Monte Carlo simulations of the production of neutrons at iThemba LABS

FLUKA Monte Carlo radiation transport code has been used to simulate neutron fluence spectrum at iThemba LABS neutron beam facility. Neutron beams with energy up to 200 MeV can be produced using different targets such as ⁷Li, ⁹Be and ¹²C bombarded with monoenergetic protons from the Separated-Sector Cyclotron. Simulated results at 66 MeV were compared with measured data. Different neutron emission angles with respect to the beam axis as well as the neutron background at different positions have been investigated.     

The role of the 14N(n,p)14C reaction in neutron irradiation of soft tissues

The cell-killing potential of the ¹⁴N(n,p)¹⁴C reaction was considered with regard to neutron absorption in human nuclear DNA and respiratory phosphates for: (A) 10¹² thermal neutrons in 1 kg of soft tissue, (B) a mono-energetic beam of 2 MeV neutrons incident in 1 kg of soft tissue such that the total collision kerma was 10 J/kg, and (C) an evenly distributed 0–66 MeV neutron beam, also incident in 1 kg such that the total collision kerma was 20 J/kg. For case (A) 0.0017 ¹⁴N(n,p)¹⁴C reactions could be expected per DNA double strand, case (B) 0.053, and case (C) 0.0039. The probabilities that a proton emitted outside the nucleus would cross nuclear DNA were estimated from ¹⁴N tissue content for adult skeletal muscle, liver, and kidney tissues, for (1) nuclear DNA being concentrated in a sphere of 1.8 μm diameter, and (2) nuclear DNA being evenly distributed in a spherical nucleus 5 μm in diameter. It was concluded that even in a nitrogen-rich tissue exposed to a collision kerma of 20 J/kg by a 0–66 MeV fast neutron beam, the ¹⁴N(n,p)¹⁴C reaction directly kills at most 10 cells in every 1000, 4 of these by DNA nitrogen absorption and 6 by the ¹⁴N(n,p)¹⁴C protons emitted elsewhere in the cell. However, the dose due to the ¹⁴N(n,p)¹⁴C reaction should be measured where exposure to thermal neutron fluxes is significant. For therapeutic neutron doses the number of respiratory phosphate molecules in which the ¹⁴N(n,p)¹⁴C reaction occurs is insignificant, and doses from ¹⁴C-decay after neutron therapy are also negligible.     

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.     

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).     

Neutron,photon and proton energy spectra at high altitude measured using a phoswich-type neutron detector

Neutron energy spectrum from 7 to 180 MeV, photon energy spectrum from 4 to 50 MeV and proton energy spectrum from 94 to 145 MeV were measured simultaneously using a phoswich-type neutron detector with particle discrimination methods at atmospheric depth of 249 g/cm², a vertical cut-off rigidity of 10.2 GV and at a heliocentric potential of 312 MV. We compared our results with other measured and calculated particle energy spectra. Our measured results give a large, sharp neutron peak around 70 MeV, although Bonner balls show a broad peak around 100 MeV due to low energy resolution. The measured photon and proton spectra are between the calculated energy spectra. This onboard study provides the first experimental neutron energy spectrum over 10 MeV with a high-energy resolution.     

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

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.     

Neutron ambient dose equivalent from 5 MeV/u 10,11B, 12,13C and 16,18O projectiles incident on a thick Al target

The neutron ambient dose equivalent has been measured from ¹⁰B, ¹¹B, ¹²C, ¹³C, ¹⁶O and ¹⁸O projectiles of energy 5 MeV/amu incident on a thick Al target at 0°, 30°, 60° and 90° with respect to the beam direction using a conventional dose equivalent meter. The calculated results obtained using previously reported empirical relations do not reproduce the experimental data. The results obtained from the PACE nuclear reaction code are closer to the experimental data as compared to the various empirical expressions. The ratio of the increase in the dose rates when the projectile is changed from the lighter to the heavier isotopes is fairly reproduced by most of the empirical formulations and the PACE code. A previously reported relation for the slope parameter is used to predict the directional distribution of the neutron dose for the projectiles used in this study. The calculated doses are lower than the experimental results in the forward directions but agree within the uncertainties at the backward directions. A new set of projectile-based parameters have been derived from the present experimental data which can be used in an empirical formulation.     

Characterisation of the high-energy neutron beam at iThemba LABS

Monoenergetic and white neutron beams with a maximum energy of 200 MeV are produced at the neutron beam facility of the iThemba Laboratory for Accelerator-Based Sciences (iTL) by (p, n) reactions on Li, Be and C targets. The iTL separated-sector cyclotron can accelerate protons from 25 MeV to 200 MeV. Time of flight (TOF) measurements can be carried out by increasing the time separation between proton bunches up to 500 ns with a beam pulse selector. In the present work, the methods of beam characterisation used at this facility are discussed. Comparisons with metrological standards are presented. Specific issues related to monitoring the neutron fields at different accelerator settings are analysed.     

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.     

Neutron dose distribution from 12C induced reactions on Ti and Ag using proton recoil scintillator

We measured the double differential neutron yield at 0°, 30°, 60° and 90° from ¹²C⁵⁺ induced reactions on thick targets of Ti and Ag at 12 MeV/amu at the Cyclotron at National Institute of Radiological Sciences, Inage, Japan, with 5″ × 5″ proton recoil scintillation detectors BC-501. The measured neutron spectra were unfolded using pulse height unfolding algorithm and energy and angular distribution obtained. Energy distribution of neutron ambient dose equivalent, H*(10) and absorbed dose, D at different angles was determined from double differential neutron spectra using ICRP recommended fluence to dose conversion coefficients.     

Radiation Protection of CSR

The construction of CSR (cooling storage ring) which includes a main ring (CSRm) and an experimental ring (CSRe) will be finished at the end of 2005. Heavy ions of carbon to uranium will be accelerated up to 900MeV/u and 400MeV/u at intensity of 10⁸ pps. The HIRFL (heavy ion research facility in Lanzhou) will be used as the injector. For the shielding design of CSR, the secondary neutrons due to the ion beam loss, their spectra and angular distributions were estimated based on the experimental results. The dose equivalent outside the shielding surface and in the surrounding environment and the neutron skyshine dose equivalent were also estimated in this study. The experimental result, neutron yield, spectrum and angular distribution for 400MeV/u ¹²C+Cu reaction were used for estimating the source term of shielding design. It is found that the most important environmental radiation impact component of CSR is the skyshine neutrons.     

Evaluation of target photon dose mixed in mono-energetic neutron fields using 7Li(p,n)7Be reaction

Target photons mixed in the 144, 250 and 565 keV mono-energetic neutron calibration fields were measured using a cylindrical NaI(Tl) detector with 7.62 cm both in diameter and in length. The ambient dose equivalent H*(10) of the photons was evaluated by applying the “G(E) function” to the measured pulse height spectrum. Neutrons induce photons by nuclear reactions in the NaI(Tl) detector and affect the pulse height spectrum. In order to eliminate the influence of these neutron events, the time-of-flight technique was applied with operating the accelerator in the pulse mode. The ratios by the ambient dose equivalent H*(10) of the photons to the 144, 250 and 565 keV neutrons were evaluated to be 3.3%, 4.7% and 0.9%, respectively. Although high energy photons ranging from 6 to 7 MeV are emitted by the ¹⁹F(p,αγ)¹⁶O reactions, the dose of the target photons is low enough to calibrate neutron dosemeters except for ones with high sensitivity to the photons.     

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

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.     

Application of 10B counter with moderator for neutron ambient dose equivalent measurement during radiation monitoring at joint institute for nuclear research

The results of an investigation into the possibility of applying a device based on a ¹⁰B neutron counter (CHM-14) with a polyethylene moderator as the dosimeter of neutron ambient dose equivalent H*(10) in radiation fields of nuclear physics installations at the Joint Institute for Nuclear Research (JINR) are presented. It is shown that the device can be used as the dosimeter of this quantity in the neutron energy range from 0.4 eV to 20 MeV with an error no larger than 30% due to the difference between the energy dependence of its response and the energy dependence of the neutron ambient dose equivalent. Applying the correction coefficients allows one to extend the energy range of neutron dose H*(10) measurement to hundreds MeV. The error due to the anisotropy of the device response does not exceed 35%.     

Deformed even-even nuclei in a quark-meson coupling model with tensor coupling

Neutron total cross sections of ¹⁹⁷Au and ^natTa have been measured at the nELBE photoneutron source in the energy range 0.1–10MeV with a statistical uncertainty of up to 2% and a total systematic uncertainty of 1%. This facility is optimized for the fast neutron energy range and combines an excellent time structure of the neutron pulses (electron bunch width 5ps) with a short flight path of 7m. Because of the low instantaneous neutron flux transmission measurements of neutron total cross sections are possible, that exhibit very different beam and background conditions than found at other neutron sources.     

Simulation and experimental study of an indigenously designed and constructed THGEM-based microdosimeter for dose-equivalent measurement

Most of the GEM/THGEM-based microdosimetric detectors presented in the literature simulate 2 μm of tissue which results in a flat neutron dose-equivalent response in the MeV region. The objective of this work was to introduce a neutron microdosimeter with a more extended flat response. In this regard, a THGEM-based microdosimeter with plexiglas walls, simulating 1 μm of tissue was designed and constructed. Its performance was investigated by both simulation and experimentation to determine the microdosimetric quantity of “lineal energy”.In the simulation study, lineal energy distribution, mean quality factor and dose-equivalent response of the microdosimeter for eleven neutron energies from 10 keV to 14 MeV, along with the energy spectrum of ²⁴¹Am-Be neutrons, were calculated by the Geant4 simulation toolkit. Obtained lineal energy distributions were compatible with the distributions determined by a Rossi counter. Also, the mean quality factors agreed well with the values reported by the ICRU report 40 which confirmed tissue equivalent behavior of the microdosimeter. They were different from the effective quality factor values within 15% between 20 keV and 14 MeV. This led to a flat dose-equivalent response with 20% difference from a median value of 0.82 in the above energy range which was an improvement compared with other THGEM-based detectors, simulating 2 μm of tissue. In spite of the satisfactory determination of the dose-equivalent, the microdosimeter had low detection sensitivity.In the experimental study, the measured lineal energy distribution of ²⁴¹Am-Be neutrons was in agreement with the simulated distribution. Further, the measured mean quality factor and dose-equivalent differed by 1.5% and 3.5%, respectively, from the calculated values. Finally, it could be concluded that the investigated microdosimeter reliably determined the desired dose-equivalent value of each neutron field with every energy spectrum lying between 20 keV and 14 MeV.     

Russian measurements of neutron energy spectra on the Mir orbital station

Results of the experiments on neutron energy spectra measurements within broad energy range from 5 x 10(-7) to 2 x 10(2) MeV aboard the Mir orbital station and equivalent neutron dose estimation are presented. Four measurement techniques were used during the experiments. The shape of spectra and their absolute values are in good agreement. According to those experiments, an equivalent neutron dose depends upon effective shielding thickness and spacecraft mass. The neutron dose mentioned is comparable with that of ionizing radiation. Neutron flux levels measured aboard the Mir station have shown that a neutron spectrometer involving broad energy range will be used within the radiation monitoring systems in manned space flights.     

Dosimetry measurements at the fast neutron therapy facility in Seattle

The fast neutron therapy facility at the University of Washington has been in routine clinical use for 25 years. 50.5 MeV protons produce neutrons in a beryllium target mounted on an isocentric gantry. Beam shaping is accomplished with a 40-leaf collimator. Dosimetry measurements for treatment planning and calibration are performed with tissue equivalent ion chambers. A layered phantom of alternating Solid Water^® and Plastic Water^® slabs has been developed for rapid dose verification measurements. The neutron field in the room has been used for radiation testing of electronic components.     

Measurement of the neutron beam polarization of BL05/NOP beamline at J-PARC

We measured the neutron beam polarization of the BL05/NOP (Neutron Optics and Physics) beamline at J-PARC with an accuracy of less than 10⁻³ using polarized ³He gas as a neutron spin analyzer. Precise polarimetry of the neutron beam is necessary to understand the beamline optics as well as for the asymmetry measurements of the neutron beta decay, which are planned in this beamline.     

CSR的辐射防护

CSR(cooling storage ring)按计划将于2005年底建成调束,届时从¹²C到²³⁸U的重离子将可以分别被加速到900和400MeV的能量. HIRFL(兰州重离子加速器Heavy Ion Research Facility in Lanzhou)将 用作CSR的注入器. 为了CSR的屏蔽设计,本文利用现有的实验数据计算了由于束流损失产生的中子及其能谱、角分布,同时也估算了屏蔽体外表面的中子剂量、环境中子剂量及天空返照中子剂量. 在源项计算中使用了400MeV/u ¹²C+Cu反应的中子产额、能谱、角分布的实验数据. 计算表明, CSR对环境剂量影响最大的是天空返照中子.