Photoneutron Source Research Articles

Photoneutron sources spectra calculation

The issues of increasing the accuracy of determining the average energy and the type of energy distribution (shape of the spectrum) of neutrons from Sb-Be- and Ra-Be-sources are considered. The indicated photoneutron sources are used to calibrate the energy scales of neutron spectrometers and to determine the energy dependence of the sensitivity of radiometers and neutron dosimeters. A method is proposed for calculating the energy spectra of radionuclide neutron sources of the Sb-Be (γ, n) and Ra-Be (γ, n) types, and the calculation results are presented. The dependence of the spectrum of the source on its design is investigated. The correctness of the results of calculating the energy spectra of radionuclide sources has been confirmed experimentally. The data obtained will be used to determine the energy dependence of the sensitivity of neutron spectrometers, radiometers and dosimeters.

Development of a Neutron Radiography System based on a 10 MeV Electron Linac

A thermal neutron radiography unit using the neutrons which emits a 10 MeV electron linac compact has been designed and simulated via MCNPX Monte Carlo code. The facility was carried out for an extensive range of values for the collimator ratio L/D, the main parameter which describes the quality of the produced radiographic images. The results show that the presented facility provides high thermal neutron flux; while with the use of single sapphire filter fulfills all the suggested values which characterize a high quality thermal neutron radiography system. A comparison with other similar facilities indicates that the use of a photoneutron source using a 10 MeV electrons beam is a useful substitutional for radiographic purposes.

Studying the Distribution of Thermal Neutrons from the W–Be output Channel of a Photoneutron Source

Distributions of thermal neutron flux density are measured with respect to the center of a W–Be photoneutron source and the axis of an output neutron channel. The measurements are made using activation detectors. The distributions are approximated by simple functions with parameters that depend on the detectors’ coordinates. The spatial distribution is obtained for thermal neutrons in the horizontal plane relative to the axis of the photoneutron source’s output channel.

Photoneutron reaction kinematics and error of commonly used approximations

An exact closed-form classical equation and a transcendental relativistic equation for the energy-angle relationship of the neutrons produced from the photoneutron reaction are derived in the laboratory coordinate system. Additional formulas are derived to describe the restriction on scattering angle in the double-valued photoneutron energy regime for incident photon energies near the kinematic threshold energy of the target nucleus. The full range of photoneutron reaction independent variables (target nucleus mass number, neutron separation energy, and photon energy) encountered in practical photoneutron production applications are explored to give the applicable range of classical kinematics and other approximations in the calculation of the photoneutron energy-angle relationship. A six-way comparison study of classical, relativistic, and approximation equations presented in this work and those implemented in radiation transport codes is performed. The well-known and widely-used approximations presented without derivation by Wattenberg and Hanson in the late 1940s are only applicable to energy spread of neutrons produced from gamma-ray photoneutron sources, the intended application of the approximations. The Wattenberg and Hanson approximations are inappropriate for general purpose radiation transport simulation and give unphysical negative photoneutron energies for incident photon energies near the kinematic threshold. Recent versions of the MCNP code appear to use erroneous equations related to inelastic neutron scattering to describe photoneutron kinematics resulting in the overestimation of photoneutron energy. Simulated photoneutron spectra are significantly hardened and cannot be used in contemporary nuclear science applications such as characterization of accelerator-driven photoneutron sources. The Monte-Carlo code developer and user communities need to exhaustively review, correct, document, verify, and validate photoneutron physics implementations in the codes that are in common use.

The influence of shielding reinforcement in a vault with limited dimensions on the neutron dose equivalent in vicinity of medical electron linear accelerator.

BackgroundHigh energy electron linear accelerators (LINACs) producing photon beams with energies higher than 10 MeV are widely used in radiation therapy. In these beams, fast neutrons are generated, which results in undesired contamination of the therapeutic beam. In this study, measurements and Monte Carlo (MC) simulations were used to obtain neutron spectra and dose equivalents in vicinity of linear accelerator.Materials and methodsLINAC Siemens Oncor Expression in Osijek University Hospital is placed in vault that was previously used for 60Co machine. Then, the shielding of the vault was enhanced using lead and steel plates. Measurements of neutron dose equivalent around LINAC and the vault were done using CR-39 solid state nuclear track detectors. To compensate energy dependence of detectors, neutron energy spectra was calculated in measuring positions using MC simulations.ResultsThe vault is a source of photoneutrons, but a vast majority of neutrons originates from accelerator head. Neutron spectra obtained from MC simulations show significant changes between the measuring positions. Annual neutron dose equivalent per year was estimated to be less than 324 μSv in the measuring points outside of the vault.ConclusionsSince detectors used in this paper are very dependent on neutron energy, it is extremely important to know the neutron spectra in measuring points. Though, patient dosimetry should include neutrons, estimated annual neutron doses outside the vault were far below exposure limit of ionizing radiation for workers.

Studying the Spatial Distribution of a Neutron Flux Using Detectors Based on Helium-3 and Boron-10

The spatial distribution of the neutron flux from the output channel of a photoneutron source is measured. The measurements are made using a moving helium counter and a position-sensitive detector based on a 10B layer. The parameters of neutron distribution are presented. The work is performed as part of preparing nuclear physics experiments and applied research using small angle scattering and diffraction.

Neutron transmission measurements at nELBE

Neutron total cross sections are an important source of experimental data in the evaluation of neutron-induced cross sections. The sum of all neutron-induced reaction cross sections can be determined with a precision of a few per cent in a relative measurement. The neutron spectrum of the photoneutron source nELBE extends in the fast region from about 100 keV to 10 MeV and has favourable conditions for transmission measurements due to the low instantaneous flux of neutrons and low gamma-flash background. Several materials of interest (in part included in the CIELO evaluation or on the HPRL of OECD/NEA) have been investigated: 197Au [1, 2], natFe [2], natW [2], 238U, natPt, 4He, natO, natNe, natXe. For gaseous targets high pressure gas cells with flat end-caps have been built that hold up to 200 bar pressure. The experimental setup will be presented including results from several transmission experiments and the data analysis leading to the total cross sections will be discussed.

A Comparative Study of Photo-Neutron Production from Flattened and Unflattened Beams in a Medical Linear Accelerator

High energy x-rays from a medical linear accelerator (LINAC) is used to treat cancer patients in radiotherapy department. The medical linac can produce bremsstrahlung x-rays from 4MV to 18MV and the commonly used linac produces 6MV, 10 MV and 15MV photons for treating major parts of the cancer patients. The photo neutron production during the radiation treatment is one of the major issues which have to be addressed with high impact because of the high LET nature associated with it. Photo neutrons are produced from several high atomic number (Z) elements present in the linac head mostly by the (ɣ, n) reaction. The flattening filter used in medical linac to produce uniform intensity distribution (made up of either tungsten, steel, lead, uranium, aluminum or a combination) is one of the photo neutron sources in the linac head. Advanced radiotherapy treatments like Intensity Modulated Radiotherapy (IMRT) and Volumetric Arc Therapy (VMAT) doesn't require a flat radiation field, and hence the flattening filter could be removed. This is called flattening filter free beam (FFF beam)1, 2. The removal of the filter will result in a forward peaked photon beam, which is very advantageous for these kinds of treatments3, 4. The purpose of this work was to find out the difference in the percentage of photo neutron production in a flattening filter free (FFF) and flattening filter (FF) beam using CR39 Solid State Nuclear Track Detector (SSNTD) from an Elekta Versa HD machine. The neutron fluence was measured at different positions with FFF and FF mode and the results were tabulated. The neutron fluence was lower for the FFF mode of operation than the FF mode. The reduction ranges from 2.27% to 44.22%.

Study of the Thermal Neutron Spectrum of the W–Be Photoneutron Source of the INR RAS

The possibility of reconstructing the neutron flux density of a W–Be photoneutron source using neutron activation analysis of Ag, Mn, Fe, Mn, and Ti samples is considered. The use of simple functions as AE–B, describing particular regions of the spectrum, as a reconstruction option is discussed. The dependence of the deviation of the measured reaction rates on those calculated in this study on the parameters of approximating functions is investigated. The convergence of the reconstruction results is obtained by increasing the amount of experimental data. This reconstruction method can be used for the operational measurement of the thermal neutron flux density of the photoneutron source and for the optimization of the source parameters.

Study of Extracted Beam Characteristics of the Photoneutron Source at the Institute for Nuclear Research of the Russian Academy of Sciences

The characteristics of the extracted beam of thermal neutrons at the IN-LUE photoneutron source are investigated. A collimator with a cross-section diameter of 30 mm, located at an angle of 67° relative to the electron beam direction, was used to extract neutrons from the source. The neutron flux density as a function of the distance from the source center and the distribution of flux density over the distance from the collimator axis (beam profile) are obtained.

Integral Benchmark Experiments on a Large Copper Block Using the GELINA Accelerator to Validate natCu Neutron Cross Sections From Different Neutron Cross-Sectional Databases

A neutronics integral benchmark experiment on a pure copper block, aimed at testing and validating recent copper nuclear data libraries, has been performed at Geel Electron LINear Accelerator Facility (GELINA). GELINA is a powerful photoneutron source using a 75- $\mu \text{A}$ , 110-MeV electron beam impinging on a depleted U rotating target, producing a white neutron spectrum ranging from the epithermal region up to about 20 MeV with a mean energy of about 1.4 MeV and intensity up to $3.2 \times 10^{13}$ n/s. A large natCu block (dimensions $60 \times 60 \times 60$ cm3) has been positioned at 100 cm from the target. Thin activation foils were used as neutron flux probes and located inside the block in six positions at different depths with respect to the main neutron propagation direction. Materials whose activation cross section are sensitive to different neutron energy ranges were used, and the measured fluxes were compared with calculations performed using the MCNP6 neutron transport code coupled to different neutron cross-sectional databases (FENDL3.1, JEFF33_T2, and ENDF 7.1). The MCNP6 calculation also used the neutron spectrum produced by the GELINA accelerator. This is the first time that a neutronics integral experiment on copper is performed using such a white neutron spectrum and the results of our comparison could be used to validate the neutron copper cross sections in the neutron energy range covered by GELINA. The C/E results, taking into account the sources of uncertainties, are satisfactory: no large differences are observed among the result obtained using the three cross-sectional databases; however, the JEFF-3.3 seems, slightly, better predict the results.

Investigation of fast neutron resonance transmission analysis based on the ultrashort pulsed electron beam-driven photoneutron source

Fast neutron resonance transmission analysis (FNRTA) is an effective method to identify the light nucleus by measuring its characteristic nuclear structure. One of the significant means of performing high-resolution FNRTA is associating a pulsed electron beam-driven photoneutron source with a neutron time of flight (TOF) detection system. The typical beam width which ranges from a few nanoseconds to several microseconds stretches the flight path to tens and even hundreds of meters for the energy resolution requirement and limits the FNRTA to laboratory analysis only. Currently, we can obtain an ultrashort-pulsed electron beam with a width of 8 ps owing to the femtosecond laser techniques. This benefits the energy resolution and accordingly shortens the flight path to a few meters, and dramatically reduces the volume of FNRTA and even makes it moveable. The work reported in this paper configured FNRTA by using an 8-ps-pulsed 45 MeV electron beam-driven photoneutron source and a 5-m TOF. Monte Carlo simulations are utilized to provide a detailed evaluation of the source and the energy resolution of TOF. The results illustrate that a pulsed neutron beam with energies ranging from sub keV to more than 10 MeV and with an intensity of 1.925× 108 n/s could be delivered. The energy resolution is evaluated to less than 1.67% within the energy region of 1–10 MeV. Preliminary resonance experiments of graphite are implemented by placing a graphite block close to the target. The resonance structure can be distinguished from the experimental results.

Studying the Thermal Neutron Spectrum of the W–Be Photoneutron Source at the Institute for Nuclear Research, Russian Academy of Sciences

The possibility is considered of reconstructing the thermal neutron flux density of a W–Be photoneutron source via neutron activation analysis of Ag, Mg, Fe, Mn, and Ti samples. The use of a simple function in the form AE−B describing the thermal region of the spectrum is discussed as an example of reconstruction. The dependence of the reaction rates’ deviation from the ones calculated on the parameters of the approximating function is investigated. The convergence of the reconstruction results as the number of experimental data grows is obtained. This means of recovery can be used for operational measurements of the thermal neutron flux density of the photoneutron source and to optimize the parameters of the source.

Determining the Energy Spectrum of Fast Neutrons in the n + 1H → n + p Reaction on the W–Be Source at the Institute for Nuclear Research, Russian Academy of Sciences

The neutron spectrum produced in the γ +9Be → 8Be + n reaction at different boundary energies of the photon spectrum is modeled in order to examine the possibility of measuring nuclear cross sections with fast quasi-monoenergetic neutron beams of the W–Be photoneutron source at the Institute for Nuclear Research. The obtained spectra are used to model the response of a hydrogen-containing detector to neutrons in the n + 1H → n + p reaction. An experimental setup for determining the energy spectrum of neutrons of the W–Be source is constructed. The spectra of secondary neutrons detected in an active hydrogen-containing target irradiated with neutrons are measured and compared to the ones modeled.

Studying the Possibility of Using Different Aerosol Filters for Neutron-Activation Analysis

Samples of different fibrous filtering materials are considered in selecting analytical filter materials suitable for sampling atmospheric aerosols and precipitates for neutron-activation analysis: quartz fiber filters, bleached gauze, LFS-2 analytical filtering material, Petryanov filters of two kinds, and an industrial nonwoven stitched filter. The filter samples are irradiated with neutrons from an IN-LUE photoneutron source based on the linear electron accelerator at the Russian Academy of Sciences’ Institute for Nuclear Research. The γ-activity of the samples is measured using a low-background gamma-spectrometer. Analysis of the activation gamma spectra corresponding to the samples of different filtering materials reveals the systematic presence of Mn, Na, and Cl (in different combinations) as their main impurities. A quartz fiber filter is the most suitable material for the neutron activation analysis. Estimates are made of the content of trace impurities in the investigated materials.

Characterization of photo-neutrons produced by 150 MeV and 1 GeV electrons impinging on high Z-metallic targets for neutron resonance spectroscopy

Monte Carlo calculations have been performed using MCNP code to study the generation, angular distribution and energy spectrum of photo-neutrons for 1 GeV and 150 MeV electron beam energies impinging on different thickness of Tungsten, Tantalum and Lead targets. It is noticed that the photo-neutron yield increases as the target thickness increases then saturates beyond an optimized thickness of the target. Moreover, the photo-neutron yield shows significant increase as the electron energy increases. At the optimized thickness, the angular distribution of photo-neutrons is found almost isotropic for 150 MeV electrons and anisotropic for 1 GeV electrons. Further, by increasing the electron energy and/or the target thickness the angular distribution is found to be forward peaked. The energy spectrum of photo-neutrons can be well described by a Maxwellian distribution for both electron energies. Such calculations can help in developing a photo-neutron source based time of flight facility (TOF) for elemental and isotopic identification via neutron resonance spectroscopy. Photo-neutron yields, angular distribution, mean energy, energy spectrum and nuclear temperature for 1 GeV and 150 MeV electron energies and different target materials are presented.

Study on background shielding for a compact photoneutron source

A compact photoneutron source (PNS), based on an electron LINAC was designed and constructed to provide required experimental nuclear data for Thorium Molten Salt Reactor (TMSR). All devices including the LINAC, the neutron production target and the detector systems were arranged in a shared hall, leading to a high background of neutron and γ-ray. The necessary shielding design was made to reduce the backgrounds. A series of preliminary measurements of total cross section in thermal neutron energy region for some key nuclides of TMSR, such as 232Th and 7Li, have been performed at this PNS. It was found, however, that additional shielding needs to be constructed to further reduce the neutron and γ-ray backgrounds. MCNP is a major tool for the background shielding design since it has the calculation capability for a coupled neutron-photon-electron transport problem. A two-step subsection simulation based on MCNP involving with variance reduction techniques (Sub-Var method) was performed to provide required flux and spectrum calculation for the new background shielding design. Based on the simulated results, a new local shielding was designed and constructed, which can reduce the thermal neutron background by three orders of magnitude and the γ-ray background by two orders of magnitude. The neutron flux and energy spectrum were measured under the new shielding condition and compared with the simulated ones, showing a good agreement with each other in the thermal energy region. Therefore, a significant improved shielding for the PNS is achieved, which is expected to provide a required assurance for the future measurements of neutron cross section at the PNS.

Performance of an electron linear accelerator for the first photoneutron source in China

A compact 15.0-MeV, 1.5-kW electron linear accelerator (LINAC) was successfully constructed to provide an electron beam for the first photoneutron source at the Shanghai Institute of Applied Physics, Shanghai, China. This LINAC consists of five main parts: a thermal cathode grid-controlled electron gun, a pre-buncher, a variable-phase-velocity buncher, a light-speed accelerating structure, and a high-power transportation beamline. A digital feedforward radio frequency compensator is adopted to reduce the energy spread caused by the transient beam loading effect. Furthermore, a real-time electron gun emission feedback algorithm is used to keep the beam stable. After months of efforts, all the beam parameters successfully met the requirements of the facility. In this paper, the beam commissioning process and performance of the LINAC are presented.

P263] Monte Carlo study of unflattened versus flattened 18 MV photon beam from medical linear accelerator as photoneutron source for BNCT applications

Purpose Studies have shown the possibility of using photoneutron emissions from medical linear accelerator as in-hospital neutron source for Boron Neutron Capture Therapy (BNCT) [1] . In this study, the effect of removing the?attening filter from an 18-MV photon beam on photoneutron production was investigated using the Monte Carlo (MC) method. Methods The 18-MV photon beam of Varian 2100 linear accelerator was simulated using the EGSnrc/dosxyznrc and Geant 4 MC codes. Beam dosimetric features, including depth doses, dose profiles, photon spectra distributions, photon angular distributions, photon x-y position and planner energy fluence were calculated for flattened and unflattened 18-MV photon beams. Results The results showed that unflattened (FFF) beam has higher photon fluencies and dose rate and thus much suited for photoneutron emission. The photon fluence and relative dose for the FFF beam decrease with increasing the distance from the central axis. The amount of dose reduction at the edge of the beam increased with field size. The mean energy decreased by a factor of 1.61% and 5.4% with field size for flattened and unflattened beam, respectively. Conclusions Because of these high photon intensities, the unflattened beam considered to be more powerful for photoneutrons production compared with the flattened beam.

Optimisation of a hybrid photoneutron source in a linear accelerator using GEANT4 and MCNPX Monte Carlo codes

Radiotherapy is important for treating oral cancer in advanced stages. Boron neutron capture therapy represents a unique modality in which neutron beams penetrate into the tissue within an epithermal range. Different methods are available for neutron production, and in this study, we used an electron accelerator. A photoneutron source based on three different energy values (10, 25, and 50 MeV) of a linear accelerator electron beam was designed using GEANT4 (GEometry ANd Tracking) and MCNPX (Monte Carlo N-Particle eXtended) simulation codes. The results indicate that a hybrid photoneutron source introduced tungsten and uranium with BeO. The statistical uncertainties in all simulations were less than 0.3 and 0.07% for MCNPX and the standard electromagnetic physics packages of GEANT4, respectively. Various cross-sectional and stopping-power data and different physics simulations produced different distributions.