Neutron dose transmission measurements for several new concrete samples including colemanite

We have prepared four baryte and four concrete samples having respectively 0%, 5%, 10% and 15% colemanite concentrations. Neutron dose transmission measurements have been done by using a source of mono-energetic neutrons (E_eff = 4.5 MeV–²⁴¹Am–Be). It has been shown that when colemanite percentages of the samples increase, neutron dose transmission values for the samples decrease. It is thus possible to enhance the neutron shielding property of baryte and ordinary concrete by adding different proportions of colemanite.     

Simulation of e–γ–n targets by FLUKA and measurement of neutron flux at various angles for accelerator based neutron source

A 6 MeV Race track Microtron (an electron accelerator) based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products where the low neutron flux requirement is desirable. The bremsstrahlung radiation emitted by impinging 6 MeV electron on the e–γ primary target, was made to fall on the γ–n secondary target to produce neutrons. The optimisation of bremsstrahlung and neutron producing target along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium and the measured fluxes were 1.1878 × 10⁵, 0.9403 × 10⁵, 0.7428 × 10⁵, 0.6274 × 10⁵, 0.5659 × 10⁵, 0.5210 × 10⁵ n/cm²/s at 0°, 30°, 60°, 90°, 115°, 140° respectively. The results indicate that the neutron flux was found to be decreased as increase in the angle and in good agreement with the FLUKA simulation.     

Design of the thermal neutron beam for neutron radiography at the Syrian MNSR

Neutron beam design was studied at the Syrian reactor (MNSR, 30 kW) with a view to generating thermal neutron beam in the vertical irradiation sites for neutron radiography. The design of the neutron collimator was performed using MCNP4C and the ENDF/B-V cross-section library. Thermal, epithermal and fast neutron energy ranges were selected as <0.4 eV, 0.4 eV–10 keV, >10 keV, respectively. To produce a good neutron beam quality, bismuth was used as photon filter. In this design, the L/D ratio of this facility had the value of 125. The thermal neutron flux at the beam exit was about 2.548 × 10⁵ n/cm² s. If such neutron beam were built into the Syrian MNSR many scientific applications would be available using the neutron radiography.     

Neutron transmission through pyrolytic graphite crystal II

The measured neutron transmissions through 6.7 mm thick pyroletic graphite (PG) crystal set at different take-off-angles with respect to the beam, as a function of wavelength, were compared with the calculated values using a general formula. An adapted version of the computer package graphite was developed in order to provide the required calculations in the neutron energy range from 0.1 MeV to 10 eV.     

Neutron flux determination in irradiation sites of an Am–Be neutron source at NNRI

Neutron flux measurements and flux distribution parameters for two irradiation sites of an Am–Be neutron source irradiator were measured by using gold (Au), zirconium (Zr) and aluminum (Al) foils. thermal neutron flux Φ_th = 1.46 × 10⁴ n cm⁻² s⁻¹ ± 0.01 × 10², epithermal neutron flux Φ_epi = 7.23 × 10² n cm⁻² s⁻¹ ± 0.001, fast neutron flux Φ_f = 1.26 × 10² n cm⁻² s⁻¹ ± 0.020, thermal-to-epithermal flux ratio f = 20.5 ± 0.36 and epithermal neutron shaping factor α = −0.239 ± 0.003 were found for irradiation Site-1; while the thermal neutron flux Φ_th = 4.45 × 10³ n cm⁻² s⁻¹ ± 0.06, the epithermal neutron Φ_epi = 1.50 × 10² n cm⁻² s¹ ± 0.003, the fast neutron flux Φ_f = 1.17 × 10 n cm⁻² s⁻¹ ± 0.011, thermal-to-epithermal flux ratio f = 29.6 ± 0.94, and epithermal neutron shaping factor α = 0.134 ± 0.001 were found for irradiation Site-2. It was concluded that the Am–Be neutron source can be used for neutron activation analysis (NAA). The Am–Be source can be used for neutron activation analysis thereby reducing the burden on GHARR-1 and increasing the research output of the nation.     

Measurements of neutron total cross-sections and resonance parameters of erbium at the Pohang Neutron Facility

We measured neutron total cross-sections of natural erbium in the neutron energy region from 0.2 to 120 eV by using the neutron time-of-flight method at the Pohang Neutron Facility, which consists of an electron linear accelerator, a water-cooled tantalum target with a water moderator, and a 12-m-long time-of-flight path. A ⁶Li-ZnS(Ag) scintillator with a diameter of 12.5 cm and a thickness of 1.6 cm was used as a neutron detector, and a group of high-purity natural erbium metallic plates with various thickness was used for the neutron transmission measurements. The present measurement was compared with the existing experimental and the evaluated data. The resonance parameters of ¹⁶⁶Er, ¹⁶⁷Er, ¹⁶⁸Er, and ¹⁷⁰Er in the neutron energy region below 120 eV were extracted from the transmission by using the multilevel R-matrix SAMMY code and were compared with the evaluated data from ENDF/B VII.0 and other previous reported results.     

Estimation of neutron production from accelerator head assembly of 15 MV medical LINAC using FLUKA simulations

For the production of a clinical 15 MeV photon beam, the design of accelerator head assembly has been optimized using Monte Carlo based FLUKA code. The accelerator head assembly consists of e-γ target, flattening filter, primary collimator and an adjustable rectangular secondary collimator. The accelerators used for radiation therapy generate continuous energy gamma rays called Bremsstrahlung (BR) by impinging high energy electrons on high Z materials. The electron accelerators operating above 10 MeV can result in the production of neutrons, mainly due to photo nuclear reaction (γ, n) induced by high energy photons in the accelerator head materials. These neutrons contaminate the therapeutic beam and give a non-negligible contribution to patient dose. The gamma dose and neutron dose equivalent at the patient plane (SSD = 100 cm) were obtained at different field sizes of 0 × 0, 10 × 10, 20 × 20, 30 × 30 and 40 × 40 cm², respectively. The maximum neutron dose equivalent is observed near the central axis of 30 × 30 cm² field size. This is 0.71% of the central axis photon dose rate of 0.34 Gy/min at 1 μA electron beam current.     

Neutron energy spectrum flux profile of Ghana’s miniature neutron source reactor core

The total neutron flux spectrum of the compact core of Ghana’s miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) × 10¹² n/cm² s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) × 10¹¹ n/cm² s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) × 10¹¹ n/cm² s. The peak values of the thermal energy range occurred in the energy range (1.8939–3.7880) × 10⁻⁰⁸ MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) × 10⁰⁹ n/cm² s at the lower energy end of the slowing down region between 8.2491 × 10⁻⁰¹ MeV and 8.2680 × 10⁻⁰¹ MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) × 10⁰⁸ n/cm² s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.     

Fluence and dose measurements for an accelerator neutron beam

The 3 MV Van de Graaff accelerator at McMaster University accelerator laboratory is extended to a neutron irradiation facility for low-dose bystander effects research. A long counter and an Anderson-Braun type neutron monitor have been used as monitors for the determination of the total fluence. Activation foils were used to determine the thermal neutron fluence rate (around 10⁶ neutrons s⁻¹). Meanwhile, the interactions of neutrons with the monitors have been simulated using a Monte Carlo N Particle (MCNP) code. Bystander effects, i.e. damage occurring in cells that were not traversed by radiation but were in the same radiation environment, have been well observed following both alpha and gamma irradiation of many cell lines. Since neutron radiation involves mixed field (including gamma and neutron radiations), we need to differentiate the doses for the bystander effects from the two radiations. A tissue equivalent proportional counter (TEPC) filled with propane based tissue equivalent gas simulating a 2 μm diameter tissue sphere has been investigated to estimate the neutron and gamma absorbed doses. A photon dose contamination of the neutron beam is less than 3%. The axial dose distribution follows the inverse square law and lateral and vertical dose distributions are relatively uniform over the irradiation area required by the biological study.     

Neutron irradiation-induced dimensional changes in MEMS glass substrates

A study is made of radiation-induced expansion/compaction in Pyrex^® (Corning 7740) and Hoya SD-2^® glasses, which are used as substrates for MEMS devices. Glass samples were irradiated with a neutron fluence composed primarily of thermal neutrons, and a flotation technique was employed to measure the resulting density changes in the glass. Transport of Ions in Matter (TRIM) calculations were performed to relate fast (∼1 MeV) neutron atomic displacement damage to that of boron thermal neutron capture events, and measured density changes in the glass samples were thus proportionally attributed to thermal and fast neutron fluences. Pyrex was shown to compact at a rate of (in Δρ/ρ per n/cm²) 8.14 × 10⁻²⁰ (thermal) and 1.79 × 10⁻²⁰ (fast). The corresponding results for Hoya SD-2 were 2.21 × 10⁻²¹ and 1.71 × 10⁻²¹, respectively. On a displacement per atom (dpa) basis, the compaction of the Pyrex was an order of magnitude greater than that of the Hoya SD-2. Our results are the first reported measurement of irridiation-induced densification in Hoya SD-2. The compaction of Pyrex agreed with a previous study. Hoya SD-2 is of considerable importance to MEMS, owing to its close thermal expansivity match to silicon from 25 to 500°C.     

Neutron transmission through pyrolytic graphite crystals

Calculation of the total cross-section, neutron transmission and removal coefficient of pyrolytic graphite (PG) for thermal neutron energies were carried out using an additive formula. The formula takes into account the variation of thermal diffuse and Bragg scattering cross-sections in terms of PG temperature and mosaic spread for neutron energies in the range 1 meV to 1 eV. A computer code PG has been developed which allow calculations for the graphite in its hexagonal close-packed structure, when its c-direction is parallel with incident neutron beam (parallel orientation).     

On the impact of neutron beam divergence and scattering on the quality of transmission acquired tomographic images

The impact of the divergence of a thermal neutron beam and the scattered neutrons on the quality of tomographic images acquired by transmission have been evaluated by using a third generation tomographic system incorporating neutron collimators under several different arrangements. The system equipped with a gaseous position sensitive detector has been placed at the main channel outlet of the Argonauta Research Reactor in Instituto de Engenharia Nuclear (CNEN-Brazil) which furnishes a thermal neutron flux of 2.3 × 10⁵ n cm⁻² s⁻¹. Experiments have then been conducted using test-objects with well-known inner structure and composition to assess the influence of the collimators arrangement on the quality of the acquired images. Both, beam divergence and scattering - expected to spoil the image quality - have been reduced by using properly positioned collimators between the neutron source and the object, and in the gap between the object and the detector, respectively. The shadow cast by this last collimator on the projections used to reconstruct the tomographic images has been eliminated by a proper software specifically written for this purpose. Improvement of the tomographic images has been observed, demonstrating the effectiveness of the proposed approach to improve their quality by using properly positioned collimators.     

Preparation of In low energy conversion electron sources

The preparation of ^114mIn sources of conversion electrons in the energy range 162-190 keV and β continuum with a 1989 keV endpoint via ion implantation of ¹¹³In into Al substrates and subsequent irradiation by thermal and epi-thermal neutrons in a nuclear reactor is described.     

Determination of thermal neutron cross-section and resonance integral for the As (n, γ)As reaction by activation method using Mn (n, γ)Mn as a monitor

Nuclear constants for use in reactor activation analysis especially (n, γ) cross-sections and absolute gamma intensities, are known to show a rather large scatter in literature. Thermal and resonance cross-sections for the ⁷⁵As (n, γ)⁷⁶As reaction is determined by the method of foil activation using ⁵⁵Mn (n, γ)⁵⁶Mn as a reference reaction. The experimental sample with and without a cadmium cover of 1-mm wall thickness was irradiated in the isotropic neutron field of the outer irradiation sites 7 of Ghana Research Reactor-1 facility which is a miniature neutron source reactor designed by the Chinese. The irradiation channel used has a neutron spectral parameter (α) found to be (0.037 ± 0.001). The induced activity in the sample was measured by gamma ray spectrometry with a high purity germanium detector. A standard solution of Arsenic was used for the analysis. The necessary correction for gamma attenuation, thermal neutrons and resonance neutron self-shielding effects were not taken into account during the experimental analysis because they were negligible. By defining cadmium cut-off energy of 0.55 eV, the result for ⁷⁵As (n, γ)⁷⁶As reaction was found to be: thermal neutron cross-section σ₀ = (4.28 ± 0.19) b and resonance integral I₀ = (61.88 ± 1.07) b.     

Neutron radiation effects on metal oxide semiconductor (MOS) devices

The main purpose of this study is to provide the knowledge and data on Deuterium-Tritium (D-T) fusion neutron induced damage in MOS devices. Silicon metal oxide semiconductor (MOS) devices are currently the cornerstone of the modern microelectronics industry. However, when a MOS device is exposed to a flux of energetic radiation or particles, the resulting effects from this radiation can cause several degradation of the device performance and of its operating life. The part of MOS structure (metal oxide semiconductor) most sensitive to neutron radiation is the oxide insulating layer (SiO₂). When ionizing radiation passes through the oxide, the energy deposited creates electron-hole pairs. These electron-hole pairs have been seriously hazardous to the performance of these electronic components. The degradation of the current gain of the dual n-channel depletion mode MOS caused by neutron displacement defects, was measured using in situ method during neutron irradiation. The average degradation of the gain of the current is about 35 mA, and the change in channel current gain increased proportionally with neutron fluence. The total fusion neutron displacement damage was found to be 4.8 × 10⁻²¹ dpa per n/cm², while the average fraction of damage in the crystal of silicon was found to be 1.24 × 10⁻¹². All the MOS devices tested were found to be controllable after neutron irradiation and no permanent damage was caused by neutron fluence irradiation below 10¹⁰n/cm². The calculation results shows that (n,α) reaction induced soft-error cross-section about 8.7 × 10⁻¹⁴ cm², and for recoil atoms about 2.9 × 10⁻¹⁵ cm², respectively.     

CR-39 detector based thermal neutron flux measurements, in the photo neutron project

PhoNeS (photo neutron source) is a project aimed at the production and moderation of neutrons by exploiting high energy linear accelerators, currently used in radiotherapy. A feasibility study has been carried out with the scope in mind to use the high energy photon beams from these accelerators for the production of neutrons suitable for boron neutron capture therapy (BNCT). Within these investigations, it was necessary to carry out preliminary measurements of the thermal neutron component of neutron spectra, produced by the photo-conversion of X-ray radiotherapy beams supplied by three LinAcs: 15 MV, 18 MV and 23 MV. To this end, a simple passive thermal neutron detector has been used which consists of a CR-39 track detector facing a new type of boron-loaded radiator. Once calibrated, this passive detector has been used for the measurement of both the thermal neutron component and the cadmium ratio of different neutron spectra. In addition, bubble detectors with a response highly sensitive to thermal neutrons have also been used. Both thermal neutron detectors are simple to use, very compact and totally insensitive to low-ionizing radiation such as electrons and X-rays. The resultant thermal neutron flux was above 10⁶ n/cm²s and the cadmium ratio was no greater than 15 for the first attempt of photo-conversion of X-ray radiotherapy beams.     

Neutron interrogation of high-enriched uranium by a 4 MeV linac

For revealing unauthorized transport (illicit trafficking) of nuclear materials, a non-destructive method reported earlier, utilizing a 4 MeV linear accelerator for photoneutron interrogation, was further developed. The linac served as a pulsed neutron source for assay of highly enriched uranium. Produced in beryllium or heavy water by bremsstrahlung, neutrons subsequently induced fission in the samples. Delayed neutrons were detected by a newly designed neutron collar built up of 14 ³He counters embedded in a polyethylene moderator. A PC controlled multiscaler served as a time analyzer, triggering the detector startup by the beam pulse. Significant progress was achieved in enhancing the detector response, hence the sensitivity for revealing illicit material. A lower sensitivity limit of the order of 10 mg ²³⁵U was determined in a 20 s measurement time with a reasonable amount of beryllium (170 g) or of heavy water (100 g) and a mean electron current of 10 μA. Sensitivity can be further enhanced by increasing the measurement time.     

Tomographic atom probe characterization of the microstructure of a cold worked 316 austenitic stainless steel after neutron irradiation

For the first time, chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after 17 years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa (7.81 × 10²⁵ neutrons.m⁻², E > 1 MeV). The samples were analysed with a laser assisted tomographic atom probe. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A high number density (6 × 10²³ m⁻³) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. Mo and P segregations at the interfaces of these clusters were also observed. Finally, Si enriched atmospheres were seen.     

An innovative research reactor design

A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < θ < π) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 × 10¹⁴ n cm⁻² s⁻¹, 3.0 × 10¹⁴ n cm⁻² s⁻¹, and 2.0 × 10¹⁴ n cm⁻² s⁻¹, respectively. Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature under steady-state operating conditions. Finally, a more compact version of the asymmetric core is developed where a maximum (non-perturbed) thermal flux of 5.0 × 10¹⁴ n cm⁻² s⁻¹ is achieved. The core life of this more compact version is estimated to be about 23 days.     

Measurement and analysis of the energy–angular distribution of secondary neutrons for Be at 5.9 and 6.4 MeV incident neutrons

The energy-angle double-differential neutron emission cross-sections of beryllium have been measured using the time of flight technique for 5.9 and 6.4 MeV incident neutrons, respectively, at 10 laboratory angles between 25° and 150°. The measured results are compared with model calculations based on the LUNF code and those of other authors and the ENDF/B-VII data. The estimation of the inelastic scattering neutron cross-sections leaving ⁹Be^∗ at the low-lying level states are also theoretically analyzed by the LUNF code. The experimental and calculated results indicated that the lowest (1.68 MeV) level still contributes to the (n, 2n) reaction with cross-sections of several 10 mb. The angular distributions and the angle-integrated elastic scattering cross-sections are also presented in comparison with other ones, these being in good agreement with the ENDF/B-VII data.