Occupational Intakes Of Radionuclides Research Articles

Application of INTDOSKIT tool for assessment of uncertainties on dose coefficients for ingestion of uranium by workers

The International Commission on Radiological Protection (ICRP) has developed the Human Alimentary Tract Model (HATM) to calculate radiation doses from the ingestion of radionuclides for the protection of workers and the public. In parallel, the ICRP's Occupational Intake of Radionuclides (OIR) series provides biokinetic models and dose coefficients based on a reference human, primarily for regulatory purposes. Although these coefficients are not usually checked for uncertainties, the investigation of such uncertainties is crucial to ensure their reliability in radiation protection. This study uses INTDOSKIT, a software tool developed using the R programming language and RStudio as the Integrated Development Environment (IDE), to calculate doses and explore uncertainties following ingestion of U-238 by workers. Two scenarios were investigated: Case I, using the latest HATM model and the systemic uranium model from ICRP publications 100 and 137 respectively, and Case II, based on data from literature, using older ICRP models. In both cases, intake of U-238 in its soluble form (Type F) was modeled and the results were validated using the ICRP OIR dataviewer software. Following validation, uncertainty and sensitivity analyses were performed. In Case I, uncertainties were assigned to the particle transport parameters in both the systemic model model and HATM as well as the uranium uptake fraction into the blood. While in Case II they were limited to the uptake fraction (f1) and the systemic uranium model. A Monte Carlo simulation of 60,000 runs was performed for both cases, sampling model parameters from their respective probability distributions to generate dose distributions. The influence of each parameter on these distributions was also analyzed. Probability distributions were inferred to the calculated dose values using the maximum likelihood estimation method and the Kolmogorov-Smirnov goodness-of-fit. The results showed that for both cases the committed effective dose coefficient, e (50), followed a lognormal distribution. Case I had a geometric mean of 3.2E-08 Sv/Bq and GSD of 2.0, while case II had a slightly lower geometric mean of 3.1E-08 Sv/Bq and GSD of 1.9. Sensitivity analysis showed that the main contributor to dose uncertainty was the fraction of uranium absorbed from the small intestine into the blood. Both cases showed similar trends, with slightly higher results in case I. Overall, this study demonstrates the effectiveness of INTDOSKIT in calculating dose coefficients and analyzing uncertainties. It suggests that while the ICRP reference values remain useful for protection, the incorporation of additional statistical measures and distribution characteristics could further enhance radiation protection strategies.

Improving the reliability of internal dosimetry for uranium workers: Application of the ICRP/OIR uranium model to long-term occupational intakes

In the Occupational Intakes of Radionuclides (OIR) report series, the International Commission on Radiological Protection (ICRP) published updated biokinetic models and dosimetric data associated with the internal exposure of workers, which are consistent with the recommendations of ICRP Publication 103. The present study focused on the application of the new OIR uranium model published in ICRP Publication 137, which is relevant for the interpretation of measurements of activity in the body and in excreta, and for calculating the committed effective dose. The main objective of this study was to assess the impacts of the new OIR biokinetic models and dose coefficients compared with the old ones based on ICRP Publications 60/78/119. CIEMAT and the University of Salamanca in Spain have been working on re-interpreting in vitro bioassay data obtained from workers with long-term exposure to inhaling uranium oxides during the fabrication of nuclear fuel elements using low enriched uranium, which is now classified as OIR type M/S material. The effects on the retention/excretion models and dose coefficients were studied for 234U, 235U, 238U, and the uranium mixture. Committed effective doses were re-assessed using BIOKMOD code by applying the new OIR excretion model and dose coefficients considering acute and/or chronic inhalation intake scenarios. A reduction by roughly a factor of four was obtained compared with former doses based on ICRP Publications 60/78/119, which is an important effect to take into consideration.

Specific Absorbed Fractions for Reference Paediatric Individuals.

The calculation of doses to organs and tissues of interest due to internally emitting radionuclides requires knowledge of the time-dependent distribution of the radionuclide, its physical decay properties, and the fraction of emitted energy absorbed per mass of the target. The latter property is quantified as the specific absorbed fraction (SAF). This publication provides photon, electron, alpha particle, and neutron (for nuclides undergoing spontaneous fission) SAF values for the suite of reference individuals. The reference individuals are defined largely by information provided in ICRP Publication 89. Some improvements and additional data are provided in this publication which define the reference individual's source and target region masses used in the Occupational Intake of Radionuclides (OIR) and Dose Coefficients for Intakes of Radionuclides by Members of the Public series of publications. The set of reference individuals includes males and females at 0 (newborn), 1, 5, 10, 15, and 20 (adult) years of age. The reference adult masses and SAFs provided in this publication are identical to those in ICRP Publication 133 and those used in the OIR series of publications. Computation of SAF values involves simulating radiation transport in computational models which represent the geometry of the reference individuals. The reference voxel phantoms of ICRP Publication 143 are used for photon and neutron transport, and most electron transport. Alpha particle transport is not necessary for large tissue regions as the short range allows for an assumption of full energy absorption (absorbed fraction of unity) for self-irradiation geometries. Additional computational models are needed for charged particles in small, overlapping, or interlaced geometries. Stylised models are described and used for electrons and alpha particles in the alimentary and respiratory tract regions. Image-based models are used to compute SAFs for charged particles within the skeleton. This publication is accompanied by an electronic supplement which includes files containing SAFs for each radiation type in each reference individual. The supplement also includes source and target region masses for each reference individual, as well as skeletal dose-response functions for photons incident upon the skeleton.© 2024 ICRP. Published by SAGE.

Интеграция моделей oir мкрз в дозиметрическую систему idose 2

Introduction: The iDose 2 dosimetry system is a tool for assessing the doses of internal irradiation of workers under the current individual dosimetry control (IDC). In this system, according to a series of measurements of the activity of radionuclides in biological objects (including those not exceeding the detection limit of the measurement technique) and information on contact times and types of compounds, estimates of the committed effective dose equivalent (CEDE) of internal irradiation, as well as their uncertainties, are made based on the Bayesian approach. It is possible to integrate practically any biokinetic models of the behavior of radionuclides in the human body, presented in the form of a system of ordinary differential equations (ODEs) with constant transition coefficients between compartments, into the iDose 2 dosimetry system without changing the source code. Purpose: Integration of new combined biokinetic models for the list of radionuclides: H-3, Sr-90, Cs-137, Pu-238, Pu-239 and Am-241 from Publications 100, 130, 134, 137 and 141 of the ICRP (conventionally called the series Occupational Intakes of Radionuclides (OIR)), for ingestion and inhalation routes of intake with AMAD = 1 and 5 microns. Material and methods: For each variant of the biokinetic model, the functions of retention/removal of radionuclides were found through the eigenvectors and eigenvalues of the matrix describing the ODE system. Results: A total of 65 new biokinetic models and 180 functions of radionuclide retention/removal in the form of a sum of exponents were integrated and quality control was carried out.

IDAC-Bio, A Software for Internal Dosimetry Based on the New ICRP Biokinetic Models and Specific Absorbed Fractions.

Radiation dosimetry is central to virtually all radiation safety applications, optimization, and research. It relates to various individuals and population groups and to miscellaneous exposure situations—including planned, existing, and emergency situations. The International Commission on Radiological Protection (ICRP) has developed a new computational framework for internal dose estimations. Important components are more detailed and improved anatomical models and more realistic biokinetic models than before. The ICRP is currently producing new organ dose and effective dose coefficients for occupational intakes of radionuclides (OIR) and environmental intakes of radionuclides (EIR), which supersede the earlier dose coefficients in Publication 68 and the Publication 72 series, respectively. However, the ICRP only publishes dose coefficients for a single acute intake of a radionuclide and for an integration period of 50 years for intake by adults and to age 70 years for intakes by pre-adults. The new software, IDAC-Bio, performs committed absorbed dose and effective dose calculations for a selectable intake scenario, e.g., for a continuous intake or an intake during x hours per day and y days per week, and for any selected integration time. The software uses the primary data and models of the ICRP biokinetic models and numerically solves the biokinetic model and calculates the absorbed doses to organs and tissues in the ICRP reference human phantoms. The software calculates absorbed dose using the nuclear decay data in ICRP publication 107. IDAC-Bio is a further development and an important addition to the internal dosimetry program IDAC-Dose2.1. The results generated by the software were validated against published ICRP dose coefficients. The potential of the software is illustrated by dose calculations for a nuclear power plant worker who had been exposed to varying levels of 60Co and who had undergone repeated whole-body measurements, and for a hypothetical member of the public subject to future releases of 148Gd from neutron spallation in tungsten at the European Spallation Source.

Basis for the ICRP’s updated biokinetic model for systemic astatine

The International Commission on Radiological Protection (ICRP) recently updated its biokinetic models for workers in a series of reports called the OIR (occupational intakes of radionuclides) series. A new biokinetic model for astatine (At), the heaviest member of the halogen family, was adopted in OIR Part 5 (ICRP in press). Occupational intakes of radionuclides: Part 5). This paper provides an overview of available biokinetic data for At; describes the basis for the ICRP’s updated model for At; and tabulates dose coefficients for intravenous injection of each of the two longest lived and most important At isotopes, 211At and 210At. At-211 (T 1/2 = 7.214 h) is a promising radionuclide for use in targeted α-particle therapy due to several favourable properties including its half-life and the absence of progeny that could deliver significant radiation doses outside the region of α-particle therapy. At-210 (T 1/2 = 8.1 h) is an impurity generated in the production of 211At in a cyclotron and represents a potential radiation hazard via its long-lived progeny 210Po (T 1/2 = 138 days). Tissue dose coefficients for injected 210At and 211At based on the updated model are shown to differ considerably from values based on the ICRP’s previous model for At, particularly for the thyroid, stomach wall, salivary glands, lungs, spleen, and kidneys.

A Preliminary Study of Radon Equilibrium Factor at a Tourist Cave in Okinawa, Japan

The International Commission on Radiological Protection (ICRP) issued its Publication 137, Occupational Intakes of Radionuclides: Part 3 in which the radon equilibrium factor is fixed as 0.4 for tourist caves; however, several studies have reported a different value for the factor and its seasonal variation has also been observed. In this study, the radon concentration, equilibrium equivalent radon concentration and meteorological data were measured, and the equilibrium factor was evaluated in a tourist cave, Gyokusen-do Cave located in the southern part of Okinawa Island in southwestern Japan. Radon concentrations were measured with an AlphaGUARD and their corresponding meteorological data were measured with integrated sensors. Equilibrium equivalent radon concentration was measured with a continuous air monitor. The measured radon concentrations tended to be low in winter and high in summer, which is similar to previously obtained results. By contrast, the equilibrium factor tended to be high in winter (0.55 ± 0.09) and low in summer (0.24 ± 0.15), with a particularly large fluctuation in summer. It was concluded that measurements in different seasons are necessary for proper evaluation of radon equilibrium factor.

A biokinetic model for systemic sodium

This paper describes an updated biokinetic model for systemic sodium (Na), developed for use in a series of reports by the International Commission on Radiological Protection (ICRP) on occupational intake of radionuclides. In contrast to the ICRP’s previous model for intake of radio-sodium by workers, the updated model depicts realistic directions of movement of Na in the body including recycling of activity between blood and tissues. The updated model structure facilitates extension of the baseline transfer coefficients for adults to different age groups and to special exposure scenarios such as transfer of radio-sodium from the mother to the foetus or the nursing infant. Dose coefficients for 22Na and 24Na based on the updated model generally do not differ greatly from those based on the ICRP’s previous Na model when both models are connected to the ICRP’s latest dosimetry system. The main exception is that the updated model yields roughly twofold higher dose coefficients for endosteal bone surface than does the previous model due to the dosimetrically cautious assumption in the updated model that exchangeable Na in bone resides on bone surface.

Dosimetry of inhaled 219Rn progeny.

During prostate cancer treatment with 223Ra. 219Rn (actinon) occurs and may be exhaled by the patient. Nurses and other hospital employees may inhale this radionuclide and its decay products. The alpha-emitting decay products of actinon deposited within a body will irradiate tissues and organs. Therefore. it is necessary to evaluate organ doses of actinon progeny. The purpose of this study is to set up a dosimetric method to assess dose coefficients for actinon progeny. The effective dose coefficients were calculated separately for three modes. The unattached mode which concerned the activity median thermodynamic diameter (AMTD) of 1 nm. and the nucleation and accumulation modes which are represented by activity median aerodynamic diameters (AMAD) of 60 and 500 nm respectively. The recent biokinetic models of actinon progeny developed in the Occupational Intakes of Radionuclides (OIR) publications series of the International Commission of Radiological Protection (ICRP) were implemented on BIOKMOD (Biokinetic Modeling) to calculate the number of nuclear transformations per activity intake of actinon progeny. The organ equivalent and effective dose coefficients were determined using the dosimetric approach of the ICRP. The inhalation dose coefficients of actinon progeny are dominated by the contribution of lung dose. The calculated dose coefficients of 211Pb and 211Bi are 5.78 × 10−8 and 4.84 × 10−9 Sv.Bq−1 for unattached particles (AMTD = 1 nm). and 1.4 × 10−8 and 3.55 × 10−9 Sv.Bq−1 for attached particles (AMAD = 60 nm). and 7.37 × 10−9 and 1.91 × 10−9 Sv.Bq−1 for attached particles (AMAD = 500 nm). These values are much closer to those of the recently published ICRP 137.

INTERNAL DOSIMETRY TOOL FOR THE IMPLEMENTATION AND USE OF NEW ICRP/OIR MODELS: A CAESIUM STUDY.

In case of accidental intake of radioactive material, the dose assessment requires information on the radionuclide distribution in the body. Measurements of retention or excretion are compared with model's predictions to estimate the intake. The reference models for internal dosimetry purposes are those proposed by ICRP and have been recently updated in the 'Occupational Intakes of Radionuclides (OIR)' series. This work focusses on the implementation of the new caesium model in a computer package, ICRP130Models (integrated in the toolbox BIOKMOD), together with new features aimed at estimating the intake and the dose from bioassay measurements. The software gives the best fitted intake and the committed effective dose, as well as the chi-squared test and the p-value as an indication of the goodness of the fit in the assessment process. The possibility to build the model as a function of a parameter allows the user to look for different options when fitting the bioassay measurements.

A biokinetic model for trivalent or hexavalent chromium in adult humans

Chromium exists in several oxidation states, with the trivalent state (Cr(III)) being the dominant naturally occurring form. Chromium in other oxidation states tends to be converted to the trivalent oxide in the natural environment and in biological systems. Chromium(III) has been shown to be an essential nutrient for humans and several non-human species. Chromium(VI), the second most stable form of chromium, is an important environmental contaminant that is mostly of industrial origin and is associated with lung cancer and nose tumours in chromium workers. This paper proposes a biokinetic model for chromium that addresses the distinctive behaviours of Cr(III) and Cr(VI) following uptake to blood of an adult human. The model is based on biokinetic data derived from relatively short-term studies involving administration of chromium tracers to adult human subjects or laboratory animals, supplemented with data on the long-term distribution of chromium in adult humans as estimated from autopsy measurements. The model is part of a comprehensive update of biokinetic models of the International Commission on Radiological Protection, used to project or evaluate radiation doses from occupational intake of radionuclides.

ICRP Publication 141: Occupational Intakes of Radionuclides: Part 4.

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).

Updated biokinetic model for systemic americium

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.

Biokinetics of 238Pu oxides: inferences from bioassay data

The bioassay data collected from several workers involved in 238Pu inhalation incidents have been analysed using the most recent biokinetic models described in the Occupational Intakes of Radionuclides (OIR) series of publications. Although all exposures were thought to be to 238Pu oxides, the observed urinary excretion patterns differed in different inhalation incidents. The urinary excretion from individuals involved in one of the incidents increased steadily with time, peaking around two to three years before decreasing. This pattern is described in Part 4 of the OIR series using the ‘238PuO2, ceramic’ model. This non-monotonic behaviour, explained as being due to fragmentation and dissolution, was not specific to the incident, but observed in other incidents. The urinary excretion data collected from individuals involved in another incident showed dissolution behaviour between Type M and Type S. Finally, the bioassay data from yet another incident showed a pattern that appears to represent behaviour more insoluble than Type S, which is possibly a result of self-heating due to the decay heat from 238Pu. The urinary excretion patterns and corresponding dose coefficients have been calculated and compared.

Effective dose coefficients for inhaled radon and its progeny: ICRP’s approach

The International Commission on Radiological Protection (ICRP) has recently published three reports on radon exposure: (i) Publication 115 on lung cancer risks from radon and radon progeny [1], (ii) Publication 126 on radiological protection against radon exposure [2] and (iii) Publication 137 on Occupational Intakes of Radionuclides (OIR), Part 3 [3]. The latter document gives doses coefficients for the inhalation of radon, thoron and their airborne progeny as well as recommendations for their use for the protection of workers. As with all other radionuclides, the effective dose coefficients are calculated with ICRP reference biokinetic and dosimetric models. Sufficient information and dosimetric data are given so that site-specific dose coefficients can be calculated based on measured aerosol parameter values.

Corrigendum.

Corrigendum to ICRP Publication 137: Occupational intakes of radionuclides: Part 3. [Ann. ICRP 46(3/4), 2017]. DOI: 10.1177/0146645317734963 . The following errors have been identified. ICRP apologises for any inconvenience or confusion caused by these errors. [Table: see text][Table: see text][Table: see text][Table: see text].

ICRP Task Group 95: internal dose coefficients.

Internal doses are calculated using biokinetic and dosimetric models. These models describe the behaviour of the radionuclides after ingestion, inhalation, and absorption to the blood, and the absorption of the energy resulting from their nuclear transformations. The International Commission on Radiological Protection (ICRP) develops such models and applies them to provide dose coefficients and bioassay functions for the calculation of equivalent or effective dose from knowledge of intakes and/or measurements of activity in bioassay samples. Over the past few years, ICRP has devoted a considerable amount of effort to the revision and improvement of models to make them more physiologically realistic representations of uptake and retention in organs and tissues, and of excretion. Provision of new biokinetic models, dose coefficients, monitoring methods, and bioassay data is the responsibility of Committee 2 and its task groups. Three publications in a series of documents replacing the ICRP Publication 30 series and ICRP Publications 54, 68, and 78 have been issued [Occupational Intakes of Radionuclides (OIR) Parts 1-3]. OIR Part 1 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 Parts 2-5 provide data on individual elements and their radioisotopes. Work is also in progress on revision of dose coefficients for radionuclide intakes by members of the public.

ICRP Publication 137: Occupational Intakes of Radionuclides: Part 3.

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).

RADON DOSIMETRY FOR WORKERS: ICRP'S APPROACH.

The International Commission on Radiological Protection (ICRP) has recently published two reports on radon exposure; Publication 115 on lung cancer risks from radon and radon progeny and Publication 126 on radiological protection against radon exposure. A specific graded approach for the control of radon in workplaces is recommended where a dose assessment is required in certain situations. In its forthcoming publication on Occupational Intakes of Radionuclides (OIR) document, Part 3, effective dose coefficients for radon and thoron will be provided. These will be calculated using ICRP reference biokinetic and dosimetric models. Sufficient information and dosimetric data will be given so that site-specific dose coefficients can be calculated based on measured aerosol parameter values. However, ICRP will recommend a single dose coefficient of 12 mSv per working level month (WLM) for inhaled radon progeny to be used in most circumstances. This chosen reference value was based on both dosimetry and epidemiological data. In this paper, the application and use of dose coefficients for workplaces are discussed including the reasons for the choice of the reference value. Preliminary results of dose calculations for indoor workplaces and mines are presented. The paper also briefly describes the general approach for the management of radon exposure in workplaces based both on ICRP recommendations and the European directive (2013/59/EURATOM).

Basis for the ICRP’s updated biokinetic model for carbon inhaled as CO2

The International Commission on Radiological Protection (ICRP) is updating its biokinetic and dosimetric models for occupational intake of radionuclides (OIR) in a series of reports called the OIR series. This paper describes the basis for the ICRP’s updated biokinetic model for inhalation of radiocarbon as carbon dioxide (CO2) gas. The updated model is based on biokinetic data for carbon isotopes inhaled as carbon dioxide or injected or ingested as bicarbonate The data from these studies are expected to apply equally to internally deposited (or internally produced) carbon dioxide and bicarbonate based on comparison of excretion rates for the two administered forms and the fact that carbon dioxide and bicarbonate are largely carried in a common form (CO2–H in blood. Compared with dose estimates based on current ICRP biokinetic models for inhaled carbon dioxide or ingested carbon, the updated model will result in a somewhat higher dose estimate for 14C inhaled as CO2 and a much lower dose estimate for 14C ingested as bicarbonate.