Effect of X-ray energies on induced photo-neutron doses

Photoneutrons induced by two high energies range from the Elekta medical linear accelerator (10 and 18 MV) were measured by nuclear track detectors (NTDs). CR-39 NTD in contact with converter screen slide films, natural boron of thickness 40 μm coated on the polyester film (BN1). Detectors were exposed at 100 cm SSD with field size 20×20 on the patient table, with chest phantom and with build-up Perspex used for high-energy exposure. CR-39 registers the thermal neutron by the (n–α) reaction with the thin layer of boron and the fast neutron was measured through the (n–p) elastic scattering with the H₂ molecules in the CR-39 constituents.It was found that the total neutron dose (thermal and fast) from the 18 MV X-ray is higher than that of 10 MV. The measured thermal neutron dose is relatively smaller than the fast neutron dose in the case of direct exposure at the two X-ray energies. On the other hand, in the case of measurements on phantom and upon the use of build-up Perspex sheets, the ratio of fast to that of thermal is less than that of direct exposure.     

The influence of field size and off-axis distance on photoneutron spectra of the 18 MV Siemens Oncor linear accelerator beam

At present, high energy electron linear accelerators (LINACs) producing photons with energies higher than 10 MeV have a wide use in radiotherapy (RT). However, in these beams fast neutrons could be generated, which results in undesired contamination of the therapeutic beams. These neutrons affect the shielding requirements in RT rooms and also increase the out-of-field radiation dose to patients. The neutron flux becomes even more important when high numbers of monitor units are used, as in the intensity modulated radiotherapy. Herein, to evaluate the exposure of patients and medical personnel, it is important to determine the full radiation field correctly. A model of the dual photon beam medical LINAC, Siemens ONCOR, used at the University Hospital Centre of Osijek was built using the MCNP611 code. We tuned the model according to measured photon percentage depth dose curves and profiles. Only 18 MV photon beams were modeled. The dependence of neutron dose equivalent and energy spectrum on field size and off-axis distance in the patient plane was analyzed. The neutron source strength (Q) defined as a number of neutrons coming from the head of the treatment unit per x-ray dose (Gy) delivered at the isocenter was calculated and found to be 1.12 × 10¹² neutrons per photon Gy at isocenter. The simulation showed that the neutron flux increases with increasing field size but field size has almost no effect on the shape of neutron dose profiles. The calculated neutron dose equivalent of different field sizes was between 1 and 3 mSv per photon Gy at isocenter. The mean energy changed from 0.21 MeV to 0.63 MeV with collimator opening from 0 × 0 cm² to 40 × 40 cm². At the 50 cm off-axis the change was less pronounced. According to the results, it is reasonable to conclude that the neutron dose equivalent to the patient is proportional to the photon beam-on time as suggested before. Since the beam-on time is much higher when advanced radiotherapy techniques are used to fulfill high conformity demands, this makes the neutron flux determination even more important. We also showed that the neutron energy in the patient plane significantly changes with field size. This can introduce significant uncertainty in dosimetry of neutrons due to strong dependence of the neutron detector response on the neutron energy in the interval 0.1–5 MeV.     

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.     

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.     

Radiation damage in triglycine sulphate due to fast neutron irradiation

Abstract

The change in electrical properties of TGS crystals due to induced defects created by fast neutron irradiation of two different energies (2 and 14 MeV) and different integrated neutron fluxes have been studied in the vicinity of phase transition. It is observed that the electrical conductivity increases with increase of neutron fluence up to 1.7 × 10¹⁰ n · cm^?2 and the values of the relative change of electrical conductivity in case of 2 MeV are higher than that of 14 MeV neutrons at the same neutron fluence (φ)     

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.     

Measuring fluence of fast neutrons with planar silicon detectors

The results of measurements of 1-MeV (Si) equivalent fast neutron fluence with silicon planar detectors are reported. The measurement method is based on the linear dependence of the reverse detector current increment on the neutron fluence: ΔI = α _I × Φ × V. This technique provides an opportunity to measure the equivalent fluence in a wide dynamic range from 108 to 10¹⁶ cm^–2 with an unknown neutron energy spectrum and without detector calibration. The proposed method was used for monitoring in radiation resistance tests of different detector types at channel no. 3 of IBR-2 and for determining the fluence of fission and leakage neutrons at the KVINTA setup.     

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.      

Plastics as nuclear track detectors for thermal neutron dosimetry

An attempt is made to determine the response of CR-39 and cellulose nitrate plastic track detectors subjected to thermal neutrons. The α-particles are produced from (n, α) reactions in lithium tetraborate convertor placed in contact with different plastics and are recorded in the detectors. The corrected track density gives a fluence sensitivity and dose sensitivity of the order of 10^?4 tracks per neutron and 10² tracks/cm² mrem respectively. A linear relationship is observed between track density and neutron fluence.     

Method for Monitoring of Neutron Fields near High-Energy Accelerators

The monitoring of neutron radiation from high-energy accelerators cannot fully rely on the standard dosimeters and radiometers manufactured in Russia, since these are sensitive only to neutrons with energies below some 10 MeV. This is because neutrons of higher energies can significantly contribute to the personnel doses both close to the accelerator shield and in the neutron multiscattered field around the shield. In this paper, we propose to measure the ambient neutron dose in energy range 10^–2 MeV to 1 GeV with a device consisting of two polyethylene balls with diameters of 3 and 10 in. housing slow-neutron detectors. The larger ball also comprises a lead converter (10'' + Pb). This device can be implemented in zonal radiation monitoring in the near-accelerator area.     

Passive neutron area monitor with pairs of TLDs as neutron detector

A passive neutron area monitor has been designed using Monte Carlo methods; the monitor is a polyethylene cylinder with pairs of thermoluminescent dosimeters (TLD600 and TLD700) as thermal neutron detector. The monitor was calibrated with a bare and a thermalzed ²⁴¹AmBe neutron sources and its performance was evaluated measuring the ambient dose equivalent due to photoneutrons produced by a 15 MV linear accelerator for radiotherapy and the neutrons in the output of a TRIGA Mark III radial beam port.     

Observation of thermoluminescence induced by fission fragments

Abstract

Thermoluminescence induced in CaF₂ powder by fission fragments emanating from a uranium foil bombarded by fast neutrons has been measured as a function of neutron fluence. A linear relationship between the glow produced and the fast neutron fluence between 5 × 10¹⁰ and 6.5 × 10¹¹ n/cm² has been obtained, thus establishing the feasibility of the use of this method for fast-neutron dosimetry. A limitation of the method is that, if the fissile foil is not separated from the phosphor after irradiation, the TL produced by the α-disintegration of ²³⁸U may eventually mask the fission-induced TL.     

Neutron spectrometry and determination of neutron ambient dose equivalents in different LINAC radiotherapy rooms

A project has been set up to study the effect on a radiotherapy patient of the neutrons produced around the LINAC accelerator head by photonuclear reactions induced by photons above ～8 MeV. These neutrons may reach directly the patient, or they may interact with the surrounding materials until they become thermalised, scattering all over the treatment room and affecting the patient as well, contributing to peripheral dose. Spectrometry was performed with a calibrated and validated set of Bonner spheres at a point located at 50 cm from the isocenter, as well as at the place where a digital device for measuring neutrons, based on the upset of SRAM memories induced by thermal neutrons, is located inside the treatment room. Exposures have taken place in six LINAC accelerators with different energies (from 15 to 23 MV) with the aim of relating the spectrometer measurements with the readings of the digital device under various exposure and room geometry conditions. The final purpose of the project is to be able to relate, under any given treatment condition and room geometry, the readings of this digital device to patient neutron effective dose and peripheral dose in organs of interest. This would allow inferring the probability of developing second malignancies as a consequence of the treatment. Results indicate that unit neutron fluence spectra at 50 cm from the isocenter do not depend on accelerator characteristics, while spectra at the place of the digital device are strongly influenced by the treatment room geometry.     

INVESTIGATION OF THE EFFECT OF SLOW NEUTRON IRRADIATION ON HALOGEN-DOPED HIGH-Tc SUPERCONDUCTORS AND ITS MECHANISM

Irradiation effect of low-fluence (-10⁸ n/cm² ) slow neutrons on halogen-doped superconductors is presented in this paper. And the mechanism of the effect is also described from the viewpoint of nuclear physics for both fast and slow neutrons on high-temperature superconductors (HTSC). It is shown ex-perimentally and theoretically that slow neutrons of low fluence has a similar irradiation effect to that of fast neutron beams with an energy E_n>0.1 MeV and fluence 10¹⁶-10¹⁸ n/cm²-However,quite differ-ent mechanisms are involved in them: Fast neutrons transfer their energies through elastic scattering in HTSC, whereas slow neutrons give off their energies during the slow neutron capture (n,γ) reaction.     

Supra-atomic structure of radiation-induced defects in synthetic quartz from neutron scattering data

The nanostructure of synthetic quartz samples irradiated with fast reactor neutrons with energies E n > 0.1 MeV has been studied by small-angle neutron scattering. The fluences are varied from 10¹⁷ to 2 × 10²⁰ neutrons/cm². In the quartz samples irradiated with fluences higher than 10¹⁷ neutrons/cm², point, extended (dislocation loops), and volume defects, namely, thermal peaks up to 50 nm in radius, are observed over the entire volume. At a fluence of 2 × 10²⁰ neutrons/cm², the total fraction of the formed defect regions, where the material is in a noncrystalline state, exceeds 10% of the sample volume. The data on the formation of a metamikt glassy phase in the quartz sample have been obtained.     

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.     

Use of CR-39 films for nuclear radiation shielding efficacy evaluation of lining materials for combat vehicles

All materials provide, to a lesser or greater extent, shielding against nuclear radiations. Armoured fighting vehicles (AFVs) have steel as the structural material, which appears to be a reasonably good gamma and neutron shield material but a shield of pure iron would not be equally effective against whole range of neutron energies as it has a few resonances in electron volt range, and it reduces energy of fast neutrons to lower energy neutrons. These neutrons will be absorbed through radiative capture and emit gamma radiations. Thus it is essential that an effective shield should contain a large amount of moderating material, hydrogen being preferred with low atomic number materials (B, C, Li) and lead (Pb) to ensure that the neutrons do not diffuse at intermediate energies in the shield as well as gamma attenuation will also take place. In order to have a suitable shield material for armoured vehicles which serves as neutron and gamma radiation attenuator, polyethylene polymer with fillers lining materials are preferred. These materials were evaluated against gamma and fast neutrons using radioactive sources for suitability to fitment into combat vehicle as per the requirement of protection factor values. The detector for gamma radiation was used as Nal(Tl) while for neutron, CR-39 film was used.      

基于D-T紧凑型中子源的快中子照相准直屏蔽体设计及中子束特性模拟研究

中子照相是一种重要的无损检测技术,它能用于火工产品、毒品和核燃料元件等的检测。基于紧凑型D-T中子发生器,完成了一个用于快中子照相的准直屏蔽体系统（BSA）的物理设计。根据D-T中子源的能谱和角分布建立了中子源模型,采用MCNP4C蒙特卡罗程序,模拟了准直屏蔽体系统中中子和γ射线的输运,准直中子束相对于单位源中子的中子注量可以达到9.30×10-6 cm-2,准直中子束中主要是能量大于10 MeV的快中子;在设置的样品平面直径14 cm的照射视野范围,准直束中子注量的不均匀度为4.30%,准直束中中子注量与γ注量的比值为17.20,中子通量和中子注量比值J/Φ为0.992,说明准直中子束有好的平行性;准直屏蔽体外的泄露中子注量率与准直束中子注量率相比降低了2个量级。所设计的准直屏蔽体能满足快中子照相的要求。Neutron radiography is an important nondestructive testing technique. It can be used to detect the explosive devices, drug and the nuclear fuel element, etc. A beam-shaping-assembly (BSA) based on a compact D-T neutron generator is designed for fast neutron radiography in this paper. D-T neutron source model is constructed based on the neutron energy spectrum and angular distribution data. The transportation of neutron and γ-ray in the BSA is simulated using MCNP4C code. The neutron fluence of the collimated neutron beam with respect to the neutron source of the unit source is 9.30×10-6 cm-2. The collimated neutron beams is mainly fast neutrons with energies greater than 10 MeV. In the irradiation field range with a diameter of 14 cm, the neutron fluence uniformity of the collimated beam is 4.3%, the ratio of the neutron fluence to the gamma fluence in the collimated beam is 17.20, and the neutron flux and the neutron fluence ratio (J/Φ) is 0.992 which indicates that the collimated neutron beam has good parallelism. The leakage neutron fluence in outside of BSA is two orders of magnitude lower than that of the collimated neutron beam. The designed BSA can meet the need of fast neutron radiography.     

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.     

Quasi-dynamic structural disorder induced by fast neutrons in Be3Al2Si6O18 crystals

The UV absorption spectra of beryl crystals exposed to fast neutrons with a fluence of 10¹⁴–10¹⁹ cm^?2 are investigated. It is found that as the fluence of particles increases, a characteristic fan-shaped broadening of the long-wavelength edge is observed for the impurity adsorption band with a charge transfer. The experimental results are interpreted on the basis of the generalized Urbach rule in the approximation of an induced quasi-dynamic disorder. The effective cross section of radiation-induced lattice disordering in the crystals under investigation is estimated at a value (σ=2.58×10^?18 cm²) close to the neutron amorphization cross sections for other crystalline silicates.