A. Lorente

Estimation of neutron-equivalent dose in organs of patients undergoing radiotherapy by the use of a novel online digital detector

Neutron peripheral contamination in patients undergoing high-energy photon radiotherapy is considered as a risk factor for secondary cancer induction. Organ-specific neutron-equivalent dose estimation is therefore essential for a reasonable assessment of these associated risks. This work aimed to develop a method to estimate neutron-equivalent doses in multiple organs of radiotherapy patients. The method involved the convolution, at 16 reference points in an anthropomorphic phantom, of the normalized Monte Carlo neutron fluence energy spectra with the kerma and energy-dependent radiation weighting factor. This was then scaled with the total neutron fluence measured with passive detectors, at the same reference points, in order to obtain the equivalent doses in organs. The latter were correlated with the readings of a neutron digital detector located inside the treatment room during phantom irradiation. This digital detector, designed and developed by our group, integrates the thermal neutron fluence. The correlation model, applied to the digital detector readings during patient irradiation, enables the online estimation of neutron-equivalent doses in organs. The model takes into account the specific irradiation site, the field parameters (energy, field size, angle incidence, etc) and the installation (linac and bunker geometry). This method, which is suitable for routine clinical use, will help to systematically generate the dosimetric data essential for the improvement of current risk-estimation models.

Testing of a High-Density Concrete as Neutron Shielding Material

Abstract We present the testing of a high-density magnetite concrete [commercially available under the name Hormirad™, developed by the Spanish company Construcciones Tecnicas de Radioterapia, S.L. (CT-RAD)] as neutron shielding material. The purpose of this work was to characterize the material behavior against neutrons, as well as to test different mixings including boron compounds in an effort to improve neutron shielding efficiency. Hormirad™ slabs of different thicknesses were exposed to a 241Am-Be neutron source under controlled conditions. The original mix, which includes a high fraction of magnetite, was then modified by adding different proportions of anhydrous borax (Na2B4O7). Looking for a comparison, the same experiment was repeated with slabs of ordinary concrete (HA-25) used to shield medical accelerator facilities. In parallel to the experiments, Monte Carlo calculations were performed with MCNP5, with some differences found with regard to the experiments, attributable to uncertainties in the elemental composition of the samples tested. Tenth-value layers have been determined for the different types of concrete tested for the 241Am-Be neutron source. The results show an advantageous behavior of the Hormirad™ when comparing it with ordinary concrete. Although borated concretes show a small improvement in neutron attenuation when they are compared with Hormirad™ alone, the resulting reduction in density and structural properties makes them less practical.

Evaluation of CdZnTe as neutron detector around medical accelerators

The operation of electron linear accelerators (LINACs) and cyclotrons can produce a mixed gamma-neutron field composed of energetic neutrons coming directly from the source and scattered lower energy neutrons. The thermal neutron detection properties of a non-moderated coplanar-grid CdZnTe (CZT) gamma-ray detector close to an 18 MV electron LINAC and an 18 MeV proton cyclotron producing the radioisotope (18)F for positron emission tomography are investigated. The two accelerators are operated at conditions producing similar thermal neutron fluence rates of the order of 10(4) cm(-2) s(-1) at the measurement locations. The counting efficiency of the CZT detector using the prompt 558 keV photopeak following (113)Cd thermal neutron capture is evaluated and a good neutron detection performance is found at the two installations.

On line neutron dose evaluation in patients under radiotherapy

Current improvements in radiotherapy require methods to evaluate their costs and benefits. A possible counterpart of the benefit is the creation of a secondary, radiation induced cancer. A new procedure is presented to assess the peripheral dose delivered to the patients due to photo-neutrons by means of a new on line digital detector. The events in the monitor have been correlated with the neutron dose by Monte Carlo simulations and experimental measurements using CR39 and TLD. This digital detector was employed at 6 different linacs, with energies ranging from 6 to 23 MV, obtaining the doses received in each organ of the patient. Additionally, the ambient dose equivalent was determined finding values from 0 to 470 mSv for complete treatments.

Experimental temperature measurements for the energy amplifier test

Abstract A uranium thermometer has been designed and built in order to make local power measurements in the First Energy Amplifier Test (FEAT). Due to the experimental conditions power measurements of tens to hundreds of nW were required, implying a sensitivity in the temperature change measurements of the order of 1 mK. A uranium thermometer accurate enough to match that sensitivity has been built. The thermometer is able to determine the absolute energetic gain obtained in a tiny subcritical uranium assembly exposed to a proton beam of kinetic energies between 600 MeV and 2.75 GeV. In addition, the thermometer measurements have provided information about the spatial power distribution and the shape of the neutron spallation cascade.

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 (241)AmBe 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.

RESPONSE MATRIX CALCULATION OF A BONNER SPHERE SPECTROMETER BASED ON A 6Lil(Eu) SCINTILLATOR

Abstract Using Monte Carlo methods the response matrix of a Bonner sphere spectrometer with a 6LiI scintillator has been calculated. The response functions were calculated for the bare detector and for polyethylene spheres 5.08, 7.62, 12.7, 20.32, 25.4, and 30.48 cm in diameter. Twenty-three beams of monoenergetic neutrons were used as sources in the energy interval from 0.025 eV to 100 MeV. The response functions were interpolated to energy points of those calculated in earlier literature works and compared with two response functions reported in the literature; good agreement was found from this comparison. The main differences were found for neutrons with energies higher than 20 MeV and, to a minor extent, for low-energy neutrons as well. These differences are mainly attributed to the cross-section libraries utilized in the different studies.

Induced radioisotopes in a linac treatment hall.

When linacs operate above 8MV an undesirable neutron field is produced whose spectrum has three main components: the direct spectrum due to those neutrons leaking out from the linac head, the scattered spectrum due to neutrons produced in the head that collides with the nuclei in the head losing energy and the third spectrum due to room-return effect. The third category of spectrum has mainly epithermal and thermal neutrons being constant at any location in the treatment hall. These neutrons induce activation in the linac components, the concrete walls and in the patient body. Here the induced radioisotopes have been identified in concrete samples located in the hall and in one of the wedges. The identification has been carried out using a gamma-ray spectrometer.

Dosimetric assessment and characterisation of the neutron field around a Howitzer container using a Bonner sphere spectrometer, Monte Carlo simulations and the NSDann and NSDUAZ unfolding codes

The Neutron Measurements Laboratory at the Nuclear Engineering Department of the Polytechnic University of Madrid consists of a bunker-like room and was built for neutron dosimetry research purposes and measurements. The facility includes a 74-GBq ²⁴¹Am-Be neutron source placed inside a neutron Howitzer container. The source can be moved to the irradiation or to the storage position. In this work, a Bonner sphere spectrometer (BSS) was used to measure the neutron fluence spectra with the source in both positions. Ambient dose equivalent rates, *(10), were measured using a calibrated neutron area monitor LB6411 (Berthold). The measured count rates were used as input to the NSDann and NSDUAZ unfolding programs to obtain the neutron fluence spectra and *(10). Monte Carlo (MC) simulation methods were used to model the system and to calculate the neutron fluence rate and the ambient dose equivalent rate at the measurement points. The comparison between NSDUAZ and NSDann resulted in relative deviations up to 6.87 % in the total neutron fluence rate and 7.18 % in *(10) values, despite the differences in the shape of the spectra obtained for the irradiation position. Comparing with the measured values, the *(10) values obtained with the unfolding programs exhibit a maximum relative deviation of 12.19 %. Taking into account the associated uncertainties, MC simulations seem to be in reasonable agreement with measurements. A maximum relative deviation of 15.65 % between computed and measured *(10) values was obtained. The computed count rates were applied to the unfolding programs to calculate the total neutron fluence rate and a maximum deviation of 12.83 % was obtained between the original values calculated by NSDann. A sensitivity test showed that the NSDann unfolding program is very sensitive to the uncertainties of the BSS count rates.

Monte Carlo-based method to determine the strength of a neutron source

The utilization of a gamma-ray spectrometer with a 7.62 Ø×7.62 cm NaI(Tl) detector, with a spherical moderator, has been studied with the aim to measure the neutron fluence rate and to determine the neutron source strength. Moderators with a large amount of hydrogen are able to slowdown and thermalize neutrons; once thermalized, there is a probability for thermal neutrons to be captured by hydrogen, producing 2.22 MeV gamma rays. The pulse-height spectrum collected in a multichannel analyzer shows a photopeak around 2.22 MeV whose net area is proportional to total neutron fluence rate and to the neutron source strength. The characteristics of this system were determined by a Monte Carlo study using the MCNP 4C code, where a detailed model of the NaI(Tl) was utilized. Spheres of diameters 3, 5, and 10 inch were used as moderators, and the response was calculated for monoenergetic and isotopic neutrons sources.