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In the framework of the ORAMED project (Optimization of RAdiation protection for MEDical staff), funded by the European Union Seventh Framework Programme, different studies were aimed at improving the quality of radiation protection in interventional radiology and nuclear medicine. The main results of the project were presented during a final workshop held in Barcelona in January 2011, the proceedings of which are available in the open literature. One of the ORAMED tasks was focused on the problem of eye-lens photon exposure of the medical staff, a topic that gained more importance especially after the ICRP decision to lower the limiting equivalent dose to 20 mSv per year. The present technical note has the scope, besides briefly summarising the physical reasons of the proposal and the practical implications, to provide, in tabular form, a set of air kerma to Hp(3) conversion coefficients based on the adoption of a theoretical cylindrical model that is well suited for reproduction of the mass and the shape of a human head. © The Author 2012. Published by Oxford University Press. All rights reserved.

In the framework of the ORAMED project (Optimization of RAdiation protection for MEDical staff), funded by the European Union Seventh Framework Programme, different studies were aimed at improving the quality of radiation protection in interventional radiology and nuclear medicine. The main results of the project were presented during a final workshop held in Barcelona in January 2011, the proceedings of which are available in the open literature. One of the ORAMED tasks was focused on the problem of eye-lens photon exposure of the medical staff, a topic that gained more importance especially after the ICRP decision to lower the limiting equivalent dose to 20 mSv per year. The present technical note has the scope, besides briefly summarising the physical reasons of the proposal and the practical implications, to provide, in tabular form, a set of air kerma to Hp(3) conversion coefficients based on the adoption of a theoretical cylindrical model that is well suited for reproduction of the mass and the shape of a human head.

This communication summarizes the results concerning the Monte Carlo (MC) modelling of Germanium detectors for the measurement of low energy photons arising from the International comparison on MC modelling for in vivo measurement of Americium in a knee phantom organized within the EU Coordination Action CONRAD (Coordinated Network for Radiation Dosimetry) as a joint initiative of EURADOS working groups 6 (computational dosimetry) and 7 (internal dosimetry). MC simulations proved to be an applicable way to obtain the calibration factor that needs to be used for in vivo measurements.

In recent years voxel models have been used to investigate the validity of radiation protection quantities evaluated, till now, employing stylized reference models. An ICRP task group (DOCAL) is developing a male and female reference voxel models to estimate a new set of radiation protection quantities. It should be taken into account that, due the intrinsic subject-dependent anatomical differences, the various developed voxel models can imply a certain variability in the corresponding sets of estimated conversion coefficients. Therefore a qualitative evaluation of the range of such discrepancies should be of interest in terms of the sensitivity of organ absorbed dose to the specific model used for its computation. The NORMAN-05 model, updated at ENEA-Radiation Protection Institute on the basis of the original NORMAN model, developed by P. Dimbylow at NRPB-UK (now HPA) in the 1990s, is a good representation of the standard Caucasian-type man and could be a valuable tool for comparison with other voxel models in order to state the inherent uncertainties associated to the estimated sets of conversion coefficients. The present paper summarizes the results obtained with NORMAN-05 and the Monte Carlo code MCNPX for neutron conversion coefficients calculation, in the energy domain from thermal to 20 MeV, in the kerma approximation that, for such energy range, as stated in the ICRU/ICRP recommendations, is a satisfactory approach.

To design a new radon gas dosemeter, Monte Carlo simulation has been used to determine the response of the device, fitted with a CR-39 SSNTD, to the different geometric conditions associated with the progress of the holder. The a counting efficiency of a SSNTD depends primarily on the energy and incidence angle into the detector, then on the etching conditions and finally on the track counting procedure. Therefore the elicited efficiency function represents all these factors. The detectors were exposed to a particles at different angle incidences and energies, by using a 244Cm source; two alternative NaOH solution etching conditions were tested, respectively 6 h and 12 h, at 70 °C and 6,5 N. The counting was performed with a light microscope, 267 X, and with an automatic track analysis procedure.

A new radon gas dosemeter (patent pending) has been designed and tested to be used as a routine referenced device to monitor radon exposures in workplaces and dwelling. It has been conceived as a practical tool to be effectively used in the practice of a qualified Service, operating under Q.A. Monte Carlo simulation has been extensively used to optimize the geometry of the sensible volume of the holder; the diffusion of the sole radon gas is achieved by the special assembly of the holder, which behaves as closed type with air gap. In this paper we introduce, besides a general description of the device, the theoretical results carried out by Monte Carlo simulation, about the sensitivity tothe radon gas and the homogeneity of the detector track density.The experimental calibrations were executed at the two ENEA facilities (the National Metrology Institute and the radon Casaccia Laboratory) and at NRPB radon chamber.

At the ENEA Institute for Radiation Protection (IRP) the fast neutron calibration facility consists of a remote control device which allows locating different sources (Am–Be, Pu–Li, bare and D2O moderated 252 Cf) at the reference position, at the desired height from the floor, inside a 10 � 10 � 3m 3 irradiation room. Either the ISO reference sources or the Pu–Li source have been characterised in terms of uncollided H � ð10Þ and neutron fluence according to the ISO calibration procedures. A spectral fluence mapping, carried out with the Monte Carlo Code MCNP TM , allowed characterising the calibration point, in scattered field conditions, according to the most recent international recommendations. Moreover, the irradiation of personal dosemeters on the ISO water filled slab phantom was simulated to determine the field homogeneity of the calibration area and the variability of the neutron field (including the backscattered component) along the phantom surface. At the ENEA Institute for Radiation Protection the calibration of neutron area monitors as well as personal dosemeters can now be performed according to the international standards, at the same time guaranteeing suitable conditions for research and qualification purposes in the field of neutron dosimetry. r 2002 Elsevier Science B.V. All rights reserved.

The new recommendation issued by the International Commission on Radiological Protection (ICRP) introducing a 20-mSv annual dose limit for the eye lens stimulated an interesting debate among the radiation protection community. In the present work the problem of estimating Hp(3) for neutron realistic workplace spectra is treated, employing the recently published Hp(3)/F conversion coefficients with the aim of establishing a workplace-dependent relationship between Hp(10) and Hp(3). The results demonstrate that, whilst the two quantities can differ by less than 10 %, in general, Hp(10) cannot be considered a conservative estimate of Hp(3).

The recent statement on tissue reactions issued by the International Commission on Radiological Protection in April 2011 recommends a very significant reduction in the equivalent dose annual limit for the eye lens from 150 to 20 mSv y-1; this has stimulated a lot of interest in eye lens dosimetry in the radiation protection community. Until now no conversion coefficients were available for the operational quantity Hp(3) for neutrons. The scope of the present work was to extend previous evaluations of Hz.ast;(10) and Hp(10) performed at the PTB in 1995 to provide also Hp(3) data for neutrons. The present work is also intended to complete the studies carried out on photons during the last 4 y within the European Union-funded ORAMED (optimisation of radiation protection for medical staff) project. © The Author 2013. Published by Oxford University Press. All rights reserved.

Background The authors report the results of femoral–tibial fusion with an Ilizarov circular external fixator following septic loosening of knee prosthesis. Materials and methods The series included 17 patients with a mean age of 62.9 years, treated from 1990 to 2007 with femoral–tibial fusion. The Cierny–Mader classification was used for clinical and anatomopathological evaluation; the Engh classification was used to assess the bone defect. Surgical treatment differed according to these criteria. Results Healing was achieved in 13 out of 17 patients at the first surgical attempt in a mean time of 9.3 months. Mean follow-up was 30 months. Of the four complications, two patients had an intolerance to the external fixator that led to its early removal, and the other two had a septic intrarticular nonunion. Conclusions The Ilizarov circular external fixator is a very reliable fixation system due to its low cost, versatility, stability under load, and low risk of septic dissemination. Nevertheless, an appropriate patient selection and a good surgeon’s experience are necessary.

Photons conversion coefficients from 20 keV to 100 MeV have been calculated with the voxel model NORMAN-05 using the MCNPX code. Both kerma approximation and electronic transport were employed and the results compared with published data. In the near future, ICRP group DOCAL will issue a new computational model, based on the GSF-GOLEM, which is intended to be the reference adult male voxel model for ICRP. NORMAN-05 well approximates the western-caucasian standard man characteristics, in terms of body height (176 cm) and mass (73 kg) and masses of the organs. It is not intended to substitute, or to be an alternative, to the future official ICRP voxel model, but thanks to its accuracy and its "standard man structure" can be useful to evaluate the intrinsic uncertainties associated with the dose quantities evaluated adopting different voxel models. For such reason data obtained with NORMAN-05 could be easily compared with those that will be derived from the future ICRP model.

The question of Monte Carlo simulation of radiation transport in voxel geometries is addressed. Patched versions of the MCNP and MCNPX codes are developed aimed at transporting radiation both in the standard geometry mode and in the voxel geometry treatment. The patched code reads an unformatted FORTRAN file derived from DICOM format data and uses special subroutines to handle voxel-to-voxel radiation transport. The various phases of the development of the methodology are discussed together with the new input options. Examples are given of employment of the code in internal and external dosimetry and comparisons with results from other groups are reported.

In recent years voxel phantoms have been developed on the basis of tomographic data of real individuals allowing new sets of conversion coefficients to be calculated for effective dose. Progress in radiation studies brought ICRP to revise its recommendations and a new report, already circulated in draft form, is expected to change the actual effective dose evaluation method. In the present paper the voxel phantom NORMAN developed at HPA, formerly NRPB, was employed with MCNP Monte Carlo code. A modified version of the phantom, NORMAN-05, was developed to take into account the new set of tissues and weighting factors proposed in the cited ICRP draft. Air kerma to organ equivalent dose and effective dose conversion coefficients for antero-posterior and postero-anterior parallel photon beam irradiations, from 20 keV to 10 MeV, have been calculated and compared with data obtained in other laboratories using different numerical phantoms. Obtained results are in good agreement with published data with some differences for the effective dose calculated employing the proposed new tissue weighting factors set in comparison with previous evaluations based on the ICRP 60 report.

The present paper summarises the activity carried out at the ENEA Radiation Protection Institute for updating the methodologies employed for the evaluation of the neutron and photon dose to the exposed workers in case of a criticality accident, in the framework of the 'International Intercomparison of Criticality Accident Dosimetry Systems' (SileÃne reactor, IRSN-CEA-Valduc June 2002). The evaluation of the neutron spectra and the neutron dosimetric quantities relies on activation detectors and on unfolding algorithms. Thermoluminescenct detectors are employed for the gamma dose measurement. The work is aimed at accurately characterising the measurement system and, at the same time, testing the algorithms. Useful spectral information were included, based on Monte Carlo simulations, to take into account the potential accident scenarios of practical interest. All along this exercise intercomparison a particular attention was devoted to the 'traceability' of all the experimental and computational parameters and therefore, aimed at an easy treatment by the user.

Radiation transport problems that involve high attenuation require non-analogue Monte Carlo. Non-analogue techniques (i.e. employing variance reduction (V.R.) ), conserve the first moment whilst reducing the second and consequently the variance. However, because V.R. methods employ in one form or another an estimate of the importance (adjoint flux) which is strictly linked to a particular response, they are directed at a single response. V.R. parameters appropriate to one response (in that they significantly reduce the statistical error) may not necessarily be appropriate to another; they may even raise the errors of other responses compared with their analogue values. Thus Monte Carlo may treat problems involving a high attenuation only if a single response (or a set of responses with similar importances) is of interest. Monte Carlo tends to run into trouble when more differential information is required (flux distributions in many energy groups for example). The technique presented here to optimize V.R. parameters to more than one response is based on the DSA (Direct Statistical Approach) which is of general application in that it is based on V.R. techniques that are themselves of general use: population or weight control through splitting and Russian roulette.

In the present report a thermal neutron assembly for calibrating dosimetric instruments at the ENEA Radiation Protection Institute is described. The characterization of such a facility in terms of spectral neutron fluence rate and ambient dose equivalent rate according to the ICRP60 document is shown in detail. The developed study was aimed at characterizing all the components of the field inside the three calibration cavities placed within a polyethylene cube. Besides from the experimental point of view, the assembly was also numerically investigated by means of the Monte Carlo code MCNP. These techniques allowed to calculate, besides the dosimetric quantities to be compared with the experimental results, also the detailed neutron spectra along the cavities. Special variance reduction algorithms developed at ENEA allowed to obtain a satisfactory statistics for each energy group of the spectra.

A new version of the MCNP Monte Carlo Code has been written that transports electrons, besides neutrons and photons, incorporating the EGS4 electron modules. The code, which is currently being benchmarked, is intended for use in the field of radiation protection and dosimetry.

The need to characterise the photon radiation field in the presence of a suitable phantom like the ICRU sphere for the design of personal dosemeters motivated an extensive amount of MCNP Monte Carlo calculations. Air kerma backscatter factors, mean energy of the backscattered radiation relative to angle of the incident beam as well as the backscattered fluence spectra for ISO and BIPM incident reference radiations were calculated for energies between 7 keV and 250 keV and for 137 Cs and 60 Co. Backscatter factor values for different angles of the incident beam were also experimentally evaluated and compared with the corresponding data obtained by Monte Carlo calculations