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Tissue reactions and stochastic effects after exposure to ionising radiation are variable between individuals but the factors and mechanisms governing individual responses are not well understood. Individual responses can be measured at different levels of biological organization and using different endpoints following varying doses of radiation, including: cancers, non-cancer diseases and mortality in the whole organism; normal tissue reactions after exposures; and, cellular endpoints such as chromosomal damage and molecular alterations. There is no doubt that many factors influence the responses of people to radiation to different degrees. In addition to the obvious general factors of radiation quality, dose, dose rate and the tissue (sub)volume irradiated, recognized and potential determining factors include age, sex, life style (e.g., smoking, diet, possibly body mass index), environmental factors, genetics and epigenetics, stochastic distribution of cellular events, and systemic comorbidities such as diabetes or viral infections. Genetic factors are commonly thought to be a substantial contributor to individual response to radiation. Apart from a small number of rare monogenic diseases such as ataxia telangiectasia, the inheritance of an abnormally responsive phenotype among a population of healthy individuals does not follow a classical Mendelian inheritance pattern. Rather it is considered to be a multi-factorial, complex trait.

This paper summarises the view of the German Commission on Radiological Protection ("Strahlenschutzkommission", SSK) on the rationale behind the currently valid dose limits and dose constraints for workers recommended by the International Commission on Radiological Protection (ICRP). The paper includes a discussion of the reasoning behind current dose limits followed by a discussion of the detriment used by ICRP as a measure for stochastic health effects. Studies on radiation-induced cancer are reviewed because this endpoint represents the most important contribution to detriment. Recent findings on radiation-induced circulatory disease that are currently not included in detriment calculation are also reviewed. It appeared that for detriment calculations the contribution of circulatory diseases plays only a secondary role, although the uncertainties involved in their risk estimates are considerable. These discussions are complemented by a review of the procedures currently in use in Germany, or in discussion elsewhere, to define limits for genotoxic carcinogens. To put these concepts in perspective, actual occupational radiation exposures are exemplified with data from Germany, for the year 2012, and regulations in Germany are compared to the recommendations issued by ICRP. Conclusions include, among others, considerations on radiation protection concepts currently in use and recommendations of the SSK on the limitation of annual effective dose and effective dose cumulated over a whole working life.

An international intercomparison was organized by Working Group 7, Internal Dosimetry, of the European Radiation Dosimetry Group in collaboration with Working Group 6, Computational Dosimetry, for measurement and Monte Carlo simulation of Am in three skull phantoms. The main objectives of this combined exercise were (1) comparison of the results of counting efficiency in fixed positions over each head phantom using different germanium detector systems, (2) calculation of the activity of Am in the skulls, (3) comparison of Monte Carlo simulations with measurements (spectrum and counting efficiency), and (4) comparison of phantom performance. This initiative collected knowledge on equipment, detector arrangements, calibration procedures, and phantoms used around the world for in vivo monitoring of Am in exposed persons, as well as on the Monte Carlo skills and tools of participants. Three skull phantoms (BfS, USTUR, and CSR phantoms) were transported from Europe (10 laboratories) to North America (United States and Canada). The BfS skull was fabricated with real human bone artificially labeled with Am. The USTUR skull phantom was made from the US Transuranium and Uranium Registries whole-body donor (case 0102) who was contaminated due to an occupational intake of Am; one-half of the skull corresponds to real contaminated bone, the other half is real human bone from a noncontaminated person. Finally, the CSR phantom was fabricated as a simple hemisphere of equivalent bone and tissue material. The three phantoms differ in weight, size, and shape, which made them suitable for an efficiency study. Based on their own skull calibration, the participants calculated the activity in the three European Radiation Dosimetry Group head phantoms. The Monte Carlo intercomparison was organized in parallel with the measurement exercise using the voxel representations of the three physical phantoms; there were 16 participants. Three tasks were identified with increasing difficulty: (1) Monte Carlo simulation of the simple CSR hemisphere and the Helmholz Zentrum München high-purity germanium detector for calculating the counting efficiency for the 59.54 keV photons of Am, in established measurement geometry; (2) Monte Carlo simulation of particular measurement geometries using the BfS and USTUR voxel phantoms and the Helmholz Zentrum München high-purity germanium detector detector; and (3) application of Monte Carlo methodology to calculate the calibration factor of each participant for the detector system and counting geometry (single or multidetector arrangement) to be used for monitoring a person in each in vivo facility, using complex skull phantoms. The results of both exercises resulted in the conclusion that none of the three available head phantoms is appropriate as a reference phantom for the calibration of germanium detection systems for measuring Am in exposed adult persons. The main reasons for this are: (1) lack of homogeneous activity distribution in the bone material, or (2) inadequate shape/size for simulating an adult skull. Good agreement was found between Monte Carlo results and measurements, which supports Monte Carlo calibration of body counters as an alternative method when appropriate physical phantoms are not available and the detector and source are well known.

This paper provides a summary of the Education and Training (E&T) activities that have been developed and organised by the European Radiation Dosimetry Group (EURADOS) in recent years and in the case of Training Courses over the last decade. These E&T actions include short duration Training Courses on well-established topics organised within the activity of EURADOS Working Groups (WGs), or one-day events integrated in the EURADOS Annual Meeting (workshops, winter schools, the intercomparison participants' sessions and the learning network, among others). Moreover, EURADOS has recently established a Young Scientist Grant and a Young Scientist Award. The Grant supports young scientists by encouraging them to perform research projects at other laboratories of the EURADOS network. The Award is given in recognition of excellent work developed within the WGs' work programme. Additionally, EURADOS supports the dissemination of knowledge in radiation dosimetry by promoting and endorsing conferences such as the individual monitoring (IM) series, the neutron and ion dosimetry symposia (NEUDOS) and contributions to E&T sessions at specific events.

The aim of the International Commission on Radiological Protection (ICRP) is to protect humans against cancer and other diseases and effects associated with exposure to ionising radiation, and also to protect the environment, without unduly limiting the beneficial use of ionising radiation. As of the second half of 2017, four committees are contributing to the overall mission of ICRP, including Committee 1 (Radiation Effects). The role of Committee 1 includes consideration of the risks and mechanisms of induction of cancer and heritable disease; discussion of the risks, severity, and mechanisms of induction of tissue/organ damage and developmental defects; and review of the effects of ionising radiation on non-human biota at population level. This paper gives an overview of the recent activities of Committee 1, and discusses the focus of its active task groups.

Since the early 1980s, the European Radiation Dosimetry Group (EURADOS) has been maintaining a network of institutions interested in the dosimetry of ionising radiation. As of 2017, this network includes more than 70 institutions (research centres, dosimetry services, university institutes, etc.), and the EURADOS database lists more than 500 scientists who contribute to the EURADOS mission, which is to promote research and technical development in dosimetry and its implementation into practice, and to contribute to harmonisation of dosimetry in Europe and its conformance with international practices. The EURADOS working programme is organised into eight working groups dealing with environmental, computational, internal, and retrospective dosimetry; dosimetry in medical imaging; dosimetry in radiotherapy; dosimetry in high-energy radiation fields; and harmonisation of individual monitoring. Results are published as freely available EURADOS reports and in the peer-reviewed scientific literature. Moreover, EURADOS organises winter schools and training courses on various aspects relevant for radiation dosimetry, and formulates the strategic research needs in dosimetry important for Europe. This paper gives an overview on the most important EURADOS activities. More details can be found at www.eurados.org .

In order to quantify radiation risks at exposure scenarios relevant for radiation protection, often extrapolation of data obtained at high doses and high dose rates down to low doses and low dose rates is needed. Task Group TG91 on 'Radiation Risk Inference at Low-dose and Low-dose Rate Exposure for Radiological Protection Purposes' of the International Commission on Radiological Protection is currently reviewing the relevant cellular, animal and human studies that could be used for that purpose. This paper provides an overview of dose rates and doses typically used or present in those studies, and compares them with doses and dose rates typical of those received by the A-bomb survivors in Japan.

Radon gas concentrations in eight basements, four living rooms and four caves from different locations in Kabul and Panjsher, Afghanistan, were measured by using eight active radon exposure meters recently developed by the Helmholtz Center in Munich, Germany. The two-phase measurements lasted from a week to a year. In the first phase of measurements which lasted one week, the mean activity concentrations ranged from 6 to 120 Bq/m3 and 25 to 139 Bq/m3 for the basements and caves, respectively. In the second phase of measurements which lasted one year, the mean activity concentrations ranged from 33 to 2064 Bq/m3 and the corresponding effective annual doses calculated for the inhabitants were in the range between 0.6 and 33.4 mSv. As some of the values are rather high and exceed the recommended recommendations by IAEA and ICRP, based on the local conditions a number of simple recommendations has been proposed for the possible reduction of effective annual dose caused by radon in the measurement locations., (© The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.)

Purpose: Estimated radiation risks used for radiation protection purposes have been based primarily on the Life Span Study (LSS) of atomic bomb survivors who received brief exposures at high dose rates, many with high doses. Information is needed regarding radiation risks from low dose-rate (LDR) exposures to low linear-energy-transfer (low-LET) radiation. We conducted a meta-analysis of LDR epidemiologic studies that provide dose-response estimates of total solid cancer risk in adulthood in comparison to corresponding LSS risks, in order to estimate a dose rate effectiveness factor (DREF)., Materials and Methods: We identified 22 LDR studies with dose-response risk estimates for solid cancer after minimizing information overlap. For each study, a parallel risk estimate was derived from the LSS risk model using matching values for sex, mean ages at first exposure and attained age, targeted cancer types, and accounting for type of dosimetric assessment. For each LDR study, a ratio of the excess relative risk per Gy (ERR Gy -1 ) to the matching LSS ERR risk estimate (LDR/LSS) was calculated, and a meta-analysis of the risk ratios was conducted. The reciprocal of the resultant risk ratio provided an estimate of the DREF., Results: The meta-analysis showed a LDR/LSS risk ratio of 0.36 (95% confidence interval [CI] 0.14, 0.57) for the 19 studies of solid cancer mortality and 0.33 (95% CI 0.13, 0.54) when three cohorts with only incidence data also were added, implying a DREF with values around 3, but statistically compatible with 2. However, the analyses were highly dominated by the Mayak worker study. When the Mayak study was excluded the LDR/LSS risk ratios increased: 1.12 (95% CI 0.40, 1.84) for mortality and 0.54 (95% CI 0.09, 0.99) for mortality + incidence, implying a lower DREF in the range of 1-2. Meta-analyses that included only cohorts in which the mean dose was <100 mGy yielded a risk ratio of 1.06 (95% CI 0.30, 1.83) for solid cancer mortality and 0.58 (95% CI 0.10, 1.06) for mortality + incidence data., Conclusions: The interpretation of a best estimate for a value of the DREF depends on the appropriateness of including the Mayak study. This study indicates a range of uncertainty in the value of DREF between 1 and about 2 after protracted radiation exposure. The LDR data provide direct evidence regarding risk from exposures at low dose rates as an important complement to the LSS risk estimates used for radiation protection purposes.

Purpose: Biologically-based mechanistic models that are used in combining current understanding of human carcinogenesis with epidemiological studies were reviewed. Assessment was made of how well they fit the data, whether they account for non-linear radiobiological low-dose effects, and whether they suggest any implications for the dose response at low doses and dose rates. However, the present paper does not make an attempt to provide a complete review of the existing literature on biologically-based models and their application to epidemiological data., Conclusion: In most studies the two-stage clonal expansion (TSCE) model of carcinogenesis was used. The model provided robust estimates of identifiable parameters and radiation risk. While relatively simple, it is flexible, so that more stages can easily be added, and tests made of various types of radiation action. In general, the model performed similarly or better than descriptive excess absolute and excess relative risk models, in terms of quality of fit and number of parameters. Only very rarely the shape of dose-response predicted by the models was investigated. For some tumors, when more detailed biological information was known, additional pathways were included in the model. The future development of these models will benefit from growing knowledge on carcinogenesis processes, and in particular from use of biobank tissue samples and advances in omics technologies. Their use appears a promising approach to investigate the radiation risk at low doses and low dose rates. However, the uncertainties involved are still considerable, and the models provide only a simplified description of the underlying complexity of carcinogenesis. Current assumptions in radiation protection including the linear-non-threshold (LNT) model are not in contradiction to what is presently known on the process of cancer development.

Background and Purpose: Systematic investigation of the energy and angular dependence of secondary neutron fluence energy distributions and ambient dose equivalents values (H*(10)) inside a pencil beam scanning proton therapy treatment room using a gantry., Materials and Methods: Neutron fluence energy distributions were measured with an extended-range Bonner sphere spectrometer featuring ³He proportional counters, at four positions at 0°, 45°, 90°, and 135° with respect to beam direction and at a distance of 2 m from the isocenter. The energy distribution of secondary neutrons was investigated for initial proton beam energies of 75 MeV, 140 MeV, and 200 MeV, respectively, using a 2D scanned irradiation field of 11 × 11 cm² delivered to a 30 × 30 × 30 cm³ PMMA phantom. Additional measurements were performed at a proton energy of 118 MeV including a 5 cm range-shifter (PMMA), yielding a Bragg peak position similar to that of 75 MeV protons., Results: Ambient dose equivalent values from 0.3 μSv/Gy (75 MeV; 90°) to 24 μSv/Gy (200 MeV; 0°) were measured inside the treatment room at a distance of 2 m from the isocenter. H*(10) values were lower (by factors of up to 7.2 (at 45°)) at 75 MeV compared to those at 118 MeV with the 5 cm range-shifter. At 0° and 45°, an evaporation peak was found in the measured neutron fluence energy distributions, at neutron energies around MeV, which contributes about 50% to total H*(10) values, for all investigated proton beam energies., Conclusions: This study showed a pronounced increase of secondary neutron H*(10) values inside the proton treatment room with increasing proton energy without beam modifiers. For example, in beam direction this increase was about a factor of 50 when protons of 75 MeV and 200 MeV were compared. The existence of a peak of secondary neutrons in the MeV region was demonstrated in beam direction (0°). This peak is due to evaporation neutrons produced in the existing surrounding materials such as those used for the gantry. Therefore, any simulation of the secondary neutrons within a proton treatment room must take these materials into account. In addition, the results obtained here show that the use of a range-shifter increases the production of secondary neutrons inside the treatment room. Using a range-shifter, the higher neutron doses observed mainly result from the higher incident proton energy (118 MeV instead of 75 MeV when no range-shifter was used), due to higher neutron production cross-sections., (© 2017 American Association of Physicists in Medicine.)

Many archaeological sites in Egypt are unique worldwide, such as ancient tombs and pyramids, because they document fundamental developments in human civilization that took place several thousands of years ago. For this reason, these sites are visited by numerous visitors every year. The present work is devoted to provide a pre-operational radiological baseline needed to quantify occupational radiation exposure at the famous Saqqara region in Cairo, Egypt. A hyperpure Ge detector has been used in the γ-ray spectrometric analysis while the (222)Rn concentration was measured using a portable radon monitor RTM 1688-2, SARAD. The mean specific activities of (226)Ra, (232)Th and (40)K in the samples collected from the interior walls of the Saqqara tombs were determined and found to show average values of 16, 8.5 and 45 Bq kg(-1), respectively. The concentration of radon was measured inside the tombs Serapeum, South tomb and the Zoser Pyramid (fifth level) and an associated average working level of 0.83 WL was obtained. In order to avoid the health hazards associated with the exposure to radon during the long period of work inside these tombs, proposed solutions are introduced.

Two people were exposed to and contaminated with 241 Am. In vivo determinations of the incorporated 241 Am were performed using a whole-body counter and two partial-body counters for the skull and lung, respectively. Additionally, urine samples were analysed to estimate the systemic activity removed from the body. To improve the geometry of the skull measurements, an optimised detector configuration was used, a calibration with three physical phantoms of the human head was conducted, and the morphological variability between the individuals was also considered. The results of the measurements indicate that activity is not deposited in the deep tissues, rather in the skin tissues close to the body surface. Unfortunately, the many open questions relating to the actual circumstances during and after the incident make the interpretation of this case difficult if at all possible.

ICRP suggested a strategy based on the distinction between a protection approach for dwellings and one for workplaces in the previous recommendations on radon. Now, the Commission recommends an integrated approach for the protection against radon exposure in all buildings irrespective of their purpose and the status of their occupants. The strategy of protection in buildings, implemented through a national action plan, is based on the application of the optimisation principle below a derived reference level in concentration (maximum 300 Bq m(-3)). A problem, however, arises that due to new epidemiological findings and application of dosimetric models, ICRP 115 (Ann ICRP 40, 2010) presents nominal probability coefficients for radon exposure that are approximately by a factor of 2 larger than in the former recommendations of ICRP 65 (Ann ICRP 23, 1993). On the basis of the so-called epidemiological approach and the dosimetric approach, the doubling of risk per unit exposure is represented by a doubling of the dose coefficients, while the risk coefficient of ICRP 103 (2007) remains unchanged. Thus, an identical given radon exposure situation with the new dose coefficients would result in a doubling of dose compared with the former values. This is of serious conceptual implications. A possible solution of this problem was presented during the workshop.

A new electronic radon/thoron monitor employing semiconductor detectors based on a passive diffusion chamber design has been recently developed at the Helmholtz Zentrum München (HMGU). This device allows for acquisition of alpha particle energy spectra, in order to distinguish alpha particles originating from radon and radon progeny decays, as well as those originating from thoron and its progeny decays. A Monte-Carlo application is described which uses the Geant4 toolkit to simulate these alpha particle spectra. Reasonable agreement between measured and simulated spectra were obtained for both (220)Rn and (222)Rn, in the energy range between 1 and 10 MeV. Measured calibration factors could be reproduced by the simulation, given the uncertainties involved in the measurement and simulation. The simulated alpha particle spectra can now be used to interpret spectra measured in mixed radon/thoron atmospheres. The results agreed well with measurements performed in both radon and thoron gas environments. It is concluded that the developed simulation allows for an accurate prediction of calibration factors and alpha particle energy spectra., (Copyright © 2016 Elsevier Ltd. All rights reserved.)

Abstract –: In this publication, the International Commission on Radiological Protection (ICRP) provides updated guidance on radiological protection from cosmic radiation in aviation, taking into account the current ICRP system of radiological protection, the latest available data on exposures in aviation, and experience gained worldwide in the management of exposures in aviation. The publication describes the origins of cosmic radiation, how it exposes passengers and aircraft crew, the basic radiological protection principles that apply to this existing exposure situation, and the available protective actions. For implementation of the optimisation principle, the Commission recommends a graded approach proportionate to the level of exposure that may be received by individuals. The objective is to keep the exposure of the most exposed individuals to a reasonable level. The Commission also recommends that information be disseminated to raise awareness about cosmic radiation, and to support informed decisions among concerned stakeholders.

Quantification of biological effects (cancer, other diseases, and cell damage) associated with exposure to ionising radiation has been a major issue for the International Commission on Radiological Protection (ICRP) since its foundation in 1928. While there is a wealth of information on the effects on human health for whole-body doses above approximately 100 mGy, the effects associated with doses below 100 mGy are still being investigated and debated intensively. The current radiological protection approach, proposed by ICRP for workers and the public, is largely based on risks obtained from high-dose and high-dose-rate studies, such as the Japanese Life Span Study on atomic bomb survivors. The risk coefficients obtained from these studies can be reduced by the dose and dose-rate effectiveness factor (DDREF) to account for the assumed lower effectiveness of low-dose and low-dose-rate exposures. The 2007 ICRP Recommendations continue to propose a value of 2 for DDREF, while other international organisations suggest either application of different values or abandonment of the factor. This paper summarises the current status of discussions, and highlights issues that are relevant to reassessing the magnitude and application of DDREF.

Since autumn 2012, the European Radiation Dosimetry Group (EURADOS) has been developing its Strategic Research Agenda (SRA), which is intended to contribute to the identification of future research needs in radiation dosimetry in Europe. The present article summarises-based on input from EURADOS Working Groups (WGs) and Voting Members-five visions in dosimetry and defines key issues in dosimetry research that are considered important for the next decades. The five visions include scientific developments required towards (a) updated fundamental dose concepts and quantities, (b) improved radiation risk estimates deduced from epidemiological cohorts, (c) efficient dose assessment for radiological emergencies, (d) integrated personalised dosimetry in medical applications and (e) improved radiation protection of workers and the public. The SRA of EURADOS will be used as a guideline for future activities of the EURADOS WGs. A detailed version of the SRA can be downloaded as a EURADOS report from the EURADOS website (www.eurados.org)., (© The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.)

The aim of the present study was to determine the internal dose in humans after the ingestion of soil highly contaminated with uranium. Therefore, an in vitro solubility assay was performed to estimate the bioaccessibility of uranium for two types of soil. Based on the results, the corresponding bioavailabilities were assessed by using a recently published method. Finally, these bioavailability data were used together with the biokinetic model of uranium to assess the internal doses for a hypothetical but realistic scenario characterized by a daily ingestion of 10 mg of soil over 1 year. The investigated soil samples were from two former uranium mining sites of Germany with (238)U concentrations of about 460 and 550 mg/kg. For these soils, the bioavailabilities of (238)U were quantified as 0.18 and 0.28 % (geometric mean) with 2.5th percentiles of 0.02 and 0.03 % and 97.5th percentiles of 1.48 and 2.34 %, respectively. The corresponding calculated annual committed effective doses for the assumed scenario were 0.4 and 0.6 µSv (GM) with 2.5th percentiles of 0.2 and 0.3 µSv and 97.5th percentiles of 1.6 and 3.0 µSv, respectively. These annual committed effective doses are similar to those from natural uranium intake by food and drinking water, which is estimated to be 0.5 µSv. Based on the present experimental data and the selected ingestion scenario, the investigated soils-although highly contaminated with uranium-are not expected to pose any major health risk to humans related to radiation.

The aim of the present study was to improve the estimation of soil-derived uranium absorption in humans. For this purpose, an in vitro solubility assay was combined with a human study by using a specific edible soil low in uranium. The mean bioaccessibility of the soil-derived uranium, determined by the solubility assay in artificial gastrointestinal fluid, was found to be 7.7% with a standard deviation of 0.2%. The corresponding bioavailability of the soil-derived uranium in humans was assumed to be log-normal distributed with a geometric mean of 0.04% and a 95% confidence interval ranging from 0.0049% to 0.34%. Both results were used to calculate a factor, denoted as fA(sol), which describes the relation between the bioaccessibility and the bioavailability of soil-derived uranium. The geometric mean of fA(sol) was determined to be 0.53% with a 95% confidence interval ranging from 0.06% to 4.43%. Based on fA(sol), it is possible to estimate more realistic values on the bioavailability of uranium for highly uranium-contaminated soils in humans by just performing the applied solubility assay. The results of this study can be further used to obtain more reliable results on the internal dose assessment of ingested highly uranium-contaminated soils.

This work compares the results of four different unfolding codes, MSANDB, MAXED, FRUIT and BONMA, which are based on different unfolding techniques. Additionally, Bayesian parameter estimation is also considered. All unfolding codes were supplied with the same set of input data acquired at the Environmental Research Station 'Schneefernerhaus' on the Zugspitze mountain, corresponding to continuous measurements of secondary neutrons from cosmic radiation. The HMGU high-energy extended Bonner sphere spectrometer (BSS), consisting of 16 measuring channels with (3)He proportional counters, was used as a reference BSS. The differences in the neutron spectra obtained with the different unfolding codes are discussed, and the uncertainties of integral quantities, like neutron fluence and ambient dose equivalent, are quantified., (© The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.)

Here the latest development stages of the HMGU active neutron dosemeter are presented. This work includes the comparison of the dosemeter's response function, calculated with Geant4, and the measurements in monoenergetic neutron fields at the Physikalisch Technische Bundesanstalt in Braunschweig, Germany. These results were used to match the response function and the count-to-dose conversion factors of the dosemeter to the Hp(10) personal dose equivalent., (© The Author 2013. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.)

For state-of-the-art discrimination of Radon and Thoron several measurement techniques can be used, such as active sampling, electrostatic collection, delayed coincidence method, and alpha-particle-spectroscopy. However, most of the devices available are bulky and show high power consumption, rendering them unfeasible for personal exposition monitoring. Based on a Radon exposure meter previously realized at the Helmholtz Center Munich (HMGU), a new electronic prototype for Radon/Thoron monitoring is currently being developed, which features small size and weight. Operating with pin-diode detectors, the low-power passive-sampling device can be used for continuous concentration measurements, employing alpha-particle-spectroscopy and coincidence event registration to distinguish decays originating either from Radon or Thoron isotopes and their decay products. In open geometry, preliminary calibration measurements suggest that one count per hour is produced by a 11 Bq m(-3) Radon atmosphere or by a 15 Bq m(-3) Thoron atmosphere. Future efforts will concentrate on measurements in mixed Radon/Thoron atmospheres.

Background: The Monte Carlo code GEANT4 was used to implement first steps towards a treatment planning program for fast-neutron therapy at the FRM II research reactor in Garching, Germany. Depth dose curves were calculated inside a water phantom using measured primary neutron and simulated primary photon spectra and compared with depth dose curves measured earlier. The calculations were performed with GEANT4 in two different ways, simulating a simple box geometry and splitting this box into millions of small voxels (this was done to validate the voxelisation procedure that was also used to voxelise the human body)., Results: In both cases, the dose distributions were very similar to those measured in the water phantom, up to a depth of 30 cm. In order to model the situation of patients treated at the FRM II MEDAPP therapy beamline for salivary gland tumors, a human voxel phantom was implemented in GEANT4 and irradiated with the implemented MEDAPP neutron and photon spectra. The 3D dose distribution calculated inside the head of the phantom was similar to the depth dose curves in the water phantom, with some differences that are explained by differences in elementary composition. The lateral dose distribution was studied at various depths. The calculated cumulative dose volume histograms for the voxel phantom show the exposure of organs at risk surrounding the tumor., Conclusions: In order to minimize the dose to healthy tissue, a conformal treatment is necessary. This can only be accomplished with the help of an advanced treatment planning system like the one developed here. Although all calculations were done for absorbed dose only, any biological dose weighting can be implemented easily, to take into account the increased radiobiological effectiveness of neutrons compared to photons.

Three facilities (CIEMAT, HMGU and HML) have used their in vivo counters to compare two leg phantoms. One was commercially produced with (241)Am activity artificially added to the bone inserts. The other, the United States Transuranium and Uranium Registries' (USTUR) leg phantom, was manufactured from (241)Am-contaminated bones resulting from an intake. The comparison of the two types of leg phantoms showed that the two phantoms are not similar in their activity distributions. An error in a bone activity estimate could be quite large if the commercial leg phantom is used to estimate what is contained in the USTUR leg phantom and, consequently, a real person. As the latter phantom was created as a result of a real contamination, it is deemed to be the more representative of what would actually happen if a person were internally contaminated with (241)Am.

The active radon exposure meter developed recently at the German Research Center for Environmental Health (Helmholtz Zentrum München) was used to measure radon concentrations in 12 tombs located in the Valley of the Kings, Egypt. Radon concentrations in air between 50 ± 7 and 12 100 ± 600 Bq m(-3) were obtained. The device was also used to measure individual radon exposures of those persons working as safeguards inside the tombs. For a measurement time of 2-3 d, typical individual radon exposures ranged from 1800 ± 400 to 240 000 ± 13 000 Bq h m(-3), depending on the duration of measurement and radon concentration in the different tombs. Based on current ICRP dose conversion conventions for workers and on equilibrium factors published in the literature for these tombs, individual effective dose rates that range from 1.5 ± 0.3 to 860 ± 50 µSv d(-1) were estimated. If it is assumed that the climatic conditions present at the measurement campaign persist for about half a year, in this area, then effective doses up to ∼ 66 mSv could be estimated for half a year, for some of the safeguards of tombs where F-values were known. To reduce the exposure of the safeguards, some recommendations are proposed.

A collaboration of the EURADOS working group on 'Internal Dosimetry' and the United States Transuranium and Uranium Registries (USTUR) has taken place to carry out an intercomparison on measurements and Monte Carlo modelling determining americium deposited in the bone of a USTUR leg phantom. Preliminary results and conclusions of this intercomparison exercise are presented here.

Wet and, to a lesser extent, dry deposition of atmospheric (129)I are known to represent the dominating processes responsible for (129)I in continental environmental samples that are remote from (129)I sources and not directly influenced by any liquid (129)I release of nuclear installations. Up to now, however, little is known about the major emitters and the related global deposition pattern of (129)I. In this work an overview over major sources of (129)I is given, and hitherto unknown time-dependent releases from these were estimated. Total gaseous (129)I releases from the US and former Soviet reprocessing facilities Hanford, Savannah River, Mayak, Seversk and Zheleznogorsk were found to have been 0.53, 0.27, 1.05, 0.23 and 0.14TBq, respectively. These facilities were thus identified as major airborne (129)I emitters. The global deposition pattern due to the (129)I released, depending on geographic latitude and longitude, and on time was studied using a box model describing the global atmospheric transport and deposition of (129)I. The model predictions are compared to (129)I concentrations measured by means of Accelerator Mass Spectrometry (AMS) in water samples that were collected from various lakes in Asia, Africa, America and New Zealand, and to published values. As a result, both pattern and temporal evolution of (129)I deposition values measured in and calculated for different types of environmental samples are, in general, in good agreement. This supports our estimate on atmospheric (129)I releases and the considered substantial transport and deposition mechanisms in our model calculations., (Copyright 2010 Elsevier B.V. All rights reserved.)

The detection of (41)Ca atoms in tooth enamel using accelerator mass spectrometry is suggested as a method capable of reconstructing thermal neutron exposures from atomic bomb survivors in Hiroshima and Nagasaki. In general, (41)Ca atoms are produced via thermal neutron capture by stable (40)Ca. Thus any (41)Ca atoms present in the tooth enamel of the survivors would be due to neutron exposure from both natural sources and radiation from the bomb. Tooth samples from five survivors in a control group with negligible neutron exposure were used to investigate the natural (41)Ca content in tooth enamel, and 16 tooth samples from 13 survivors were used to estimate bomb-related neutron exposure. The results showed that the mean (41)Ca/Ca isotope ratio was (0.17 +/- 0.05) x 10(-14) in the control samples and increased to 2 x 10(-14) for survivors who were proximally exposed to the bomb. The (41)Ca/Ca ratios showed an inverse correlation with distance from the hypocenter at the time of the bombing, similar to values that have been derived from theoretical free-in-air thermal-neutron transport calculations. Given that gamma-ray doses were determined earlier for the same tooth samples by means of electron spin resonance (ESR, or electron paramagnetic resonance, EPR), these results can serve to validate neutron exposures that were calculated individually for the survivors but that had to incorporate a number of assumptions (e.g. shielding conditions for the survivors).

(41)Ca is produced mainly by absorption of low-energy neutrons on stable (40)Ca. We used accelerator mass spectrometry (AMS) to measure (41)Ca in enamel of 16 teeth from 13 atomic bomb survivors who were exposed to the bomb within 1.2 km from the hypocenter in Hiroshima. In our accompanying paper (Wallner et al., Radiat. Res. 174, 000-000, 2010), we reported that the background-corrected (41)Ca/Ca ratio decreased from 19.5 x 10(-15) to 2.8 x 10(-15) with increasing distance from the hypocenter. Here we show that the measured ratios are in good correlation with gamma-ray doses assessed by electron paramagnetic resonance (EPR) in the same enamel samples, and agree well with calculated ratios based on either the current Dosimetry System 2002 (DS02) or more customized dose estimates where the regression slope as obtained from an errors-in-variables linear model was about 0.85. The calculated DS02 neutron dose to the survivors was about 10 to 80 mGy. The low-energy neutrons responsible for (41)Ca activation contributed variably to the total neutron dose depending on the shielding conditions. Namely, the contribution was smaller (10%) when shielding conditions were lighter (e.g., outside far away from a single house) and was larger (26%) when they were heavier (e.g., in or close to several houses) because of local moderation of neutrons by shielding materials. We conclude that AMS is useful for verifying calculated neutron doses under mixed exposure conditions with gamma rays.

Occupational exposures to ionising radiation mainly occur at low-dose rates and may accumulate effective doses of up to several hundred milligray. The objective of the present study is to evaluate the evidence of cancer risks from such low-dose-rate, moderate-dose (LDRMD) exposures. Our literature search for primary epidemiological studies on cancer incidence and mortality risks from LDRMD exposures included publications from 2002 to 2007, and an update of the UK National Registry for Radiation Workers study. For each (LDRMD) study we calculated the risk for the same types of cancer among the atomic bomb survivors with the same gender proportion and matched quantities for dose, mean age attained and mean age at exposure. A combined estimator of the ratio of the excess relative risk per dose from the LDRMD study to the corresponding value for the atomic bomb survivors was 1.21 (90% CI 0.51 to 1.90). The present analysis does not confirm that the cancer risk per dose for LDRMD exposures is lower than for the atomic bomb survivors. This result challenges the cancer risk values currently assumed for occupational exposures.

The assessment of the exposure to cosmic radiation onboard aircraft is one of the preoccupations of bodies responsible for radiation protection. Cosmic particle flux is significantly higher onboard aircraft than at ground level and its intensity depends on the solar activity. The dose is usually estimated using codes validated by the experimental data. In this paper, a comparison of various codes is presented, some of them are used routinely, to assess the dose received by the aircraft crew caused by the galactic cosmic radiation. Results are provided for periods close to solar maximum and minimum and for selected flights covering major commercial routes in the world. The overall agreement between the codes, particularly for those routinely used for aircraft crew dosimetry, was better than +/-20 % from the median in all but two cases. The agreement within the codes is considered to be fully satisfactory for radiation protection purposes.

EPCARD.Net as completely new code is based on the same approved physical algorithm as EPCARD version 3.34. As a major feature, many significant changes were made in the information technology area. There are only a few physical improvements adopted in the parameters database of the new EPCARD.Net. These are mainly 'dynamic' fluence-to-dose conversion coefficients and the most recent model of the world grid cut-off rigidity. Differences between EPCARD.Net and EPCARD version 3.34 are discussed in terms of effective dose. These differences turned out to be less than approximately 8 %.

Two Bonner sphere spectrometers (BSSs) have recently been installed to measure secondary neutrons from cosmic radiation continuously, one at the environmental research station 'Schneefernerhaus' at an altitude of 2650 m in Germany and the other at the Koldewey station close to the North Pole in Ny-Alesund, Spitsbergen. After unfolding, both systems provide neutron fluence energy distributions as a function of time. Based on these distributions and on fluence-to-dose conversion coefficients, mean ambient dose equivalent rate values of 75.0 +/- 2.9 nSv h(-1) and 8.7 +/- 0.6 nSv h(-1) were obtained for October 2008, respectively (quoted uncertainties represent standard deviations of 124 values obtained during the measurement period). Ambient dose equivalent rates measured by means of an extended rem counter at the Schneefernerhaus agree with those based on the BSS neutron energy distributions within 5 %. The ambient dose equivalent rate was also calculated based on simulated FLUKA neutron energy distributions in the atmosphere. Even without detailed modelling of the local environment, an agreement better than 30 % was obtained between the ambient dose equivalent rate based on the FLUKA distributions and those based on the BSS measurements at the Schneefernerhaus, for neutrons above about 20 MeV. This agreement is expected to be even better if the influence of the local environment on the measured neutron fluence energy distribution will be calculated.

A large amount of radioiodine isotopes (mainly (131)I, t(1/2) = 8 days) was released from the accident at Chernobyl Nuclear Power Plant (CNPP) in April-May 1986. An increase in childhood-thyroid cancer in the contaminated areas in Belarus, Russia and the Ukraine was demonstrated to be caused by radioiodine released at the time of the accident. However, there is a lack of quantitative data on the (131)I levels in the local environment (e.g. air, plant, soil). At this point, a long-lived iodine isotope, (129)I (t(1/2) = 15.7 million years), also released with a certain ratio to (131)I from CNPP, could be used for estimating the (131)I levels in the environment. In this paper we present analytical results of the (129)I concentrations and (129)I/(127)I atom ratios in soil samples collected from the CNPP exclusion zone (30-km zone), with the aim of assessing current contamination levels and distribution patterns. For the analysis of the iodine fraction in the investigated soil samples, the pyrohydrolysis method was utilized for separation of (127)I and (129)I nuclides, and subsequently their concentration was determined using inductively coupled plasma mass spectrometry (ICP-MS) and accelerator mass spectrometry (AMS), respectively. The concentration of (129)I and the (129)I/(127)I atom ratio in the surface soil samples in the 30 km-zone of CNPP ranged from 4.6 to 170 mBq/kg, and from 1.4 x 10(-6) to 13 x 10(-6), respectively. These values are significantly higher than those from global (129)I fallout, indicating that most of the measured (129)I was due to the deposition of the accident. Stable iodine concentrations in this area were found to be very low (below 1 ppm) for most of the samples, suggesting the environmental iodine levels in this area to be potentially low. The (129)I/(137)Cs activity ratio in surface and sub-surface soils was not so constant, i.e., in the range (7.3-20.2) x 10(-7). This might be due to the different behavior of deposition and/or migration of these nuclides in soil. These results suggest the obtained data of (129)I to be useful for the reconstruction of the (131)I deposition in the contaminated areas.

Air crew members and airline passengers are continuously exposed to cosmic radiation during their flights. Particles ejected by the sun during so-called solar particle events (SPEs) in periods of high solar activity can contribute to this exposure. In rare cases the dose from a single SPE might even exceed the annual dose limit of 1 mSv above which dose monitoring of air crews is legally required in Germany. Measurements performed by means of neutron monitors have already shown that the relative intensity of secondary neutrons from cosmic radiation is enhanced during an SPE, particularly at regions close to the magnetic poles of the Earth where shielding of the cosmic radiation by the geomagnetic field is low. Here we describe a Bonner sphere spectrometer installed at the Koldewey station at 79 degrees N, i.e. about 1,000 km from the geographic North pole, which is designed to provide first experimental data on the time-dependent energy spectrum of neutrons produced in the atmosphere during an SPE. This will be important to calculate doses from these neutrons to air crew members. The system is described in detail and first results are shown that were obtained during quiet periods of sun activity.

The first measurements of (63)Ni produced by A-bomb fast neutrons (above approximately 1 MeV) in copper samples from Hiroshima encompassed distances from approximately 380 to 5062 m from the hypocenter (the point on the ground directly under the bomb). They included the region of interest to survivor studies (approximately 900 to 1500 m) and provided the first direct validation of fast neutrons in that range. However, a significant measurement gap remained between the hypocenter and 380 m. Measurements close to the hypocenter are important as a high-value anchor for the slope of the curve for neutron activation as a function of distance. Here we report measurements of (63)Ni in copper samples from the historic Hiroshima Atomic Bomb Dome, which is located approximately 150 m from the hypocenter. These measurements extend the range of our previously published data for (63)Ni providing a more comprehensive and consistent A-bomb activation curve. The results are also in good agreement with calculations based on the current dosimetry system (DS02) and give further experimental support to the accuracy of this system that forms the basis for radiation risk estimates worldwide.

In the process of developing a new dosimetry system for atomic bomb survivors in Hiroshima and Nagasaki (DS02), an intercomparison study between (152)Eu and (36)Cl measurements was proposed, to reconcile the discrepancy previously observed in the Hiroshima data between measurements and calculations of thermal neutron activation products. Nine granite samples, exposed to the atomic-bomb radiation in Hiroshima within 1,200 m of the hypocenter, as well as mixed standard solutions containing known amounts of europium and chlorine that were neutron-activated by a (252)Cf source, were used for the intercomparison. Gamma-ray spectrometry for (152)Eu was carried out with ultra low-background Ge detectors at the Ogoya Underground Laboratory, Kanazawa University, while three laboratories participated in the (36)Cl measurement using accelerator mass spectrometry (AMS): The Technical University of Munich, Germany, the Lawrence Livermore National Laboratory, USA and the University of Tsukuba, Japan. Measured values for the mixed standard solutions showed good agreement among the participant laboratories. They also agreed well with activation calculations, using the neutron fluences monitored during the (252)Cf irradiation, and the corresponding activation cross-sections taken from the JENDL-3.3 library. The measured-to-calculated ratios obtained were 1.02 for (152)Eu and 0.91-1.02 for (36)Cl, respectively. Similarly, the results of the granite intercomparison indicated good agreement with the DS02 calculation for these samples. An average measured-to-calculated ratio of 0.98 was obtained for all granite intercomparison measurements. The so-called neutron discrepancy that was previously observed and that which included increasing measured-to-calculated ratios for thermal neutron activation products for increasing distances beyond 1,000 m from the hypocenter was not seen in the results of the intercomparison study. The previously claimed discrepancy could be explained by insufficient understanding of the measured data.