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cited By 5; Typical patient radiation doses in diagnostic imaging Radiologists should be able to appreciate the radiation dose delivered to patients for routine diagnostic procedures. The radiology report should include data necessary to calculate the patient dose in Gray. Using the effective dose, it is possible to compare with other source of radiation exposure. Simple formulas, taking into account different anatomical regions, derived from dose-area product (conventional radiography) or dose-length product (CT) are provided to calculate the effective dose in Sievert. For conventional (non-interventional) radiography, the effective dose for a given exam is inferior or equal to the yearly background radiation. For CT, the effective dose corresponds to 1 to 10 years of yearly background radiation. © 2010. Éditions Françaises de Radiologie.; Il est essentiel que tout radiologue soit capable d’apprécier le niveau de la dose délivrée par les explorations usuelles. La transcription sur le compte-rendu d’examen des éléments permettant de connaître la dose délivrée est obligatoire. À partir de ces données, il est possible par des calculs d’évaluer le niveau de dose et ainsi de connaître l’ordre de grandeur de l’irradiation exprimée en Gray ou sous multiples. La comparaison avec les autres sources d’exposition est possible par l’appréciation de la dose efficace qui tient compte de l’effet biologique attendu. Des formules simples tenant compte de la région anatomique sont proposées à partir des produits dose-surface (radiologie conventionnelle) et dose-longueur (TDM) pour évaluer cette dose efficace, qui s’exprime en sievert et sous-multiples (mSv). En radiographie conventionnelle (non interventionnelle), la dose efficace délivrée par une exploration est inférieure ou égale au niveau de l’irradiation naturelle annuelle. En tomodensitométrie, la dose efficace délivrée par une exploration représente, selon les régions anatomiques, de 1 à 10 ans d’irradiation naturelle.

cited By 16; The purpose of this study was to evaluate the in vivo dose-response relation of chromosome aberration formation and distribution in a context of localised and fractionated radiotherapy. Cytogenetic analysis was applied to eight patients, all treated for the same tumour localisation; the same localisation was used to prevent the variability usually observed between patients treated with radiotherapy and to allow the corresponding roles of the size of irradiation field and of the dose rate to be studied. The yield of dicentrics, centric rings and fragments was measured in blood samples taken before treatment, during the course of radiotherapy and up to 6 months after. After the first fraction of radiotherapy, we observed that the whole-body dose estimated from the yield of dicentrics and rings was higher (0.35±0.2 Gy) than the calculated equivalent whole-body dose (0.07±0.04 Gy). By contrast, the partial-body dose derived from the Qdr (quotient of dicentrics and rings) model was estimated to be 2.2±0.3 Gy, which agreed quite well with the dose delivered to the tumour (2.1±0.1 Gy). We also found a correlation between the yield of induced chromosome aberrations and the target field size (p=0.014). U-value analysis showed that the distribution of dicentrics and rings was overdispersed, despite the fractionation of the exposure, and a positive correlation between the U-value and the dose rate was observed (p=0.017). Overall, these results suggest that the proportion of undamaged lymphocytes could increase with the dose rate. © 2010 The British Institute of Radiology.

Purpose. To evaluate current exposure levels from pediatric MDCT examinations in order to issue recommendations for the routine clinical practice that may be used for establishing future diagnostic reference levels (DRL).Materials and methods. A survey was conducted at hospital sites affiliated with the SFIPP (Société Francophone d'Imagerie Pédiatrique et Prénatale). Tube potential and volume computed tomography dose index (CTDI) vol. values were obtained for three age groups (1, 5 and 10 years) for typical scanning indications of eight anatomical regions.Results. Twenty of 29 sites provided us with a copy of their protocols. All ages groups combined, 97% of protocols used a tension level ≤120 kVp. For age groups 1, 5 and 10 years respectively, the 75th percentiles of dose distributions were : 31, 39.5 and 49.5 mGy for the «head» protocol; 24, 22 and 24 mGy for «head and neck»; 11, 11, and 11mGy for «paranasal sinuses»; 46, 71 and 87 mGy for «petrous bone»; 3, 3.5 and 5.5 mGy for «chest»; 2, 3 and 4 mGy for «low dose lung»; 4, 4.5 and 7 mGy for «abdomen-pelvis»; 7, 11 and 12 mGy for «bone». The inter-center discrepancy for doses were still high.Conclusion. This study provided an evaluation of current practices in reference centers for pediatric imaging and allowed readjustment of MDCT dose recommendations. It could help in the development of DRLS for pediatric MDCT.; Objectif. Évaluer les niveaux actuels d’exposition en scanner multicoupes pédiatrique et en déduire des recommandations pour la pratique pouvant servir de base à l’élaboration de futurs niveaux de référence diagnostiques (NRD). Matériels et méthodes. Une enquête de pratique a été réalisée auprès des sites hospitaliers correspondants de la SFIPP (Société Francophone d’Imagerie Pédiatrique et Prénatale). Les valeurs protocolaires de haute tension et de dose théorique absorbée (IDSV) ont été relevées pour trois âges types (1, 5 et 10 ans) et des indications types sur huit régions anatomiques. Résultats. Vingt des 29 sites contactés ont transmis leurs protocoles. Tous âges confondus, 97 % des protocoles comportaient un niveau de haute tension ≤ 120 kV. Pour les âges de 1, 5 et 10 ans respectivement, les 75es centiles des distributions de dose étaient : 31, 39,5 et 49,5 mGy pour le protocole « crâne » ; 24, 22 et 24 mGy pour le « massif facial » ; 11, 11 et 11 mGy pour les « sinus » ; 46, 71 et 87 mGy pour les « rochers » ; 3, 3,5 et 5,5 mGy pour le « thorax standard » ; 2, 3 et 4 mGy pour le « poumon basse dose » ; 4, 4,5 et 7 mGy pour « l’abdomen-pelvis » ; 7, 11 et 12 mGy pour l’« os ». Les écarts dosimétriques intercentres restaient élevés. Conclusion. L’étude a permis d’évaluer les pratiques actuelles au sein de centres radiopédiatriques de référence et de réajuster les recommandations dosimétriques. Elle pourrait servir de base pour l’établissement de NRD en scanographie pédiatrique.

Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. This dose distribution can be assessed by physical dosimetric reconstruction methods. Physical dosimetric reconstruction can be achieved using experimental or numerical techniques. This article presents the laboratory-developed SESAME-Simulation of External Source Accident with MEdical images-tool specific to dosimetric reconstruction of radiological accidents through numerical simulations which combine voxel geometry and the radiation-material interaction MCNP(X) Monte Carlo computer code. The experimental validation of the tool using a photon field and its application to a radiological accident in Chile in December 2005 are also described. © 2009 Health Physics Society.

As part of the EU Fifth Framework Programme IDEAS project 'General Guidelines for the Evaluation of Incorporation Monitoring Data', and in collaboration with the International Atomic Energy Agency, a new intercomparison exercise for the assessment of doses from intakes of radionuclides was organised. Several cases were selected, to cover a wide range of practices in the nuclear fuel cycle and medical applications. The cases were: (1) acute intake of HTO, (2) acute inhalation of the fission products 137Cs and 90Sr, (3) acute inhalation of 60Co, (4) repeated intakes of 131I, (5) intake of enriched uranium and (6) single intake of Pu isotopes and 241Am. This intercomparison exercise especially focused on the effect of the Guidelines proposed by the IDEAS project for harmonisation of internal dosimetry. © The Author 2007. Published by Oxford University Press. All rights reserved.

A didactic software, MEthodes DOsimètriques de REférence (MEDOR), is being developed to provide help in the interpretation of biological data. Its main purpose is to evaluate the pertinence of the application of different models. This paper describes its first version that is focused on inhalation exposure to actinide aerosols. With this tool, sensitivity analysis on different parameters of the ICRP models can be easily done for aerosol deposition, in terms of activity and particle number, actinide biokinetics and doses. The user can analyse different inhalation cases showing either that dose per unit intake cannot be applied if the aerosol contains a low number of particles or that an inhibition of the late pulmonary clearance by particle transport can occur which contributes to a 3-4 fold increase in effective dose as compared with application of default parameters. This underlines the need to estimate systematically the number of deposited particles, as well as to do chest monitoring as long as possible. © The Author 2007. Published by Oxford University Press. All rights reserved.

The interpretation of individual activity measurement after a contamination by 241Am or its parent nuclide 241Pu is based on the reference americium (Am) biokinetic model published by the International Commission on Radiological Protection in 1993 [International Commission on Radiological Protection. Age-dependent doses to members of the public from intake of radionuclides: Part 2 Ingestion dose coefficients. ICRP Publication 67. Ann. ICRP 23(3/4) (1993)]. The authors analysed the new data about Am biokinetics reported afterwards to propose an update of the current model. The most interesting results, from the United States Transuranium and Uranium Registries post-mortem measurement database [Filipy, R. E. and Russel, J. J. The United States Transuranium and Uranium Registries as sources for actinide dosimetry and bioeffects. Radiat. Prot. Dosim. 105(1-4), 185-187 (2003)] and the long-term follow-up of cases of inhalation intake [Malatova, I., Foltanova, S., Beckova, V., Filgas, R., Pospisilova, H. and Holgye, Z. Assessment of occupational doses from internal contamination with 241Am. Radiat. Prot. Dosim. 105(1-4), 325-328 (2003)], seemed to show that the current model underestimates the retention in the massive soft tissues and overestimates the retention in the skeleton and the late urinary excretion. However, a critical review of the data demonstrated that all were not equally reliable and suggested that only a slight revision of the model, possibly involving a change in the balance of activity between massive soft tissues, cortical and trabecular bone surfaces, may be required. © The Author 2007. Published by Oxford University Press. All rights reserved.

The need for harmonisation of the procedures for internal dose assessment has been recognised within an EU research project under the 5th Framework Programme. The aim of the IDEAS project was to develop general guidelines for standardising assessments of intakes and internal doses. It started in October 2001 and ended in June 2005. The project is closely related to some goals of the work of Committee 2 of the International Commission on Radiological Protection and since 2003 there has been close co-operation between the two groups. The general philosophy of the guidelines is focusing on the principles of harmonisation, accuracy and proportionality. The proposed system of 'level of task' to structure the approach of internal dose evaluation is also reported. Some details of the internal structure of the guidelines for the different pathways of intake are provided. © The Author 2007. Published by Oxford University Press. All rights reserved.

The Institute for Radiological Protection and Nuclear Safety owns a graphite-moderated AmBe neutron field facility, SIGMA, that has to be reconstructed. Monte-Carlo simulations were performed to study the design of a new thermal facility based on IRSN existing facilities. Studies related to an update version of SIGMA concerned the enhancement of the thermal neutrons contribution to the dose equivalent. Calculations were mainly performed for a 252Cf neutron source distribution located at the centre of a graphite moderator block. A quasi-pure thermal neutron field was obtained with a 2.4 × 2.4 × 2.4-m3 block of graphite. A second acceptable neutron field was obtained with 3.3-MeV mono-energetic neutrons created by a 400-kV accelerator coupled to a graphite assembly of 1.5 × 1.5 × 1.5 m3. The characteristics of the studied thermal fields with the requirement for a reference calibration field are compared, and the advantages and drawbacks of the different producing methods are discussed. © The Author 2007. Published by Oxford University Press. All rights reserved.

The European Commission is funding within its Sixth Framework Programme a three-year project (2005-2007) called CONRAD, COordinated Network for RAdiation Dosimetry. The organisational framework for this project is provided by the European Radiation Dosimetry Group EURADOS. One task within the CONRAD project, Work Package 6 (WP6), was to provide a report outlining research needs and research activities within Europe to develop new and improved methods and techniques for the characterisation of complex radiation fields at workplaces around high-energy accelerators, but also at the next generation of thermonuclear fusion facilities. The paper provides an overview of the report, which will be available as CERN Yellow Report. © The Author 2007. Published by Oxford University Press. All rights reserved.

Lung cancer incidence in Sprague-Dawley rats was simulated by a biologically based carcinogenesis model, which is formulated mathematically in terms of a stochastic state-vector model. Doses to the sensitive target cells in the bronchial epithelium of the rat lung were calculated by a stochastic dosimetry model, considering the distinct monopodial branching structure and the crossfire of alpha particles from alveolar tissue to bronchial epithelium. Bronchial and alveolar cellular doses could reasonably be approximated by lognormal distributions, with geometric standard deviations (GSD) between 7 and 10, depending on exposure conditions. Based on a dose-exposure conversion factor of 8.5 mGy WLM-1 and a GSD of 8, lung cancer incidences were calculated for each cumulative exposure category in the rat inhalation study, consisting of different exposure rates and exposure times. The fair agreement between theoretical predictions and experimental data over the whole exposure range emphasises the necessity to incorporate the full cellular dose distributions rather than their mean values. © 2006 Oxford University Press.

EVIDOS is an EC sponsored project that aims at an evaluation and improvement of radiation protection dosimetry in mixed neutron/photon fields. This is performed through spectrometric and dosimetric investigations during different measurement campaigns in representative workplaces of the nuclear industry. The performance of routine and, in particular, novel personal dosemeters and survey instruments is tested in selected workplace fields. Reference values for the dose equivalent quantities, H*(10) and Hp(10) and the effective dose E, are determined using different spectrometers that provide the energy distribution of the neutron fluence and using newly developed devices that determine the energy and directional distribution of the neutron fluence. The EVIDOS project has passed the mid-term, and three measurement campaigns have been performed. This paper will give an overview and some new results from the third campaign that was held in Mol (Belgium), around the research reactor VENUS and in the MOX producing plant of Belgonucléaire. © 2006 Oxford University Press.

The QUADOS EU cost shared action conducted an intercomparison on the usage of numerical methods in radiation protection and dosimetry. The eight problems proposed were intended to test the usage of Monte Carlo and deterministic methods by assessing the accuracy with which the codes are applied and also the methods used to evaluate uncertainty in the answer gained through these methods. The overall objective was to spread good practice through the community and give users information on how to assess the uncertainties associated with their calculated results. © 2006 Oxford University Press.

The purpose of this work is to present an innovative approach for the creation and application of voxel phantoms associated with the Monte Carlo calculation (MCNP) for the calibration of whole-body counting systems dedicated to the measurement of fission and activation products. The new method is based on a graphical user interface called 'OEDIPE' that allows to simulate a whole measurement process using all measurement parameters, the final goal being to approach a numerical calibration of the facilities. The creation of voxel phantoms and validation of the method are presented in this paper using the IGOR phantom. Finally, the efficiency of the method is discussed, in particular, with the perspective of validating IGOR as a suitable human-equivalent phantom and for the assessment of uncertainties in dose estimation due to the inhomogeneous distribution of activity in the body, correlated to the bio-kinetic behaviour of the radionuclides. © The Author 2005. Published by Oxford University Press. All rights reserved.

The paper presents the OEDIPE (French acronym that stands for tool for personalised internal dose assessment) and SESAME (for simulation of external source accident with medical images) computational tools, dedicated to internal and external dose assessment, respectively, and currently being developed at the Institute for Radiological Protection and Nuclear Safety. The originality of OEDIPE and SESAME, by using voxel phantoms in association with Monte Carlo codes, lies in their ability to construct personalised voxel phantoms from medical images and automatically generate the Monte Carlo input file and visualise the expected results. OEDIPE simulates in vivo measurements to improve their calibration, and calculates the dose distribution taking both internal contamination and internal radiotherapy cases into account. SESAME enables radiological overexposure doses to be reconstructed, as also victim, source and accident environment modelling. The paper presents the principles on which these tools function and an overview of specificities and results linked to their fields of application. © The Author 2005. Published by Oxford University Press. All rights reserved.

A Monte Carlo code, initially developed for the calculation of microdosimetric spectra for alpha particles in cylindrical airways, has been extended to allow the computation (i) of additional microdosimetric parameters and (ii) for realistic exposure conditions in human bronchial airways with respect to surface activity distribution and airway geometry. The objective of the present study was to investigate the effects of non-uniform distributions of radon progeny activities in bronchial airways on cellular energy deposition parameters. Significant variations of hit frequencies, doses and microscopic energy deposition patterns were observed for epithelial cell nuclei, depending strongly on the assumed activity distributions. Thus, epithelial cells located at different positions in a given bronchial airway may experience a wide range of biological responses. The results obtained suggest that the hit frequency may be the primary physical parameter for alpha particles, supplemented by microdosimetric single event spectra, to be related to biological effects for chronic low level exposures. © The Author 2005. Published by Oxford University Press. All rights reserved.

In some Spanish nuclear power plants of pressurised water reactor (PWR) type, albedo thermoluminescence dosemeters are used for personal dosimetry while survey meters, based on a thermal-neutron detector inside a cylindrical or spherical moderator, are used for dose rate assessment in routine monitoring. The response of both systems is highly dependent on the energy of the existing neutron fields. They are usually calibrated by means of ISO neutron sources with energy distributions quite different from those encountered at these installations. Spectrometric measurements with a Bonner sphere system (BSS) allow us to determine the reference dosimetric values. The UAB group, under request from the National Coordinated Research Action, was in charge of characterising the neutron fields and evaluating the response of personal dosemeters at several measurement points inside the containment building of the Catalan Nuclear Power Plant Vandellòs II. The neutron fields were characterised at five places using the UAB-BSS and a home made unfolding code called MITOM. The results obtained confirm the presence of low-energy components in the neutron field in most of the selected points. Moreover, we have found no influence of the nuclear fuel burning on the shape of the spectrum. © Oxford University Press 2004; all rights reserved.

Individual irradiation doses were determined by electron paramagnetic resonance spectroscopy of the tooth enamel of the inhabitants of Zaborie, the most contaminated inhabited settlement not evacuated after the Chernobyl accident. Dose determination was performed using a specially developed automatic spectrum processing procedure. Spectrum processing was carried out in different operating modes, and average results were taken in order to reduce the contribution of uncertainty in dose determination caused by spectrum processing. The absorbed doses determined in enamel were corrected to take into account the contribution of natural background radiation and to determine the individual excess dose due to radioactive contamination of the territory. Individual excess doses are compared to calculated individualized doses to teeth, estimated using the local radioactive contamination levels, dose rates, and information concerning individual behavior. The individual excess doses measured by electron paramagnetic resonance spectroscopy and the calculated individualized doses are fully independent. Mean square variation between results of two methods was found to be 34 mGy, which is consistent with error estimation for both methods. This result can validate both the methodology of signal processing presented here when using electron paramagnetic resonance dosimetry of tooth enamel for low doses and the methodology of individualized dose calculation.