Dosimetry and fast neutron energies characterization of photoneutrons produced in some medical linear accelerators

This work focusses on the estimation of induced photoneutrons energy, fluence, and strength using nuclear track detector (NTD) (CR-39). Photoneutron energy was estimated for three different linear accelerators, LINACs as an example for the commonly used accelerators. For high-energy linear accelerators, neutrons are produced as a consequence of photonuclear reactions in the target nuclei, accelerator head, field-flattening filters and beam collimators, and other irradiated objects. NTD (CR-39) is used to evaluate energy and fluence of the fast neutron. Track length is used to estimate fast photoneutrons energy for linear accelerators (Elekta 10 MV, Elekta 15 MV, and Varian 15 MV). Results show that the estimated neutron energies for the three chosen examples of LINACs reveals neutron energies in the range of 1–2 MeV for 10 and 15 MV X-ray beams. The fluence of neutrons at the isocenter (Φ_total) is found to be (4×10⁶ n cm² Gy^?1) for Elekta machine 10 MV. The neutron source strengths Q are calculated. It was found to be 0.2×10¹² n Gy^?1 X-ray at the isocenter. This work represents simple, low cost, and accurate methods of measuring fast neutrons dose and energies.     

Boron thin films and CR-39 detectors in BNCT: A method to measure the 10B(n,α)7Li reaction rate

The working principle of the Boron Neutron Capture Therapy (BNCT) is the selective delivery of a greater amount of boron to the tumor cells than to the healthy ones, followed by the neutron irradiation that will induce the emission of α-particles and recoil ⁷Li nuclei through the ¹⁰B(n,α)⁷Li reaction. The objective of this work is to present a setup composed of a boron thin film coupled with CR-39. Alpha and ⁷Li particle coming from the boron films are used to quantify neutron boron reaction and are detected by CR-39. The nuclei compounding of this detector, H, C and O, will undergo fast neutrons reactions, which will be detected in the CR-39 itself. In this way, the ¹⁰B(n,α)⁷Li reaction and the contribution of fast neutrons to the flux can be determined at the same time. These measurements are essential for treatment planning as well as for studies of the biodistribution of ¹⁰B-carrier drugs and tissue microdosimetry. The boron films were deposited on stainless steel substrates through the sputtering technique and irradiated with thermal neutrons at the reactor IEA-R1 located at IPEN, São Paulo/SP, Brazil. Here we show the first results on the characterization of these thin films and calibration of the proposed setup.     

Comparative Study of Depth Dose Distribution of Fast Neutrons Using Solid State Nuclear Track Detectors

A comparative study was made for measuring neutron depth dose distribution in a water phantom by means of different solid state nuclear track detectors. The used detectors were Makrofol, CR-39 and LR-115 with threshold energies of 1 and 0.1 MeV, respectively. Two exposure positions were used, the source in contact with the surface and at 30 cm from the phantom surface. The fluence distribution of ²⁵²Cf as a function of distance in air and in the water phantom were investigated. It was found that the neutron build up factor depends on the irradiation geometry, its values were 1.7 and 1.4 for both exposure positions. These results are in good agreement with the calculated results given in literature. The neutron attenuation and relaxation length had been also determined.     

Neutron response study of two CR-39 personal dosemeters with air and Nylon converters

A neutron personal dosemeter using CR-39 as a detector and hydrogenated materials as proton converters for fast neutron detection plus an air layer for thermal neutron detection is being developed in our laboratory. To increase the CR-39 response to thermal neutrons, the air converter was substituted with Nylon in some dosemeters. Several dosemeters with these two configurations were mounted on a water-filled phantom and exposed under different incidence angles (0, 30 and 60) to: (i) Three ISO neutron sources (²⁴¹Am–Be, bare ²⁵²Cf and moderated ²⁵²Cf with Cd shielding), and (ii) two realistic neutron sources (SIGMA and moderated ²⁵²Cf) at the IRSN (Cadarache) facilities. The irradiated detectors were electrochemically etched and evaluated in order to determine their dose equivalent response in terms of H_p(10,α). The results obtained are compared to those obtained from Monte Carlo simulations using the MCNPX code.     

Measurements of neutron radiation and induced radioactivity for the new medical linear accelerator,the Varian TrueBeam

Contemporary linear accelerators applied in radiotherapy generate X-ray and electron beams with energies up to 20 MeV. Such high-energy therapeutic beams induce undesirable photonuclear (γ,n) and electronuclear (e,e'n) reactions in which neutrons and radioisotopes are produced. The originated neutron can also induce reactions such as simple capture, (n,γ), reactions that produce radioisotopes. In this work measurements of the non-therapeutic neutrons and the induced gamma radiation were carried out in the vicinity of a new medical accelerator, namely the Varian TrueBeam. The TrueBeam is a new generation Varian medical linac making it possible to generate the X-ray beams with a dose rate higher than in the case of the previous models by Varian. This work was performed for the X-ray beams with nominal potentials of 10 MV (flattening filter free), 15 MV and 20 MV, and for a 22 MeV electron beam. The neutron measurements were performed by means of a helium chamber and the induced activity method. The identification of radioisotopes produced during emission of the therapeutic beams was based on measurements of the energy spectra of gammas emitted in decays of the produced nuclei. The gamma energy spectra were measured with the use of the high-purity germanium detector. The correlation between the neutron field and the mode and nominal potential was observed. The strongest neutron fluence of 3.1 × 10⁶ cm⁻² Gy⁻¹ and 2.0 × 10⁶ cm⁻² Gy⁻¹ for the thermal and resonance energies, respectively, was measured during emission of the 20 MV X-ray beam. The thermal and resonance neutron fluence measured for the 15 MV X-rays was somewhat less, at 1.1 × 10⁶ cm⁻² Gy⁻¹ for thermal neutrons and 6.7 × 10⁵ cm⁻² Gy⁻¹ for resonance neutrons. The thermal and resonance neutron fluences were smallest for the 10 MV FFF beam and the 22 MeV electron beam and were around two orders of magnitude smaller than those of the 20 MV X-ray beam. This work has shown that the neutron reactions are dominant because of relatively high cross sections for many elements used in the accelerator construction. The detailed analysis of the measured spectra made it possible to identify 11 radioisotopes induced during TrueBeam delivery. In this work the following radioisotopes were identified: ⁵⁶Mn, ¹²²Sb, ¹²⁴Sb, ¹³¹Ba, ⁸²Br, ⁵⁷Ni, ⁵⁷Co, ⁵¹Cr, ¹⁸⁷W, ²⁴Na and ³⁸Cl.     

Effects of radiations on the characteristics of alpha and fission tracks in CR-39 detectors

Abstract

The effects of neutron, gamma and alpha radiations on the alpha and fission fragment tracks registration and revelation properties of CR-39 detectors (CR-39 and CR-39(DOP) were studied. It was found that the ratio of the bulk etch rate of irradiated to unirradiated (V_G(irr.)/V_G(unirr.) detectors is linearly dependent on dose. An exponential decrease in fission track densities with increase in neutron fluence was observed. The ratio of V_G(irr.)/V_G(unirr.) was found to be high in CR-39 than that in CR-39(DOP) exposed to the same reactor neutron fluence. The decrease in fission track densities with increase in neutron fluence was observed to be faster in CR-39 than in CR-39(DOP). This indicates that doping with dioctyl phthalate improves the radiation resistance of CR-39 detectors. It was observed that in detectors exposed to an alpha flux of the order of 9.36 × 10⁶ / cm², the fission track density was reduced by 11% and thereafter it remained constant. The results also indicate that thermal neutron fluence up to 7.01 ×10¹¹ neutrons/cm² does not affect the alpha and fission track densities. I.R. spectra were also studied to find out the nature of chemical changes produced by these radiations on CR-39.     

The dose comparison between the THOR and HFR epithermal neutron beams

This study is a part of the beam comparison campaign, inter-center dose comparison, between boron neutron capture therapy facilities at the Tsing Hua Open-pool Reactor and the High Flux Reactor. The clinical information exchange can improve the dosimetry uncertainty for medical physics in a mixed field. The method of paired Mg(Ar) and TE(TE) ionization chambers was used to determine the gamma-ray and neutron dose rates. Furthermore, activation foils, including gold, copper, and manganese, were employed to estimate the thermal and epithermal neutron fluxes. Measurements were performed free in air and also in a PMMA phantom. All the chambers were calibrated using a ⁶⁰Co primary standard source at the Institute of Nuclear Energy Research, Taiwan. Spectrum dependent neutron sensitivity of TE(TE) chamber is one of the important parameters to evaluate dose components. The requested neutron spectra were calculated by the Monte Carlo code MCNP. The measured thermal neutron fluxes, gamma-ray and neutron dose rates of the THOR beam in the phantom were 2.6, 2.2, and 2.1 times of the HFR beam at 2.5-cm depth, respectively. The higher thermal neutron flux and neutron and gamma-ray dose rates are due to the higher epithermal neutron beam intensity of the THOR.     

The use of recombination chambers at radiation therapy facilities

The paper presents an overview of the applications of recombination chambers for dosimetric measurements at radiotherapy facilities. The chambers were used at electron, proton and heavy ion accelerators, in the beam and in the vicinity of the accelerators at very different dose rates. The examples of measurements discussed in the paper include: the determination of the absorbed dose and radiation quality parameters of a 170 MeV proton beam and BNCT (boron neutron capture therapy) beam, neutron dose measurements at a phantom surface outside the beam of a 15 MV electron medical accelerator, determination of ambient dose equivalent, H1 (10) outside the irradiated phantom in the proton therapy treatment room at JINR (Dubna, Russia), and at working places outside the shielding of the heavy ion therapy facility at GSI (Darmstadt, Germany).     

In-phantom neutron dose distribution for bladder cancer cases treated with high-energy photons

This work presents an estimation of the neutron dose distribution for common bladder cancer cases treated with high-energy photons of 15 MV therapy accelerators. Neutron doses were measured in an Alderson phantom, using TLD 700 and 600 thermoluminescence dosimeters, resembling bladder cancer cases treated with high-energy photons from 15 MV LINAC and having a treatment plan using the four-field pelvic box technique. Thermal neutron dose distribution in the target area and the surrounding tissue was estimated. The sensitivity of all detectors for both gamma and neutrons was estimated and used for correction of the TL reading. TLD detectors were irradiated with a Co⁶⁰ gamma standard source and thermal neutrons at the irradiation facility of the National Institute for Standards (NIS). The TL to dose conversion factor was estimated in terms of both Co⁶⁰ neutron equivalent dose and thermal neutron dose. The dose distribution of photo-neutrons throughout each target was estimated and presented in three-dimensional charts and isodose curves. The distribution was found to be non-isotropic through the target. It varied from a minimum of 0.23 mSv/h to a maximum of 2.07 mSv/h at 6 cm off-axis. The mean neutron dose equivalent was found to be 0.63 mSv/h, which agrees with other published literature. The estimated average neutron equivalent to the bladder per administered therapeutic dose was found to be 0.39 mSv Gy^?1, which is also in good agreement with published literature. As a consequence of a complete therapeutic treatment of 50 Gy high-energy photons at 15 MV, the total thermal neutron equivalent dose to the abdomen was found to be about 0.012 Sv.     

硼中子俘获治疗的蒙特卡罗方法模拟

用通用蒙特卡罗程序MCNP模拟了粒子在人脑中的输运过程. 吸收剂量率主要来自以下四个反应:¹⁰B(n,α)⁷Li,¹⁴N(n,p)¹⁴C,¹H(n,γ)²D,快中子弹性散射反应.对肿瘤区的贡献主要来自硼中子吸收反应.结果表明,超热中子比热中子适合于深肿瘤的治疗,而热中子对浅肿瘤的治疗有优越性,比如皮肤癌.同确定论方法的结果相比,蒙特卡罗方法不失为一种模拟中子俘获治疗的好工具.     

Peripheral doses in radiotherapy: A comparison between IMRT,VMAT and Tomotherapy

The goal of this intercomparison is to determine the peripheral doses during treatment of prostate and head and neck (H&N) cancers. In the case of prostate cancer, two different treatment techniques are compared: intensity-modulated radiation therapy (IMRT – 10 MV and 18 MV), on a Varian Clinac 2100 C/D and Tomotherapy. VMAT (also on a Varian Clinac 2100 C/D) was compared to Tomotherapy, for H&N cancer. The treatment devices are located at the university hospitals of Leuven and Brussels, respectively. A common treatment protocol was agreed between the two clinical centers and this same protocol was used by each partner. For the higher energy modalities (10 MV and 18 MV) we also assessed the neutron contribution to the total dose, by using bubble detectors. In this way, the performance (in terms of peripheral doses) of the different treatment techniques, when faced with the same dose distribution constraints, was evaluated. The doses were evaluated with an anthropomorphic phantom loaded with TLD detectors. Summarizing our results, we can conclude that low energy radiation techniques, namely VMAT and Tomotherapy, have more interesting performances when compared to IMRT at energies of 10 MV and 18 MV, with respect to peripheral dose. On the one hand the former are associated with lower photon doses and, on the other hand, there is no contribution from neutrons to the total dose.     

Photoparticle production in solid state detectors used for neutron dosimetry in the presence of photons

Photon-induced neutron, proton and alpha particle production in polyethylene and CR-39 has been estimated for the photon energy range of 2–30 MeV, using our previously established methods and photonuclear cross-section data for hydrogen, carbon and oxygen. The rarer isotopes of the constituents of CR-39 and polyethylene, namely ²H, ¹³C, ¹⁷O and ¹⁸O, have been taken into account. Neutrons and protons are produced in polyethylene and CR-39 for photon energies above 2.2 MeV, the (γ, np) threshold for ²H. Photoparticles produced in these materials may need to be taken into consideration when using them for neutron dosimetry in the presence of photons in this energy range, especially when the neutron flux is several orders of magnitude less than that of the photons.     

Dosimetry of 40 and 70 MeV/n O beams

Using CR-39 plastic track detectors the range values of ¹⁶O ions at two different energies (initially in the beam line, 39.97 MeV/n and 69.98 MeV/n) were measured after escaping the beam pipe and found to be (3050 ± 40) μm and (8210 ± 90) μm, respectively. The longitudinal and projected angular spread of oxygen ions of an initial energy of 69.98 MeV/n in the region of the Bragg peak was derived from the measured geometrical parameters of tracks. Based on a calibration curve (etch rate ratio vs total linear energy transfer in CR-39) and the measured track length distribution at the range end of oxygen ions, the complete depth dose profile of a 67.7 MeV/n ¹⁶O beam in CR-39 (plateau, extended Bragg peak and residual ionization caused by projectile like fragments) was obtained.     

Characterization of the radiation shielding properties of US and Russian EVA suits using passive detectors

Radiation measurements using passive detectors were carried out to assess the shielding properties of the US Extravehicular Mobility Unit (EMU) space suit and the Russian Orlan-M suit during irradiations of the suits and a tissue-equivalent phantom by monoenergetic proton and electron beams at the Loma Linda University Medical Center (LLUMC). During irradiations of 6 MeV electrons and 60 MeV protons, absorbed dose as a function of depth was measured using thermoluminescent detector (TLD) exposed behind swatches of the two suit materials and inside the two extravehicular activity (EVA) helmets. Considerable reduction in electron dose was measured behind all suit materials on exposure to 6 MeV electrons. Slowing of the proton beam in the suit materials led to an increase in dose measured on exposure to 60 MeV protons. During 232 MeV proton irradiations, measurements were made with TLD and CR-39 plastic nuclear track detector (PNTD) at five organ locations inside a tissue-equivalent phantom, exposed both with and without the two EVA suits. The EVA helmets produced a 13% to 27% reduction in total absorbed dose and a 0% to 25% reduction in dose equivalent when compared to measurements made in the phantom head alone. Differences in absorbed dose and dose equivalent between the suit and non-suit irradiations for the lower portions of the two EVA suits tended to be smaller. Proton-induced target fragmentation was found to be a significant source of increased dose equivalent, especially within the two EVA helmets, and average quality factor inside the EMU and Orlan-M helmets was 2% to 14% greater than that measured in the bare phantom head.     

Experimental and Monte-Carlo studies of the spatial distribution of neutrons around extended Pb-spallation target

The spatial distribution of thermal and fast neutrons on the surface of a paraffin moderator surrounding a cylindrical lead target, irradiated with 1 GeV protons was studied. The lead target had 8 cm diameter and 20 cm length. The thickness of the paraffin around the target was 6 cm. The slow and fast neutron distributions were determined using LR 115 2B and CR-39 detectors via the ¹⁰B(n,) reaction and neutron induced nuclear recoils, respectively. The observed slow and fast neutron distributions on the surface of the paraffin were compared with Monte-Carlo simulations using the MCNPX-2.1.5 code.     

Characterization of a BNCT beam using neutron activation and indirect neutron radiography

This study aims to measure the neutronic characteristics of the 14-cm diameter boron neutron capture therapy beam at the Tsing Hua Open-pool Reactor and to provide essential information for the neutron source adjustment and validation. The measurements were performed both in air and inside a cubic PMMA phantom at the beam, with and without an 18-cm long extended PE collimator. The neutron intensity was determined by neutron activation analysis; the indirect neutron radiography and the cadmium difference method were coupled to provide the two-dimensional neutron flux distributions.According to the measurements, the angular distribution is highly forward for epithermal neutrons but much more divergent for thermal ones. A PE extended collimator will modify the beam characteristics and could benefit the treatment for head and neck tumors, which often needs higher therapeutical boron dose at the shallow region.     

Microdosimetry for the characterization of the THOR epithermal neutron beam

The epithermal neutron beam of the Tsing Hua Open-pool Reactor (THOR) was constructed for the study of boron neutron capture therapy (BNCT). The THOR epithermal neutron beam was mainly composed of thermal neutrons, fast neutrons, and photons. For fast neutrons and photons, the absorbed dose and the relative biological effectiveness (RBE) were used to characterize radiation dose and radiation quality. The short-ranged alpha particles and lithium ions produced from ¹⁰B(n,α)⁷Li reactions in the BNCT required cellular- and micro-dosimetry characterizations. Due to the non-uniform microdistribution of boron in cells, these characterizations should depend on the source–target geometry. In this case, the geometry-dependent specific cellular dose and lineal energy could be used to describe radiation dose and radiation quality. In the present work, cellular- and micro-dosimetry were studied for the THOR epithermal neutron beam. The specific cellular dose and lineal energy were calculated for thermal neutron-induced α-particles and ⁷Li-ions with different source–target geometry and various cell sizes. Applying the linear energy dependent-biological weighting function, the geometry-dependent RBE of thermal neutron-induced heavy particles was determined. Finally, the effective RBE of the THOR epithermal neutron beam was estimated for tumors and normal tissues of specified ¹⁰B concentrations. This effective RBE should be multiplied by the total absorbed dose to determine the corresponding biological dose required in the treatment planning.     

Neutron response study using poly allyl diglycol carbonate

The results of an experimental work aimed at improving the performance of the CR-39 nuclear track detector for neutron dosimetry applications are reported. A set of CR-39 plastic detectors was exposed to ²⁵²Cf neutron source, which has the emission rate of 0.68 × 10⁸ s⁻¹, and neutron dose equivalent rate 1 m apart from the source is equal to 3.8 mrem/h. The detection of fast neutrons performed with CR-39 detector foils, subsequent chemical etching and evaluation of the etched tracks by an automatic track counting system was studied. It is found that the track density increases with the increase of neutron dose and etching time. The track density in the detector is directly proportional to the neutron fluence producing the recoil tracks, provided the track density is in the countable range. This fact plays an important role in determining the equivalent dose in the field of neutron dosimetry. These results are compared with previous work. It is found that our results are in good agreement with their investigations.      

Sample to moderator volume ratio effects in neutron yield from a PGNAA setup

Performance of a prompt gamma ray neutron activation analysis (PGNAA) setup depends upon thermal neutron yield at the PGNAA sample location. For a moderator, which encloses a sample, thermal neutron intensity depends upon the effective moderator volume excluding the void volume due to sample volume. A rectangular moderator assembly has been designed for the King Fahd University of Petroleum and Minerals (KFUPM) PGNAA setup. The thermal and fast neutron yield has been measured inside the sample cavity as a function of its front moderator thickness using alpha particle tracks density and recoil proton track density inside the CR-39 nuclear track detectors (NTDs). The thermal/fast neutron yield ratio, obtained from the alpha particle tracks density to proton tracks density ratio in the NTDs, shows an inverse correlation with sample to moderator volume ratio. Comparison of the present results with the previously published results of smaller moderators of the KFUPM PGNAA setup confirms the observation.     

无损检测用15MeV电子直线加速器主束内光中子剂量的测量与MC模拟计算

通过实验和模拟计算研究了无损检测用15MeV电子直线加速器X射线主束内的中子剂量. 加速器采用了铜复合靶和钨加含硼聚乙烯的屏蔽结构, 能够有效地减少光中子的产生, 中子产额在1/1000n/γ以下. 但由于主束内光子剂量很大, 中子的绝对强度也不容忽视. 针对加速器周围强X射线脉冲辐射场的特点, 采用了被 动型的中子剂量测量方法, 加速器正常工作情况下, 使用CR-39片和双电离室测量了等中心处中子对X射线的剂量当量比率, 分别为0.19mSv/Gy X-ray和0.060mSv/Gy X-ray. 利用MCNP5模拟计算了实验相应点的中子对X射线的剂量当量比率, 为0.092mSv/Gy X-ray, 与实验测量结果在数量级上一致. 加速器主射束上D_n/D_γ<1/1000,小于辐射防护标准对中子泄漏剂量的规定值, 从而验证了屏蔽结构的安全性.