Ambient Dose Equivalent Research Articles

Field calibration of PADC track etch detectors for local neutron dosimetry in man using different radiation qualities

In order to quantify the dose from neutrons to a patient for contemporary radiation treatment techniques, measurements inside phantoms, representing the patient, are necessary. Published reports on neutron dose measurements cover measurements performed free in air or on the surface of phantoms and the doses are expressed in terms of personal dose equivalent or ambient dose equivalent. This study focuses on measurements of local neutron doses inside a radiotherapy phantom and presents a field calibration procedure for PADC track etch detectors. An initial absolute calibration factor in terms of Hp(10) for personal dosimetry is converted into neutron dose equivalent and additional calibration factors are derived to account for the spectral changes in the neutron fluence for different radiation therapy beam qualities and depths in the phantom. The neutron spectra used for the calculation of the calibration factors are determined in different depths by Monte Carlo simulations for the investigated radiation qualities. These spectra are used together with the energy dependent response function of the PADC detectors to account for the spectral changes in the neutron fluence. The resulting total calibration factors are 0.76 for a photon beam (in- and out-of-field), 1.00 (in-field) and 0.84 (out-of-field) for an active proton beam and 1.05 (in-field) and 0.91 (out-of-field) for a passive proton beam, respectively. The uncertainty for neutron dose measurements using this field calibration method is less than 40%.The extended calibration procedure presented in this work showed that it is possible to use PADC track etch detectors for measurements of local neutron dose equivalent inside anthropomorphic phantoms by accounting for spectral changes in the neutron fluence.

Intercomparison of the response of different photon and neutron detectors around a spent fuel cask

An intercomparison of the response of different photon and neutron detectors was performed in several measurement positions around a spent fuel cask (type TN 12/2B) filled with 4 MOX and 8 UO 2 15 × 15 PWR fuel assemblies at the nuclear power plant Gösgen (KKG) in Switzerland. The instruments used in the study were both active and passive, photon and neutron detectors calibrated either for ambient or personal dose equivalent. The aim of the measurement campaign was to compare the responses of the radiation instruments to routinely used detectors. It has been shown that especially the indications of the neutron detectors are strongly dependent on the neutron spectra around the cask due to their different energy responses. However, routinely used active photon and neutron detectors were shown to be reliable instruments. ► Response of different detectors around a spent fuel cask is compared. ► We use active and passive, photon and neutron detectors. ► We measure on two positions the neutron spectrum with Bonner spheres. ► Indications of neutron detectors are strongly dependent on neutron spectra. ► Routinely used detectors were shown to be reliable instruments.

Calculation of conversion coefficients for air kerma to ambient dose equivalent using transmittedspectra of megavoltage X-rays through concrete

With the fast advancement of technology, (60)Co teletherapy units are largely being replaced with medical linear accelerators. In most cases, the linear accelerator tends to be installed in the same room in which the (60)Co teletherapy unit was previously placed. If in-depth structural remodelling is out of the question, high-density concrete is usually used to improve shielding against primary, scatter and leakage radiation originating in the new equipment. This work presents a study based on Monte Carlo simulations of the transmission of some clinical photon spectra (from 6, 10, 15, 18 and 25 MV accelerators) through concrete, considering two different densities. Concrete walls with thickness ranging from 0.70 to 2.0 m were irradiated with 30 cm×30 cm primary beam spectra. The results show that the thickness of the barrier decreases up to ∼65 % when barite (high-density concrete) is used instead of ordinary concrete. The average energies of primary and transmitted beam spectra were also calculated. In addition, conversion coefficients from air kerma to ambient dose equivalent, H*(d)/K(air), and air kerma to effective dose, E/K(air), for photon spectra from the transmitted spectra were calculated and compared. The results suggest that the 10-mm depth is not the best choice to represent the effective dose.

Neutron features at the UPM neutronics hall

The neutronics hall of the Nuclear Engineering Department at the Polytechnical University of Madrid has been characterized. The neutron spectra and the ambient dose equivalent produced by an 241AmBe source were measured at various source-to-detector distances on the new bench. Using Monte Carlo methods a detailed model of the neutronics hall was designed, and neutron spectra and the ambient dose equivalent were calculated at the same locations where measurements were carried out. A good agreement between measured and calculated values was found.

Ferramenta computacional para avaliação de kerma no ar em aplicações de radioproteção em áreas de internação de pacientes: proposição de um método simples para avaliação experimental

OBJETIVO: Apresentar uma ferramenta de análise de dados que pode ser utilizada para proteção de pacientes e trabalhadores em áreas de uso de equipamentos móveis. MATERIAIS E MÉTODOS: Foi desenvolvida uma ferramenta, em planilha ativa Excel®, que utiliza medidas de exposição para gerar um banco de dados de fatores de forma e calcular o kerma no ar ao entorno de um leito. O banco de dados inicial foi coletado com três equipamentos móveis. Um espalhador não antropomórfico foi utilizado, sendo realizadas medidas de exposição em uma malha de (4,2 × 4,2) m², ao passo de 0,3 m. RESULTADOS: A ferramenta calcula o kerma no ar (associado à exposição de pacientes expostos e ao equivalente de dose ambiente) à radiação secundária. Para distâncias inferiores a 60,0 cm, valores acima do limite máximo de equivalente de dose ambiente definido para área livre (0,5 mSv/ano) foram verificados. Os dados coletados a 2,1 m foram sempre inferiores a 12% do referido limite. CONCLUSÃO: A ferramenta é capaz de auxiliar na proteção radiológica de pacientes e trabalhadores, quando associada à coleta de dados adequada, pois possibilita a determinação de áreas livres ao entorno de leitos em áreas onde equipamentos móveis geradores de radiação X são utilizados.

Measurement of cosmic-ray neutron dose onboard a polar route flight from New York to Seoul

Exposure to cosmic radiation in operation of a jet aircraft is considered to be a part of the occupational exposure. Cosmic radiation doses received in aviation are generally evaluated by numerical model simulations. The precision of the model calculation should be verified by measurements. From the viewpoint of radiological protection, neutrons are the most contributing radiation component and have to be precisely measured. Neutron measurements were thus performed in a long-haul flight using a relatively new transportable neutron monitor (WENDI-II) which responds fairly well to the cosmic-ray neutrons. The in-flight measurement was carried out on 5-6 November 2009 on a polar route flight from New York/John F. Kennedy airport to Seoul/Incheon airport. The flying time was ~14 h. The observations obtained as 1 cm ambient dose equivalent were compared with model calculations using a computer program developed by the authors for the calculation of aviation route doses 'JISCARD EX'. Good agreements between the measured and calculated values were observed over the polar route where the geomagnetic cut-off rigidity is the lowest.

Field test of wide-range environmental neutron dosemeter using PADC

the cyclotron facility of National Institute of Radiological Sciences (NIRS) in Japan.G The neutron ambient dose equivalent was measured using the dosemeter and a rem counter. The neutron energy spectrum and the ambient dose equivalent in this neutron field were also evaluated by the measurement of the multi-moderator spectrometer, so-called Bonner ball, inserting gold activation foil.

Comparison of operator radiation exposure during coronary sinus catheter placement via the femoral or jugular vein approach

To evaluate and compare operator radiation exposure during the catheter placement in the coronary sinus via the femoral vein with a steerable catheter or the jugular vein with a fixed curve catheter. A total of 156 patients undergoing electrophysiological study or radiofrequency catheter ablation were prospectively assigned in a random fashion to either the femoral vein access (FVA) with a steerable curve deca-polar catheter (n= 80) or the jugular vein access (JVA) with a fixed curve deca-polar catheter (n = 76). All the catheterization procedures were performed by the same operator who had extensive experience in both accesses. Operator radiation exposure was measured with an electronic radiation dosimeter attached to the breast pocket of the operator on the outside of the lead apron and estimates of the ambient dose equivalent were derived. The operator radiation exposure was reduced significantly by using the FVA compared with the JVA (1.8 ± 1.3 vs. 8.6 ± 6.5 μSv; P < 0.001). The fluoroscopy time (62.7 ± 45.8 vs. 61.9 ± 46.5 s; P = NS) and dose-area product (3.2 ± 2.3 vs. 3.1 ± 2.1 Gy cm(2); P = NS) were not statistically different. Operator radiation exposure can be significantly reduced by using the FVA approach with a steerable curve catheter compared with the JVA approach with a fixed curve catheter, without increasing the fluoroscopy time and dose-area product.

Optimal shielding design for bunkers of compact cyclotrons used in the production of medical radionuclides

There are several options to consider in the design of a vault that will house a cyclotron for radioisotopes production with regards to the door entrance. Alternatives are a direct-shielded door, a simple maze, or a double-legged maze. In this work, the impact of the neutron and photon doses at the vault entrance was evaluated for these options. Monte Carlo simulations were carried out in order to assess photon and neutron ambient dose equivalents. Simulations results were compared to experimental measurements taken inside a vault with a direct-shielded door. The double-legged maze is the configuration that provides the higher degree of radiological protection at the vault entrance. In addition, the fact of the location of the cyclotron target plays an important role in the ambient dose equivalents. The comparison performed between measurements and results of MCNP simulations confirmed a favorable agreement; maze legs reduce the neutron energies at the entrances of the vault. However, a degree of dose reduction similar to the one obtained for a two-legged maze can be achieved with the addition of shielding against neutrons in the inner maze entrance to act as a second door. A choice of a vault design is more evident by comparing the results of this study.

Evaluation of H*(10) using the developed spherical type neutron dose monitor

An instrument for evaluating the neutron ambient dose equivalent has been developed. It has the characteristic of uniform response to wide energy of neutrons. The monitor is four-layered spherically shaped, based on moderation and absorption of neutrons. Neutron dose can be evaluated from the linear combination of three specific responses of thermoluminescent dosimeters (TLDs), which are located at three depths in the moderator. TLDs were arranged between layers of two consecutive depths on 12 radial axes at even intervals so that the monitor is equally sensitive to all directions of neutrons. In order to verify the usefulness of dose evaluation by the monitor, irradiation experiments were conducted at the FRS, JAEA. The D2O-moderated 252Cf was used for the calibration of the monitor. Experiments were also conducted by using two neutron sources of 252Cf bare and 241Am-Be. As a result, the evaluated dose for each irradiation was obtained close to the actual irradiated dose. It was confirmed that the method of dose evaluation by the developed monitor can be applied to practical neutron fields where the distance of neutron source is unknown.

The simulated workplace field with a high-energy neutron component produced by irradiating a Fe-target with 200 MeV/u 12C-ions

Abstract Recent developments in accelerator physics have led to new challenges for radiation protection dosimetry. Doses have to be determined for workplace fields which are characterized by high-energy radiation, a dominant contribution from neutrons, high intensities and pulsed time structure This may present problems for active measuring devices. As is well known, the ambient dose equivalent is often underestimated by area monitors operating in high-energy neutron fields behind shielding. Therefore, it is desirable to calibrate survey monitors in a characterized neutron field with the type of spectral fluence distribution that is expected behind shielding, i.e. where the main dose from neutrons arises from two peaks with mean energies of about 1 MeV and 100 MeV, respectively. Such a neutron fluence distribution is produced by the irradiation of a Fe-target with 200 MeV/u 12C-ions. Measurements with the extended range Bonner sphere spectrometer NEMUS of PTB were performed at two positions inside the experimental area Cave A of the heavy-ion synchrotron SIS at GSI. The measured neutron spectra show different fluence contributions for the two peaks at the two positions. The results were compared to Monte Carlo Simulations with MCNPX and FLUKA.

Influence of Bonner sphere response functions above 20 MeV on unfolded neutron spectra and doses

Abstract Monte Carlo (MC) codes for neutron transport calculations such as MCNP, MCNPX, FLUKA, PHITS, and GEANT4, crucially rely on cross sections that describe the interaction of neutrons with nuclei. For neutron energies below 20 MeV, evaluated cross sections are available that are validated against experimental data. In contrast, for high energies (above 20 MeV) experimental data are scarce and, for this reason, every neutron transport code is based on theoretical nuclear models to describe interactions of neutrons with nuclei in matter. Here we report on the calculation of a complete set of response functions for a Bonner spheres spectrometer (BSS), by means of GEANT4 using the Bertini and Binary Intranuclear Cascade (INC) models for energies above 20 MeV. The recent results were compared with those calculated by MCNP/LAHET and MCNP/HADRON MC codes. It turns out that, whatever code used, the response functions were rather similar for neutron energies below 20 MeV, for all 16 detector/moderator combinations of the considered BSS system. For higher energies, however, differences of more than a factor of 2 were observed, depending on neutron energy, detector/moderator combination, MC code, and nuclear model used. These differences are discussed in terms of neutron fluence rates measured at the environmental research station (UFS), “Schneefernerhaus”, (Zugspitze mountain, Germany, 2650 m a.s.l.) for energies below 0.4 eV (thermal neutrons), between 0.4 eV and 100 keV (epithermal neutrons), between 100 keV and 20 MeV (evaporation neutrons), and above 20 MeV (cascade neutrons). In terms of total neutron fluence rates, relative differences of up to 4% were obtained when compared to the standard MCNP/LAHET results, while in terms of total ambient dose equivalent, relative differences of up to 8% were obtained. Both the GEANT4 Binary INC and Bertini INC gave somewhat larger fluence and dose rates in the epithermal region. Most relevant for dose, however, those response functions calculated with the GEANT4 Bertini INC model provided about 18% less neutrons in the cascade region, which led to a roughly 13% smaller contribution of these neutrons to ambient dose equivalent. It is concluded that doses from secondary neutrons from cosmic radiation as deduced from BSS measurements are uncertain by about 10%, simply because of some differences in nuclear models used by various neutron transport codes.

Evaluation of neutron production in new accelerators for radiotherapy

Ambient dose equivalent, H*(10), and personal dose equivalent, Hp(10), were calculated in different points located inside two different treatment rooms. 15-MV Varian and 15-MV Elekta accelerators were used in these studies. The geometry of both accelerators heads and treatment rooms were built up to perform the Monte Carlo simulations. The patient was also simulated using an ICRU phantom. Calculations were done using the MCNPX code. Ambient dose equivalents rates from neutrons range between 1.2 and 419 mSv/h in the Elekta treatment room and between 0.96 and 1140 mSv/h in the Varian treatment room, depending on the location. These values suggest a larger neutron production in the Varian than in the Elekta accelerator.

Comparison of measured and calculated in-air secondary neutrons in passive carbon-ion radiotherapy

Undesired radiation exposure in normal tissues around a treatment volume in proton and carbon-ion radiotherapies is less than that in the conventional radiotherapies due to physical and/or biological properties of charged particles. Such exposure is always considered in a treatment planning, however, undesired exposure in normal tissues far from the treatment volume cannot be considered in the treatment planning, because it is caused by secondary radiation as well as leakage primary particles. Though this exposure is considerably lower than that near the treatment volume, it may be not negligible to estimate the risk of secondary cancer especially for the young patients. In particular, the assessment of the secondary neutrons that inevitably produced within the patient and beam line devices is very important due to the potency of their biological effect. The distributions of the absorbed dose and the biological effectiveness in phantom/patient are required to assess the risk, and Monte Carlo calculation plays a key role due to a difficulty of the measurements. In this study, comparison of measured and calculated in-air neutrons at the patient position in the Heavy Ion Medical Accelerator in Chiba (HIMAC) treatment room are performed to verify the accuracy of the Monte Carlo code, PHITS. Our calculations underestimated epithermal neutrons measured by Bonner sphere system. This discrepancy may be caused by an insufficiency of the calculational geometry modeling, consequently an underestimation of neutrons scattered and moderated by the beam line devices. However, it is unlikely that the underestimation significantly contribute to the dose estimation in phantom. On the other hand, the calculation reproduced the measured ambient dose equivalents well because they were dominated by neutrons above 0.1 MeV. This result showed that the PHITS code has a potential ability to evaluate the neutron exposure of the patient in passive carbon-ion radiotherapy.

Microdosimetric measurements in the secondary radiation field produced in 12C-therapy irradiations

The ambient dose equivalent from the secondary radiation produced during irradiation of a cylindrical water phantom with 200 MeV/u (12)C-ions was investigated at the biophysics cave at GSI Helmholtzzentrum für Schwerionenforschung in Darmstadt, Germany. Pencil-like ion beams were delivered by the heavy-ion synchrotron SIS18 using the slow extraction mode. Since the secondary radiation field outside the phantom is complex in its particle composition and particle energy distribution, microdosimetric methods developed for the dosimetry of the cosmic radiation field at flight altitudes, which is similar in terms of complexity, were applied. Lineal energy distributions and the ambient dose equivalent were measured with a tissue-equivalent proportional counter at different particle emission angles. An additional veto counter allowed the identification of the different contributions of charged and neutral particles. A significant increase in the mean quality factor was observed at large emission angles which could be attributed to the decreasing contributions of charged particles compared to the (relative) contributions from neutrons.

Criteria for establishing shielding of multi-detector computed tomography (MDCT) rooms

The aim of this work is to compare two methods used for determining the proper shielding of computed tomography (CT) rooms while considering recent technological advances in CT scanners. The approaches of the German Institute for Standardisation and the US National Council on Radiation Protection and Measurements were compared and a series of radiation measurements were performed in several CT rooms at the Lausanne University Hospital. The following three-step procedure is proposed for assuring sufficient shielding of rooms hosting new CT units with spiral mode acquisition and various X-ray beam collimation widths: (1) calculate the ambient equivalent dose for a representative average weekly dose length product at the position where shielding is required; (2) from the maximum permissible weekly dose at the location of interest, calculate the transmission factor F that must be taken to ensure proper shielding and (3) convert the transmission factor into a thickness of lead shielding. A similar approach could be adopted to use when designing shielding for fluoroscopy rooms, where the basic quantity would be the dose area product instead of the load of current (milliampere-minute).

ＣＲ‐３９を用いた中性子周辺線量当量測定用線量計の開発

A CR-39 has good advantages such as cumulative type dosimeter, small fading effect and gamma-ray insensitive. Therefore, we developed the wide energy-range environmental neutron dosimeter using eight CR-39s for area monitoring in this study. This dosimeter is made of octagonal columnar polyethylene block which height is 60 mm and bottom side is 25 mm. The dosimeter contains two types of CR-39s for fast neutron detection and slow neutron detection. Four CR-39s for fast neutron detection are used for detection of recoil protons produced by H (n, p) reactions. Four CR-39s for slow neutron detection are used with boron nitride converter to detect alpha-rays produced by 10B (n, α) 7Li reactions. Ambient dose equivalent is obtained by adding the number of etch-pits observed in four CR-39s for fast neutron detection to the number of etch-pits observed in four CR-39s for slow neutron detection with appropriate constants respectively. Dosimeters were irradiated with some energetic neutrons and evaluated results of ambient dose equivalent were compared with results from neutron transport calculations. Energy response of dosimeter shows good agreement with neutron fluence to ambient dose equivalent conversion coefficients. Directional dependence of dosimeter is at the same level as the rem-counter.

Comparison of Unfolding Methods for Determining Neutron Spectrum and Ambient Dose Equivalent

The neutron spectra and ambient dose equivalent were determined by unfolding measured Bonner sphere system data using different unfolding methods. These methods included a maximum entropy method (MAXED code), nonlinear least-squares method (GRAVEL code) with several different starting spectra, and a genetic algorithm method. These algorithms were used to unfold measured Bonner sphere data that had been collected using a LiI(Eu) detector and activation foils. The Bonner sphere system was exposed to neutrons from a known AmBe source and at the proton accelerator facility at the University of Texas M. D. Anderson Cancer Center. The total fluence rates and total ambient dose equivalents of the neutron field agree within 7.5%, regardless of unfolding algorithm or starting spectrum. In contrast, the fluence-weighted average energy varied dramatically, depending on the starting spectrum used in the unfolding process. These findings offer insight and guidance into the use of unfolding algorithms and starting spectra for neutron spectroscopy.

TEPC Measurements and Monte Carlo Calculations for Evaluating Ambient Dose Equivalent Response in Mixed Radiation Fields around the Shielded Area of a Carbon Ion Accelerator

The tissue equivalent proportional counter is a versatile instrument used to determine the absorbed dose and the quality of an unknown radiation field, but it does not directly measure the operational quantity ambient dose equivalent H*(10). In this paper, simulated and measured values of H*(10) are compared, and some corrections are applied in order to give a reliable estimate of H*(10) in a complex radiation field generated by high-energy carbon ions.

Measurement of Gamma and Neutron Dose from the Operation of the ANKA Storage Ring

ANKA is a ramped 2.5-GeV electron synchrotron light source, which can be operated at all energies from 500 MeV to 2.5 GeV. An online area monitoring network makes it possible to measure the radiation from the decaying beam at eight stations every minute. It consists of two stationary and six mobile stations with a gamma and a neutron detector, which are connected with a central computer by a bus system. Measuring devices and software were developed or customized for ANKA.Special machine shifts were dedicated to the systematic investigation of beam lifetime and ambient equivalent dose, which depend on the beam energy, beam current, and operating conditions of the storage ring.The highest dose rates are measured in the forward direction of insertion devices. The beam is lost at the device with the smallest aperture. Local shielding of the storage ring and operating conditions can be improved with time-resolved dose rate measurements at locations with high dose levels. Low beta optics improves the lifetime of electron beams circulating in vacuum chambers with small aperture. This helps to reduce the dose rate. Furthermore, this has the advantage of reducing the dose rate without the drawback of extra costs for additional shielding.