[Show abstract][Hide abstract] ABSTRACT: It is generally known that the use of high-energy electron linear accelerators (LINACs) in radiotherapy medical treatments may generate secondary neutrons, mainly via photonuclear (gamma, n ) giant dipole resonance reactions of incident photons with all the heavy materials present inside the gantry and along the beam line. A detailed knowledge (i.e., fluence energy distribution) of this parasite radiation, which is approximately isotropic and not confined within the primary LINAC beam field, would be of great interest to estimate the associated radiological risk for the patient and the working staff. It has been shown, in this study, that our recently developed passive Bonner sphere system, using pure gold activation foils as central detectors, is well adapted to measure neutron spectra at pulsed and intense mixed n-gamma fields with high-energy photon component. This system was used to characterize the neutron field around a new generation medical electron LINAC. Two measurement positions (isocenter and maze entrance) inside the treatment room of this facility, with the machine operating in Bremsstrahlung photon mode, were chosen. The obtained specific <sup>198</sup>Au saturation activities were processed by means of the NUBAY unfolding code, which performs a Bayesian estimation of a parameterized spectrum, to derive the final neutron spectra. Another unfolding method (MAXED), based on the maximum entropy principle and which may depend to some extent on considered the initial guess or default spectrum, was also applied to check the robustness of the NUBAY solutions as well as to carry out a sensitivity analysis to confirm their stability and to corroborate the associated uncertainties on their output results. Also presented are the obtained integral quantities, in terms of neutron fluence and ambient dose equivalent rates normalized to the primary LINAC photon dose, together with an estimation of their associated uncertainties due to the unfolding code used.
IEEE Transactions on Nuclear Science 11/2009; · 1.22 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: The ROSPEC device is a multi-detector system, which has been designed by Bubble Technologies Industries (BTI at Chalk River, ON, Canada) to assess neutron spectra, and hence neutron dose quantities, at workplace fields. It is made up of six gaseous proportional counters that detect neutrons via the elastic (n,p) scattering (four hydrogenous counters) and with the (3)He(n,p)T reaction (two (3)He-filled counters). Results of the calibration of a similar rotating spectrometer (ROSPEC) have been described by Rosenstock et al.((1)). For energy and fluence calibration purposes, measurements were performed with the accelerator for metrology and neutron applications in external dosimetry (AMANDE) facility at the Laboratory of Neutron Metrology and Dosimetry (Institute of Radiation Protection and Nuclear Safety, IRSN, France). This facility provides monoenergetic neutron radiation fields from 2 keV to 20 MeV. Two kinds of experiments were carried out. First, the ROSPEC was used in its rotational mode for the ISO energies. Then, each detector was irradiated with all the available neutron energies, in a well defined position with the rotation of the device stopped. The energy values of the neutron beam were calculated using the TARGET code. A BC501-A liquid scintillation spectrometer provided the fluence values for energies beyond 1.2 MeV, a methane-filled SP2 counter from 800 keV to 1.4 MeV and an H(2)-filled SP2 counter from 144 to 800 keV. Reference data for 70 keV monoenergetic neutrons were obtained using the IRSN Long Counter. Results showed that the ROSPEC device was in agreement with the absolute neutron fluences within 10%. Moreover, the new energy calibration factors are in good agreement with those derived by BTI.
[Show abstract][Hide abstract] ABSTRACT: This paper presents a measurement campaign carried out in six medical electron linear accelerator facilities to evaluate occupational exposure associated with photons and secondary neutrons for several encountered configurations. No strong correlation between dose and configuration has been observed. Assuming realistic hypotheses, the annual effective dose to staff due to external exposure has been estimated to 0.9 mSv, 54% associated with the decay of activation products and 7% due to exposure to leakage neutrons.
[Show abstract][Hide abstract] ABSTRACT: Proton recoil spectra were calculated for various spherical proportional counters using Monte Carlo simulation combined with the finite element method. Electric field lines and strength were calculated by defining an appropriate mesh and solving the Laplace equation with the associated boundary conditions, taking into account the geometry of every counter. Thus, different regions were defined in the counter with various coefficients for the energy deposition in the Monte Carlo transport code MCNPX. Results from the calculations are in good agreement with measurements for three different gas pressures at various neutron energies.
Nuclear Instruments and Methods in Physics Research Section A Accelerators Spectrometers Detectors and Associated Equipment 01/2008; 593(3):485-491. · 1.14 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: The dosimetry of the experimental reactor CALIBAN was established for photon and neutron components, and the kerma variations were evaluated according to position (distance from the core, height, angle with the median room axis). Dosimetry was performed with TLD (Al2O3) for the photon component, and with passive silicon diodes, activation detectors (Au, In, Ni, Mg, Cu) and alanine pellets measured by EPR spectrometry for the neutron component. The neutron energy distribution was experimentally evaluated at various distances based on the activities measured at the activated foil using the SNAC2 software. It was then compared with those calculated with the Monte Carlo code MCNPX.
[Show abstract][Hide abstract] ABSTRACT: The Institute for Radiological protection and Nuclear Safety was engaged in the EC funded EVIDOS project to provide reference spectrometry data using its Bonner sphere system. The data were processed by means of two unfolding codes, NUBAY and GRAVEL, both provided by the Physikalisch-Technische Bundesanstalt. The NUBAY program, based on Bayesian parameter estimation methods, assumes a parameterised spectrum and provides posterior probability distributions for the parameters. The code GRAVEL, an iterative algorithm based on SAND-II, was used with various default spectra, among them the NUBAY solution. The BS measurements were used to establish the neutron fluence energy distributions and reference values for the neutron ambient dose equivalent. As this quantity depends strongly on the high energy neutrons, a sensitivity analysis was done by unfolding the BS data with GRAVEL using the NUBAY solution spectrum as default with various changes in the parameters of the high energy peak. This new method of analysing Bonner sphere data allowed the determination of reliable neutron spectra, as well as a very good estimate of the corresponding integral quantities with small associated uncertainties.
[Show abstract][Hide abstract] ABSTRACT: At the Krsko Nuclear Power Plant (NPP), albedo dosimeters are used for personal neutron dosimetry. Spectrometric measurements allow determination of reference dosimetric values of realistic neutron fields to be used for calibration of albedo dosimeters. The Laboratory for Neutron Metrology and Dosimetry from the Institute for Radiological Protection and Nuclear Safety (IRSN) was in charge of characterising neutron fields in the plant at two representative points with high neutron and gamma dose rate. Calibration of the dosimeters in the workplace used to be performed only by a spherical survey meter. Based on the reference dosimetric values, the Plant Dosimetry Laboratory has verified the response of albedo dosimeters.
[Show abstract][Hide abstract] ABSTRACT: The photon contribution to ambient dose equivalent in several wide-spectrum reference neutrons fields of the Institute for Radiological Protection and Nuclear Safety were measured using a Geiger-Müller counter. For the investigated fields, the ratio of photon to neutron ambient dose equivalent ranged between 0.03 and 0.20. The results show that the Geiger-Müller tube is a versatile instrument for dosimetry in mixed photon-neutron fields if sufficient information for the calculation of corrections is available.
[Show abstract][Hide abstract] ABSTRACT: The neutron spectrometer ROSPEC is made up of 6 spherical proportional counters with different gas fillings, on a rotating platform, and which covers the energy range 0.025 eV to 4.5 MeV. A complete study is performed within the framework of a French working group in order to improve the spectrometer performance. First, the study of the devices performance in reference neutron fields was performed, with respect to the influence of rotation and disturbance from other counters. It appeared that the fluence scattered by other counters could represent between 7 and 40% of the total fluence, but with a slight influence on the initial spectrum. Secondly, we determined the response matrix for each counter from two main parameters, the wall and the gas amplification effects. The wall effects were calculated using the Snidow algorithm and compared to the MCNPX simulation. The change in the electric field was calculated using a finite element method to define amplification region for simulation. Finally, since the ROSPEC unfolding code sometimes gave negative fluences, the unfolding code UMG was used to correct this default. We compared the results from both unfolding codes.
[Show abstract][Hide abstract] ABSTRACT: Expertise and research activities in the field of dosimetry at the “Institut de radioprotection et de sûreté nucléaire” (IRSN) have been initiated at CEA during the 60’s. The first laboratories studies have been focused to the development of radiation dosimetry techniques of workers. Today, external and internal dosimetry laboratories of the institute are involved in all subjects concerning doses measurements and follow-up for workers in normal situation or dose reconstruction following accidental situation. In 1984, the creation of the LARD came with the impulsion of the head of department Portal and of Blanc with the goal to ease the contacts between the different laboratories and especially to help them to know better European Community in order to integrate research programs granted by the commission. After a brief historical introduction, three fields of activity currently studied by IRSN are presented, dosimetry reconstruction following accidental external irradiation, dosimetric and spectrometric characterization at work places exposed to neutron radiation and research made for the improvement of in vivo monitoring in internal dosimetry. Les activités de recherche et d’expertise en dosimétrie de l’Institut de radioprotection et de sûreté nucléaire (IRSN) sont les héritières de celles commencées dans les années soixante au CEA. Les premiers laboratoires créés dans les années soixante étaient destinés à développer les techniques dosimétriques des rayonnements externes pour la protection du personnel. Aujourd’hui, les laboratoires de dosimétrie externe et interne de l’institut couvrent assez largement les différentes problématiques liées à la dosimétrie des rayonnements ionisants, que ce soit pour la mesure et le suivi des doses en situation normale d’exploitation ou pour la reconstitution des doses à la suite d’une situation accidentelle. En 1984, la création des LARD est intervenue sous l’impulsion du chef du service de dosimétrie Portal et de Blanc dans le but de favoriser les contacts entre les divers laboratoires et notamment de les aider à pénétrer les arcanes de la commission européenne pour s’intégrer aux programmes de recherche soutenus par la commission. Après un bref rappel historique, trois des champs d’activité actuellement traités par l’IRSN sont présentés, la reconstitution dosimétrique suite à une irradiation externe accidentelle, la caractérisation dosimétrique et spectrométrique des postes de travail exposés aux neutrons et la recherche pour l’amélioration de la mesure directe en dosimétrie interne.
[Show abstract][Hide abstract] ABSTRACT: The French laboratories in charge of 'neutron' dosimetry using the spectrometer 'ROSPEC', formed a working group in 2001. The participants began to study the behaviour of the instrument with a comparison exercise in broad energy neutron fields recommended by the International Organisation for Standardisation (ISO) and available at the LMDN in Cadarache. The complete version of the ROSPEC is made up of six spherical proportional counters fixed to a rotating platform. These counters cover different energy ranges which overlap each other to provide a link between the detectors, within the energy range from thermal neutrons to 4.5 MeV. The irradiation configurations chosen were ISO standard sources (252Cf, (252Cf+D2O)(/Cd), 241Am-Be) and the SIGMA facility. The results show that the 'thermal and epithermal' neutron fluence was widely overestimated by the spectrometer in all configurations.
[Show abstract][Hide abstract] ABSTRACT: An international intercomparison of criticality accident dosimetry systems took place in the SILENE reactor, in June 2002. Participants from 60 laboratories irradiated their dosemeters (physical and biological) using two different configurations of the reactor. In preparation for this intercomparison, the leakage radiation fields were characterised by spectrometry and dosimetry measurements using the ROSPEC spectrometer associated with a NE-213 scintillator, ionisation chambers, GM counters, diodes and thermoluminescence dosemeters (TLDs). For this intercomparison, a large area was required to irradiate the dosemeters both in free air and on phantoms. Therefore, measurements of the uniformity of the field were performed with activation detectors and TLDs for neutron and gammas, respectively. This paper describes the procedures used and the results obtained.
[Show abstract][Hide abstract] ABSTRACT: In criticality accident dosimetry and more generally for high dose measurements, special techniques are used to measure separately the gamma ray and neutron components of the dose. To improve these techniques and to check their dosimetry systems (physical and/or biological), a total of 60 laboratories from 29 countries (America, Europe, Asia) participated in an international intercomparaison, which took place in France from 9 to 21 June 2002, at the SILENE reactor in Valduc and at a pure gamma source in Fontenay-aux-Roses. This intercomparison was jointly organised by the IRSN and the CEA with the help of the NEA/OCDE and was partly supported by the European Communities. This paper describes the aim of this intercomparison, the techniques used by the participants and the two radiation sources and their characteristics. The experimental arrangements of the dosemeters for the irradiations in free air or on phantoms are given. Then the dosimetric quantities measured and reported by the participants are summarised, analysed and compared with the reference values. The present paper concerns only the physical dosimetry and essentially experiments performed on the SILENE facility. The results obtained with the biological dosimetry are published in two other papers of this issue.
[Show abstract][Hide abstract] ABSTRACT: In the context of accidental or intentional radiation exposures (nuclear terrorism), it is essential to separate rapidly those individuals with substantial exposures from those with exposures that do not constitute an immediate threat to health. Low-frequency electron paramagnetic resonance (EPR) spectroscopy provides the potential advantage of making accurate and sensitive measurements of absorbed radiation dose in teeth without removing the teeth from the potential victims. Up to now, most studies focused on the dose-response curves obtained for gamma radiation. In radiation accidents, however, the contribution of neutrons to the total radiation dose should not be neglected. To determine how neutrons contribute to the apparent dose estimated by EPR dosimetry, extracted whole human teeth were irradiated at the SILENE reactor in a mixed neutron and gamma-radiation field simulating criticality accidents. The teeth were irradiated in free air as well as in a paraffin head phantom. Lead screens were also used to eliminate to a large extent the contribution of the gamma radiation to the dose received by the teeth. The EPR signals, obtained with a low-frequency (1.2 GHz) spectrometer, were compared to dosimetry measurements at the same location. The contribution of neutrons to the EPR dosimetric signal was negligible in the range of 0 to 10 Gy and was rather small (neutron/gamma-ray sensitivity in the range 0-0.2) at higher doses. This indicates that the method essentially provides information on the dose received from the gamma-ray component of the radiation.
Radiation Research 09/2003; 160(2):168-73. · 2.70 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: Five scintillation detectors of different scintillator size and type were characterized. The pulse height scale was calibrated in terms of electron light output units using photon sources. The response functions for time-of-flight (TOF)-selected monoenergetic neutrons were experimentally determined and also simulated with the NRESP code over a wide energy range. A comparison of the measured and calculated response functions allows individual characteristics of the detectors to be determined and the response matrix to be reliably derived. Various applications are discussed.
Nuclear Instruments and Methods in Physics Research Section A Accelerators Spectrometers Detectors and Associated Equipment 01/2002; 476(s 1–2):186–189. · 1.14 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: The new ICPR60 recommendations and the consideration of the ALARA principle have led the operators of nuclear facilities to evaluate with a higher care, the doses received by workers. The aim of this paper is to present a recent study concerning mixed field characterisation at a workplace located in a reprocessing laboratory.As a first step, neutron spectrum determination was achieved by two ways: simulation using MCNP code and experimental measurements with Bonner spheres and recoil proton counters. Neutron spectrum allowed the evaluation of dosimetric quantities. Measurements were then performed with different devices routinely used in radioprotection. The authors describe the measurement techniques, present the results obtained, and finally compare and discuss them.
Nuclear Instruments and Methods in Physics Research Section A Accelerators Spectrometers Detectors and Associated Equipment 01/2002; 476(1):440-445. · 1.14 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: The Institute for Protection and Nuclear Safety (IPSN) standard neutron detector in the energy range 60–800 keV is a spherical proportional counter of HARWELL type SP2 nominally filled with 300 kPa hydrogen. It was characterised in the monoenergetic neutron fields of PTB at the energies of 144, 250 and 565 keV, where the neutron energy and fluence were determined with the PTB reference instruments. The neutron fields produced at the same energies with the accelerator facility of Bruyères-le-Châtel were then investigated with the calibrated SP2 counter and various PTB instruments in order to determine the mean energy and the neutron fluence. The energy scale and a neutron fluence monitor were calibrated.
Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment. 01/2002;