crosslinking polysaccharides dextran and pullulan. They were functionalized with fucoidan then radiolabelled with Tc-99 m. A mean labelling efficiency of 92±1 % was measured. [ 99m Tc] MP-fucoidan (43±2 MBq) was injected to 24 rats via the penis vein. Rats were euthanized at 30, 60, 120 and 240 min after injection (4 rats at each time point). Samples of each organ, as well as the injected microparticle suspensions, were aliquoted for counting. Four animals were sacrificed for blood clearance studies and four were sacrificed for image analysis and quantification of the cortical, medullary, papillary kidney, and pelvis uptake. A compartmental model was realised using SAAM II and organ data were fitted. The area under the curve was then used to compute the residence times in each rat organs and converted to human residence time values. Absorbed and effective human doses in organs were estimated using (1) the OLINDA/EXM 1.1 software with the hermaphroditic mathematical phantoms and (2) the OEDIPE software associated to the MCNPX Monte Carlo code and the ICRP reference computational male and female phantoms, using the updated tissue weighting factors in the ICRP Publication 103.Results: The highest human residence times were found in the liver, kidneys, and urinary bladder wall. The largest doses were found in the kidneys and then in the urinary bladder wall and liver. The human effective doses were 6.06 μSv/MBq for the hermaphroditic mathematical phantom and 5.95 μSv/MBq for the ICRP adult reference computational phantom.Conclusions: Animal-based human dose estimates support a future first-inhuman testing of [ 99m Tc] MP-fucoidan following IV injection.

Purpose: To collect values of parameters describing the physico-chemical properties of different uranium compounds in order to quantify their variability and to propose specific parameters for different workplaces and compounds.Material and methods: The published values of absorption parameters, gastrointestinal absorption fractions, activity median aerodynamic diameters and geometric standard deviations of the particle size activity distribution were collected.Results: Average and median values for each chemical form and workplace were determined for these parameters. These values can be used when no precise information is known for dose assessment following internal contamination by uranium. This review presents and discusses the variability of these parameters for the different uranium chemical forms and workplaces. Finally, sensitivity of the dose coefficients to these parameters was quantified.Conclusion: Specific parameter values for different workplaces and compounds were proposed and the v...

Purpose: To improve the dosimetry of incorporated americium (Am) and to contribute to radiation protection by characterizing the absorption kinetics of inhaled Am compounds.Material and methods: In vitro dissolution tests, animal experiments and human contamination cases published in the literature were reviewed. The data were analyzed with biokinetic models consistent with the current publications of the International Commission on Radiological Protection.Results: Material-specific dissolution parameter values with consequent assignment to absorption Types are proposed as well as representative central values for the different chemical forms of Am.Conclusions: The absorption of Am oxide is consistent with the moderate absorption Type M while Am nitrate appears more soluble. Am associated with plutonium oxide usually follows its slow absorption Type S. However, the large variability observed stresses the value of investigating the specific absorption kinetics for Am compounds which represent a sig...

In the event of an accident involving radioactive material, there could potentially be a large number of people requiring internal contamination monitoring. To answer to this challenge, the IRSN, the French institute for radiological protection and nuclear safety, has developed since 2007 a fleet of 10 body counting mobile units: 8 light and heavy emergency units and 2 expertise mobile units, unique in Europe, allowing on-site monitoring of up to 2500 people per day. Their use for the follow up of French nationals coming back from Japan after the Fukushima Accident is described and discussed.

The fourth workshop of the Multidisciplinary European Low Dose Initiative (MELODI) was organised by STUK—Radiation and Nuclear Safety Authority of Finland. It took place from 12 to 14 September 2012 in Helsinki, Finland. The meeting was attended by 179 scientists and professionals engaged in radiation research and radiation protection. We summarise the major scientific findings of the workshop and the recommendations for updating the MELODI Strategic Research Agenda and Road Map for future low dose research activities.

L’uranium est naturellement present dans l’alimentation, a un niveau tres variable, conduisant a des activites detectables dans les excreta, independamment d’une eventuelle exposition professionnelle. L’interpretation des resultats de surveillance radiotoxicologique individuelle des travailleurs exposes professionnellement au dioxyde d’uranium doit donc prendre en compte l’apport alimentaire pour etablir une estimation dosimetrique. Pour ce faire, les auteurs ont teste differentes approches dans le suivi d’un travailleur et proposent une methode originale d’interpretation des donnees radiotoxicologiques par comparaison des resultats obtenus dans un groupe de travailleurs exposes a ceux collectes dans une population de reference. Cette methode a permis d’estimer une dose efficace engagee annuelle pour un groupe de travailleurs exposes au dioxyde d’uranium.

Recent epidemiology studies highlighted the detrimental health effects of exposure to low dose and low dose rate ionizing radiation (IR): nuclear industry workers studies have shown increased leukaemia and solid tumour risks following cumulative doses of

International audience; The assessment and management of risks associated with exposures to ionising radiation are defined by the general radiological protection system, proposed by the International Commission on Radiological Protection (ICRP). This system is regarded by a large majority of users as a robust system although there are a number of dissenting voices, claiming that it is not suitable for estimating the risks resulting from internal exposures. One of the specific issues of internal exposure involves short-range radiations such as Auger and beta particles. Auger- and beta-emitting radionuclides can be distributed preferentially in certain tissue structures and even in certain cellular organelles, according to their chemical nature and the vector with which they are associated. Given the limited range of the low-energy electrons in biological matter, this heterogeneous distribution can generate highly localised energy depositions and exacerbate radiotoxic responses at cellular level. These particularities in energy distribution and cellular responses are not taken into account by the conventional methods for the assessment of risk. Alternative systems have been proposed, based on dosimetry conducted at the cellular or even molecular level, whose purpose is to determine the energy deposition occurring within the DNA molecule. However, calculation of absorbed doses at the molecular level is not sufficient to ensure a better assessment of the risks incurred. Favouring such a microdosimetric approach for the risk assessments would require a comprehensive knowledge of the biological targets of radiation, the dose-response relationships at the various levels of organisation, and the mechanisms leading from cellular energy deposition to the appearance of a health detriment. The required knowledge is not fully available today and it is not yet possible to link an intracellular energy deposition to a probability of occurrence of health effects or to use methods based on cellular dosimetry directly. The imperfections of the alternative approaches proposed so far should not discourage efforts. Protection against exposure to Auger and low-energy beta emitters would benefit from mechanistic studies, dedicated to the study of energy depositions of the radionuclides in various cellular structures, but also from radiotoxicological studies to define the relative biological effectiveness of the various Auger emitters used in medicine and of certain low-energy beta emitters, whose behaviour may depend greatly on their chemical form during intake. The scientific expertise, as well as the human and physical resources needed to conduct these studies, is available. They could be now mobilised into international low-dose research programmes, in order to ultimately improve the protection of people exposed to these specific radionuclides. © 2013 IOP Publishing Ltd.

International audience; Dose assessment after intakes of radionuclides requires application of biokinetic and dosimetric models and assumptions about factors influencing the final result. In 2006, a document giving guidance for such assessment was published, commonly referred to as the IDEAS Guidelines. Following its publication, a working group within the European networks CONRAD and EURADOS was established to improve and update the IDEAS Guidelines. This work resulted in Version 2 of the IDEAS Guidelines, which was published in 2013 in the form of a EURADOS report. The general structure of the original document was maintained; however, new procedures were included, e.g. the direct dose assessment method for 3H or special procedure for wound cases applying the NCRP wound model. In addition, information was updated and expanded, e.g. data on dietary excretion of U, Th, Ra and Po for urine and faeces or typical and achievable values for detection limits for different bioassay measurement techniques. © The Author 2015. Published by Oxford University Press. All rights reserved.

International audience; The TECHREC project, funded by the European Commission, will provide Technical Recommendations for Monitoring Individuals for Occupational Intakes of Radionuclides. It is expected that the document will be published by the European Commission as a report in its Radiation Protection Series during 2016. The project is coordinated by the European Radiation Dosimetry Group (EURADOS) and is being carried out by members of EURADOS Working Group 7 (Internal Dosimetry). This paper describes the aims and purpose of the Technical Recommendations, and explains how the project is organised. © Crown copyright 2015.

The potential health impacts of chronic exposures to uranium, as they occur in occupational settings, are not well characterized. Most epidemiological studies have been limited by small sample sizes, and a lack of harmonization of methods used to quantify radiation doses resulting from uranium exposure. Experimental studies have shown that uranium has biological effects, but their implications for human health are not clear. New studies that would combine the strengths of large, well-designed epidemiological datasets with those of state-of-the-art biological methods would help improve the characterization of the biological and health effects of occupational uranium exposure. The aim of the European Commission concerted action CURE (Concerted Uranium Research in Europe) was to develop protocols for such a future collaborative research project, in which dosimetry, epidemiology and biology would be integrated to better characterize the effects of occupational uranium exposure. These protocols were developed from existing European cohorts of workers exposed to uranium together with expertise in epidemiology, biology and dosimetry of CURE partner institutions. The preparatory work of CURE should allow a large scale collaborative project to be launched, in order to better characterize the effects of uranium exposure and more generally of alpha particles and low doses of ionizing radiation. © 2016 IOP Publishing Ltd.

International audience; In case of internal contamination with plutonium materials, a treatment with diethylene triamine pentaacetic acid (DTPA) can be administered in order to reduce plutonium body burden and consequently avoid some radiation dose. DTPA intravenous injections or inhalation can start almost immediately after intake, in parallel with urinary and fecal bioassay sampling for dosimetric follow-up. However, urine and feces excretion will be significantly enhanced by the DTPA treatment. As internal dose is calculated from bioassay results, the DTPA effect on excretion has to be taken into account. A common method to correct bioassay data is to divide it by a factor representing the excretion enhancement under DTPA treatment by intravenous injection. Its value may be based on a nominal reference or observed after a break in the treatment. The aim of this study was to estimate the influence of this factor on internal dose by comparing the dose estimated using default or upper and lower values of the enhancement factor for 11 contamination cases. The observed upper and lower values of the enhancement factor were 18.7 and 63.0 for plutonium and 24.9 and 28.8 for americium. For americium, a default factor of 25 is proposed. This work demonstrates that the use of a default DTPA enhancement factor allows the determination of the magnitude of the contamination because dose estimated could vary by a factor of 2 depending on the value of the individual DTPA enhancement factor. In case of significant intake, an individual enhancement factor should be determined to obtain a more reliable dose assessment. © 2016 Health Physics Society.

International audience; The knowledge of the absorbed radiation dose is essential in order to interpret the results of animal experiments with chronic ingestion of radionuclides by rodents. In order to evaluate this absorbed dose, we applied the dose conversion factors proposed by ICRP publication 108 to mouse chronic ingestion experiments with 20 kBq.L-1 137Cs- or 90Sr-contaminated water. The results indicated that whole-body absorbed doses were 9 mGy and 10 mGy by the end of 20 weeks of 137Cs or 90Sr ingestion, respectively. These results of dose calculations were compared with results obtained with a more refined method using published organ-specific absorbed fractions of energy. There was good agreement between the two methods, indicating that, despite the simple hypotheses used to apply the ICRP 108 dose conversion factors to our mouse model, this method allows one to calculate in a simple and reliable way the absorbed radiation dose received by the rodents during long-term experiments on chronic ingestion of radionuclides. © 2012 EDP Sciences.

A workshop dedicated to cancer risks associated with low-dose internal contamination was organised in March 2011, in Paris, in the framework of the DoReMi (Low Dose Research towards Multidisciplinary Integration) European Network of Excellence. The aim was to identify the best epidemiological studies that provide an opportunity to develop a multidisciplinary approach to improve the evaluation of the cancer risk associated with internal contamination. This workshop provided an opportunity for in-depth discussions between researchers working in different fields including (but not limited to) epidemiology, dosimetry, biology and toxicology. Discussions confirmed the importance of research on the health effects of internal contamination. Several existing epidemiological studies provide a real possibility to improve the quantification of cancer risk associated with internal emitters. Areas for future multidisciplinary collaborations were identified, that should allow feasibility studies to be carried out in the near future. The goal of this paper is to present an overview of the presentations and discussions that took place during this workshop.

En vue d’optimiser la protection des travailleurs vis-a-vis des rayonnements ionisants, la reglementation francaise impose des limites de dose et une demarche de reduction progressive de l’exposition, dans la continuite des recommandations de la Commission internationale de protection radiologique (CIPR). Afin de verifier le respect des limites et contraintes de dose lorsqu’un risque significatif de contamination interne existe, des programmes de surveillance sont mis en place, comprenant des mesures radiotoxicologiques periodiques. Cependant, des incertitudes dans l’interpretation dosimetrique de ces mesures sont introduites par leurs variabilites propres et par la connaissance incomplete des circonstances de l’exposition. Ces incertitudes ont ete prises en compte par des techniques statistiques bayesiennes. La methodologie developpee a ete appliquee a l’analyse du programme de surveillance des travailleurs des ateliers de purification du plutonium du site AREVA NC de La Hague. A partir du seuil de decision du comptage nucleaire, une dose minimale detectable (DMD) par ce programme de routine avec un niveau de confiance donne peut etre calculee a l’aide du logiciel « Optimisation des programmes de surveillance de la contamination interne » (OPSCI). Cette methodologie s’avere un support utile a l’optimisation des programmes de surveillance systematique par la recherche du meilleur compromis entre leur sensibilite et leur cout

Iodine-131 (131I) is one of the most frequently used radionuclides for diagnosis and therapy of thyroid diseases. It is administered orally in the treatment of cancer to eliminate the residual postoperative microscopic tumor foci, and the residual normal thyroid tissue for early detection of recurrence [1]. The comparative behavior of 131I concentration into two animalmodels with total and partial thyroid has been investigated in our previous work [2]. The accumulated activities have been measured in fourteen organs. In this study, the mean absorbed doses resulting from 131I accumulated in all organs have been evaluated using RODES software [3, 4]. With this software, mean absorbed doses were calculatedfor selected organs (thyroid, lungs, heart, liver, kidneys, stomach, spleen, large and small intestine, testes, urinary bladder wall) by combining the specific absorbed fractions (SAF) of energy with radiation emission spectra and biokinetic data determined from our previous experimental study [2]. Calculations were based on the 131I photon and electron emissions reported by [5] and SAFs previously calculated by Monte Carlo simulation in the voxel phantom of an adult male rate [3, 4]. The obtained results show high absorbed dosesdeliveredto stomach and lungs for both models compared to other organs. The dose received by the testes and salivary glands is found to be higher in the case of the rat model without thyroid. Conversely, the spleen and bladder wall received lower doses in this latter model compared to those received by the rat model with thyroid. One can also note that the difference in mean absorbed dose received by liver, lungs, heart, and walls of the stomach is not significant between the two rat models.

Epidemiological studies of the occupational exposure of miners and domestic exposures of the public have provided strong and complementary evidence of the risks of lung cancer following inhalation of radon progeny. Recent miner epidemiological studies, which include low levels of exposure, long duration of follow-up, and good quality of individual exposure data, suggest higher risks of lung cancer per unit exposure than assumed previously by the International Commission on Radiological Protection (ICRP). Although risks can be managed by controlling exposures, dose estimates are required for the control of occupational exposures and are also useful for comparing sources of public exposure. Currently, ICRP calculates doses from radon and its progeny using dose conversion factors from exposure (WLM) to dose (mSv) based on miner epidemiological studies, referred to as the epidemiological approach. Revision of these dose conversion factors using risk estimates based on the most recent epidemiological data gives values that are in good agreement with the results of calculations using ICRP biokinetic and dosimetric models, the dosimetric approach. ICRP now proposes to treat radon progeny in the same way as other radionuclides and to publish dose coefficients calculated using models, for use within the ICRP system of protection.

cited By 3; Lung cancer risk following radon inhalation is strongly established. On the contrary, much less is known about the possible health effects of radon ingestion via drinking water. This article summarizes the current knowledge of dosimetry following radon ingestion. Results from epidemiological studies of radon in drinking water are also presented and discussed. © 2010 EDP Sciences.; Si le risque de cancer du poumon lié à l’inhalation de radon est aujourd’hui largement reconnu, les possibles conséquences sanitaires de l’ingestion d’eau contenant du radon apparaissent moins bien documentées. Cet article synthétise les connaissances actuelles sur la dosimétrie correspondant à l’ingestion de radon. Les résultats des études épidémiologiques sur les effets sanitaires du radon dans l’eau de boisson sont également présentés et discutés.

Les travailleurs du cycle nucleaire, notamment ceux impliques dans la preparation ou le retraitement du combustible nucleaire, peuvent etre exposes a une contamination interne par des composes radioactifs (uranium, plutonium…). Afin de se premunir contre un risque sanitaire du a cette exposition, la reglementation francaise definit des limites d’exposition annuelles a ne pas depasser. Ces limites sont exprimees en termes de dose efficace engagee resultant de la distribution des radionucleides dans le corps humain et de l’absorption des rayonnements ionisants qu’ils emettent par les tissus radiosensibles. Pour s’assurer du non-depassement de ces limites et pour identifier tout evenement anormal, la potentielle contamination interne des travailleurs doit etre surveillee. Cette surveillance est realisee par des mesures de la radioactivite retenue dans le corps ou presente dans les excreta. A partir du scenario d’exposition, le resultat de ces mesures peut etre interprete en termes de dose efficace engagee. Cependant, une incertitude dans l’interpretation dosimetrique des resultats est introduite par la variabilite de la mesure et par la connaissance incomplete des conditions de contamination. Des methodes statistiques ont ete developpees pour prendre en compte cette incertitude dans l’optimisation des programmes de surveillance individuelle, l’objectif etant de garantir le respect des objectifs dosimetriques de la radioprotection, avec le niveau de confiance souhaite et des moyens operationnels raisonnables. Ces methodes ont ete implementees dans le logiciel OPSCI developpe en collaboration entre l’IRSN et AREVA. Ce logiciel permet, par la prise en compte des incertitudes : – d’estimer la dose minimale detectable par un programme de surveillance de routine a partir des informations disponibles sur la forme physico-chimique des composes manipules, des niveaux de radioactivite en presence et des limites de detection des techniques disponibles pour la mesure des radionucleides incorpores ; – de calculer la dose efficace engagee suite a un incident de contamination a partir des resultats de mesure d’activite retenue et/ou excretee, ainsi que l’incertitude associee ; – d’aider le medecin du travail a choisir le programme de surveillance le mieux adapte aux conditions d’exposition potentielle ou averee. Des cas concrets d’utilisation de ce logiciel seront presentes afin de montrer son interet pratique. Cet outil apporte ainsi un vrai support au medecin du travail en lui donnant les elements necessaires pour repondre aux questions sur la representativite et les limites des protocoles de surveillance en situation normale ou anormale et sur la quantification des expositions internes a la radioactivite.

The individual monitoring of internal exposure of workers comprises two steps: measurement and measurement interpretation. The latter consists in reconstructing the intake of a radionuclide from the activity measurement and calculating the dose using a biokinetic model of the radionuclide behavior in the human body. Mathematically, reconstructing the intake is solving an inverse problem described by a measurement-model equation. The aim of this paper is to propose a solution to this inverse problem when the measurement-model parameters are considered as uncertain. For that, an analysis of the uncertainty on the intake calculation is performed taking into account the dispersion of the measured quantity and the uncertainties of the measurement-model parameters. It is shown that both frequentist and Bayesian approaches can be used to solve the problem according to the measurement-model formulation. A common calculation algorithm is proposed to support both approaches and applied to the examples of tritiated water intake and plutonium inhalation by a worker. © 2009 Health Physics Society.

Modelling Pu/Am decorporation by DTPA. A new tool has been developed to optimize DTPA efficacy as concerns reduction of effective dose after 239 Pu wound. For example, the simulations show, for moderately soluble compounds (type M), a 1/3 decrease of effective dose is obtained after repeated early treatment (24 i.v. for 4 months), whereas a decrease by a factor 5 can be reached if treatments continue for 5 years at 2 week interval. By contrast, for poorly soluble compounds (type S), negligible efficacy is observed after early treatments, and a 3 time decrease of dose is obtained for treatments performed at 2 week interval for 50 years. Some of the hypotheses retained for modelling DTPA decorporation are validated from new experimental data published recently, and structure of a new model which can be applied both to Pu and Am is reported, taking into account urinary and faecal excretion, structure being suitable for different doses of DTPA and using various galenic forms.

Although great efforts have been made to improve the physical anthropomorphic phantoms used to calibrate in vivo measurement systems, each of these phantoms represents a single average counting geometry and usually contains a uniform distribution of a radionuclide in the tissue substitute. As a matter of fact, significant corrections must be made to phantom-based calibration factors in order to obtain absolute calibration efficiencies applicable to a given individual. The OEDIPE software has been developed at the Institute for Radiological Protection and Nuclear Safety. It allows the direct and fast construction of a voxel phantom from medical images with a realistic distribution of activity between organs and its conversion into computer files to be used online for Monte Carlo calculations. OEDIPE allows determining and visualizing the variation of activity retention in the segmented organs of the phantom and running MCNPX with these calculated source-organs at different times after intake. The influence of the biokinetics of radionuclides on the in vivo measurement as well as the uncertainty on the estimated activity is quantified by comparing the numerical calibration coefficients obtained by considering realistic and dynamic biokinetic distributions of activity with the counting efficiency commonly obtained from simple, homogeneous, and static activity distributions in organs or in the whole body. As an application, the effect of long-term retention tissues was studied through a case of inhalation of an insoluble compound where the thoracic lymph nodes make a significant contribution to the lung counting efficiency. Future work will concern contamination through other routes of intake, such as a wound, and mixtures of radionuclides.

To improve calibration methods of in vivo counting, our laboratory has developed a computer tool to model internal contamination and assess in vivo activity and corresponding organ absorbed doses. The aim of the recent work was to define a more realistic source based on biokinetic models. The influence of the biokinetic parameters on the in vivo counting was studied through the simulation of an acute inhalation intake of Am. The tissue distribution of activity predicted by the biokinetic model was visualized. Two equivalent methods for determination of the efficiency related to the total activity distributed in the body were used. The comparison between the efficiency taking the biokinetics into account and the classically estimated efficiency quantifies the influence of the activity distribution in the body and provides conversion factors for correcting the classical efficiency to account for biokinetics. ©2009Health Physics Society.

La conversion en dose d’une exposition au radon fait l’objet de debats. L’UNSCEAR a defini un coefficient de conversion de la concentration en radon vers la dose efficace a partir de modeles dosimetriques et d’informations epidemiologiques. La CIPR recommande d’evaluer le risque lie au radon a partir des etudes epidemiologiques. Pour permettre la prise en compte d’une exposition combinee a d’autres sources de rayonnements ionisants, elle a adopte des conventions de conversion en dose efficace inferieures de 10 a 30 % au coefficient de l’UNSCEAR, sur la base du suivi epidemiologique des mineurs d’uranium. Ainsi, des niveaux de reference differents pour la concentration en radon peuvent etre deduits d’un meme niveau de dose efficace suivant que l’on applique l’un ou l’autre facteur de conversion. Cependant, il convient de rappeler que ces differences sont negligeables par rapport aux incertitudes associees aux etudes tant dosimetriques qu’epidemiologiques et que les niveaux de reference ne sont que des outils pour aider a reduire aussi bas que raisonnablement possible l’exposition au radon et le risque qui en resulte.

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.

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.

Recent developments in the software package OEDIPE - enabling the simulation of in vivo counting - to use the biokinetic data in the modelling of the internal contamination. As part of the improvement of calibration techniques of in vivo counting, the Labo- ratory of Internal Dose Assessment of IRSN has developed a computer tool, "OEDIPE", to model internal contamination, to simulate in vivo counting and to cal- culate internal dose. The previous version of this software enabled to easily model sources made of a single radionuclide and located in a single organ but could not repre- sent a mixture of radionuclides which distribution evolves from the time of intake according to the biokinetics of the radionuclides. The purpose of this article is to des- cribe a new facility developed in the software to allow the modelling of sources heterogeneous in composition and in distribution and optionally to automatically inte- grate the distribution of the activity in different tissues estimated by biokinetic calculation at any time since the intake. These new developments give the opportunity to study much more complex contamination cases and the influence of the biokinetics on the in vivo counting, leading to a better assessment of the calibration coefficients and of the corresponding uncertainties.

Introduction Data are sparse about the potential health risks of chronic low-dose contamination of humans by uranium (natural or anthropogenic) in drinking water. Previous studies report some molecular imbalances but no clinical signs due to uranium intake. Objectives In a proof-of-principle study, we reported that metabolomics is an appropriate method for addressing this chronic low-dose exposure in a rat model (uranium dose: 40 mg L−1; duration: 9 months, n = 10). In the present study, our aim was to investigate the dose–effect pattern and identify additional potential biomarkers in urine samples. Methods Compared to our previous protocol, we doubled the number of rats per group (n = 20), added additional sampling time points (3 and 6 months) and included several lower doses of natural uranium (doses used: 40, 1.5, 0.15 and 0.015 mg L−1). LC–MS metabolomics was performed on urine samples and statistical analyses were made with SIMCA-P+ and R packages. Results The data confirmed our previous results and showed that discrimination was both dose and time related. Uranium exposure was revealed in rats contaminated for 9 months at a dose as low as 0.15 mg L−1. Eleven features, including the confidently identified N1-methylnicotinamide, N1-methyl-2-pyridone-5-carboxamide and 4-hydroxyphenylacetylglycine, discriminated control from contaminated rats with a specificity and a sensitivity ranging from 83 to 96 %, when combined into a composite score. Conclusion These findings show promise for the elucidation of underlying radiotoxicologic mechanisms and the design of a diagnostic test to assess exposure in urine, in a dose range experimentally estimated to be above a threshold between 0.015 and 0.15 mg L−1. Electronic supplementary material The online version of this article (doi:10.1007/s11306-016-1092-8) contains supplementary material, which is available to authorized users.

L’estimation des doses absorbee, equivalente et efficace resultant d’une contamination interne se base sur la mesure de l’activite par anthroporadiametrie (in vivo ) ou par analyse radiotoxicologique in vitro. La valeur d’activite mesuree est interpretee en termes de dose a l’aide de modeles biocinetiques et dosimetriques qui decrivent le devenir des radionucleides dans le corps d’un individu de reference et permettent de calculer l’energie qu’ils y deposent. Pour s’assurer du respect des limites reglementaires de dose, des programmes de surveillance sont mis en place. Les mesures d’activite peuvent etre interpretees suivant une procedure standard comme celle proposee par le groupe IDEAS. Cependant, une incertitude significative est introduite dans le calcul de la dose par les incertitudes de mesure, par la meconnaissance partielle des conditions d’exposition et par le realisme imparfait des modeles. En plus de cette incertitude, l’extrapolation aux faibles doses d’effets biologiques observes apres une irradiation plus intense, la toxicite chimique des radionucleides et l’heterogeneite du depot d’energie rendent difficile l’estimation du risque resultant d’une contamination interne. En conclusion, il convient de garder a l’esprit que la dose efficace est un outil de gestion de la radioprotection mais pas un indicateur quantitatif du risque individuel.

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. © The Author 2007. Published by Oxford University Press. All rights reserved.