Am-Be Source Research Articles

Neutron Activation Analysis of Some Building Material

Neutron activation analysis (NAA), based on the comparator method, has the potential to fulfill the requirements of a primary ratio method as defined in 1998 by the Comité Consultatif pour la Quantité de Matière — Métrologie en Chimie (CCQM, Consultative Committee on Amount of Substance — Metrology in Chemistry Studies of Neutron Activation Analysis (NAA) have been carried out on different Egyptian building material samples. The technique of neutron activation analysis is based on the measurement of radiation released by the decay of radioactive nuclei formed by neutron irradiation of the material. The most suitable source of neutrons for such an application is usually a research reactor. The samples that can be analyzed with this method stem from a number of different fields, including medicine, nutrition, biology, chemistry, forensics, the environment and mining. Neutron activation analysis can be performed in a variety of ways. This depends on the element and the corresponding radiation levels to be measured, as well as on the nature and the extent of interference from other elements present in the sample. Most of the methods used are non-destructive, based on the detection of gamma radiation emitted by the irradiated material after or during the irradiation. Next to education and training, neutron activation analysis is the most widely used application of research reactors. Almost any reactor operating at 10-30 kilowatt of thermal power is capable of providing a sufficient neutron flux to irradiate samples for selective applications of this analysis technique. Another method of NAA by using two Am-Be isotopic neutron sources of activity 5 Ci were used in this investigation. The accomplished gamma rays were measured using 70 % HPGe spectrometer. This work demand to estimate the elements contained in cement products and its quality control. X-ray Fluorescence (XRF) measurements were done for confirming our results, and for determining the average neutron flux of 3.7× 103 n/cm2sec. The Natural radioactivities of these samples were measured before the analysis to know the background level of 40K, 238U and232Th nuclei. The results investigated that NAA agree with the results of XRF and the world range of the cement concentration of the essential elements Ca, Al, Na, Fe, Mn, V, Sr and Si.

Design of portable shield for neutron sources using MCNP Computer Code

In this work a proposed design of portable shield for neutron sources was performed theoretically using MCNP5 code. The shield is composed of three layers; lead layer for gamma attenuation, PMMA (Poly-methyl methacrylate) layer for fast neutron thermalization and a layer of B2O3 for thermal neutron absorption. The shield materials are combined in stainless steel container of 0.5 mm thick. The source design verification was carried out by experimental measurements at different source-center distance. The energy spectrum of 241Am-Be source is considered in order to study the neutron moderation due to successive elastic scattering with PMMA composition. The neutron and photon total dose rates was calculated at the surface and one meter from the shield. The results of MCNP5 model were compared with the international transport regulation to ensure the suitability of neutron shield proposed design. The results show that the proposed design of the portable shield satisfies the dose rate limits and can be used for control of external radiation level during neutron source transportation, ensure the safety transport conditions and protection of radiation workers in different applications.

Multiwalled boron nitride microtubes (BNMTs) shielding in thermal neutron absorbing

ABSTRACTIn this study, multiwalled boron nitride microtubes (BNMTs) were synthesised from amorphous boron powder in pure nitrogen atmosphere controlled graphite furnace at 1300°C for 6 h with amorphous boron/manganese (IV) oxide (MnO2) catalyst mass ratio of 1/1. It was found that the synthesised BNMTs outer diameters and the lengths vary between 0.68 and 1.57 µm and between 6.79 and 31.23 µm, respectively, by the scanning electron microscope analysis. The surface layer of AISI 1040 steel material was coated with multiwalled BNMTs at 950°C for 4 h in a foundry furnace. Transport phenomena of BNMT molecules on surfaces of the steel were considered by diffusion. Neutrons from obtained 241Am-Be source were irradiated on the 58Fe nuclei. The irradiation properties of the without and with BNMT coated steel materials were studied in the field of gamma emitted by 60Co. The radiation exposure rate of gamma radiation sources used at 1099 and 1291 keV, respectively. The higher peak area was obtained in steel coated with BNMT according to non-coated steel from irradiation results of samples. These results show that BNMT coated steel was more capable for radiation shielding.

Neutron spectra from Neutron Standards Laboratory (LPN/CIEMAT) sources with two Bonner sphere spectrometers

The Neutron Standards Laboratory (LPN: Laboratorio de Patrones Neutrónicos) of the Ionizing Radiation Metrology Laboratory (LMRI) of the CIEMAT (Centro de Investigaciones Energéticas Medioambientales y Tecnológicas) has two calibration neutron sources, 241Am-Be and 252Cf. Studies to verify the shielding of the facility and to characterize the neutron field in the irradiation room have been carried out using Monte Carlo methods. However, spectral measurements were not performed yet. In this study, we measured the total and direct neutron spectra of both calibration neutron sources. Two different Bonner Sphere Spectrometers (BSS) systems were allocated to 115 cm from the sources, on the calibration bench, with and without the shadow cones (SC). From the count rates obtained with both BSS systems (BSS-6LiI and BSS-3He), the total neutron spectrum and direct neutron spectrum were unfolded. Total and direct spectra obtained for the 241Am-Be and 252Cf sources with the MCNPX and different BSS systems, unfold codes and response matrices presented similar shape and are consistent with each other.

IMPROVEMENTS IN THE THERMAL NEUTRON CALIBRATION UNIT, TNF2, AT LNMRI/IRD.

The standard thermal neutron flux unit, TNF2, in the Brazilian National Ionizing Radiation Metrology Laboratory was rebuilt. Fluence is still achieved by moderating of four 241Am-Be sources with 0.6 TBq each. The facility was again simulated and redesigned with graphite core and paraffin added graphite blocks surrounding it. Simulations using the MCNPX code on different geometric arrangements of moderator materials and neutron sources were performed. The resulting neutron fluence quality in terms of intensity, spectrum and cadmium ratio was evaluated. After this step, the system was assembled based on the results obtained from the simulations and measurements were performed with equipment existing in LNMRI/IRD and by simulated equipment. This work focuses on the characterization of a central chamber point and external points around the TNF2 in terms of neutron spectrum, fluence and ambient dose equivalent, H*(10). This system was validated with spectra measurements, fluence and H*(10) to ensure traceability.

NEUTRON FIELD CHARACTERISTICS AT RADIATION METROLOGY LABORATORY OF STUK.

Radiation metrology laboratory at Radiation and Nuclear Safety Authority (STUK) is Finnish national laboratory for ionizing radiation. In order to establish personal dose equivalent available for neutron calibration service, a project was started to evaluate the neutron reference fields in the present facility. The irradiation room conditions were characterized in order to establish reference conditions for personal dose equivalent. To verify the validity of the calculations, ambient dose equivalent rates and room return were measured and calculated for 241AmBe and 252Cf sources. First results of Monte Carlo calculations and measurements are presented in this article.

Calibrations of the FDS Long Counter for Neutron Fluence Measurement

Long counters are widely used for monitoring neutron fluence owing to their constant response in a wide energy range. In this study, an extended long counter named FDS-LC (FDS Long Counter), having a flat response over a wide neutron energy range from 1 keV to 20 MeV, was developed to monitor high neutron fluence. The geometry and basic structure of FDS-LC was optimized by using Monte Carlo simulations, and it consists of the BF3 thermal neutron counter, the inner and outer polyethylene moderators, borated polyethylene absorption layer, and chromium and lead metal neutron multiplier. The parameters such as the effective center, the energy response, and the angluar response of the FDS-LC were estimated using Super Monte Carlo code. The experimental validation of these parameters were performed by using 241Am-Be source and T(d, n)3He neutron source at the China Institute of Atomic Energy in Beijing. The results showed that the fluctuation of the response in the energy range from 1 keV to 20 MeV was less than 12% and the effective center positions were approximately equal to The comparison of the simulation and experimental results of the angular response function showed good agreement with a maximum deviation less than 15.7%.

RESPONSE OF CsI:Pb SCINTILLATOR CRYSTAL TO NEUTRON RADIATION

The helium-3 world crisis requires a development of new methods of neutron detection to replace commonly used 3He proportional counters. In the past decades, great effort was made to developed efficient and fast scintillators to detect radiation. The inorganic scintillator may be an alternative. Inorganic scintillators with much higher density should be selected for optimal neutron detection efficiency taking into consideration the relevant reactions leading to light emission. These detectors should, then, be carefully characterized both experimentally and by means of advanced simulation code. Ideally, the detector should have the capability to separate neutron and gamma induced events either by amplitude or through pulse shape differences. As neutron sources also generate gamma radiation, which can interfere with the measurement, it is necessary that the detector be able to discriminate the presence of such radiation. Considerable progress has been achieved to develop new inorganic scintillators, in particular increasing the light output and decreasing the decay time by optimized doping. Crystals may be found to suit neutron detection. In this report, we will present the results of the study of lead doped cesium iodide crystals (CsI:Pb) grown in our laboratory, using the vertical Bridgman technique. The concentration of the lead doping element (Pb) was studied in the range 5х10-4 M to 10-2 M . The crystals grown were subjected to annealing (heat treatment). In this procedure, vacuum of 10-6 mbar and continuous temperature of 350°C, for 24 hours, were employed. In response to neutron radiation, an AmBe source with energy range of 1 MeV to 12 MeV was used. The activity of the AmBe source was 1Ci Am. The fluency was 2.6 х 106 neutrons/second. The operating voltage of the photomultiplier tube was 1700 V; the accumulation time in the counting process was 600 s and 1800 s. The scintillator crystals used were cut with dimensions of 20 mm diameter and 10 mm height.

Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.

Optimisation of pulse shape discrimination using EJ299-33 for high energy neutron detection in proton beam therapy

It is widely understood that proton beam therapy has considerable clinical benefits over photon therapy for treating certain types of tumours. Protons deposit most of their energy in a very localised area, the so-called Bragg peak, sparing surrounding healthy tissue and critical organs from radiation. However, secondary neutrons and gamma rays are generated in the beam nozzle and inside the patient. Clinically, it is highly desirable to monitor the neutron dose the patient is exposed to, and this requires a neutron detector sensitive to high energies. EJ299-33 is a solid plastic scintillator capable of discriminating neutrons from gamma rays using pulse shape analysis of scintillation light. EJ299-33 has the potential to detect neutrons with energies up to 100 MeV and does not present leakage and flammability hazards generally associated with liquid scintillators. Experimental measurements with 60Co, 137Cs and 241AmBe sources were performed to calibrate and optimise pulse shape discrimination parameters. We also performed experimental measurements at the Clatterbridge Cancer Centre in a 60 MeV passive scattered beam to detect high energy neutrons.

DEVELOPMENT, CHARACTERIZATION AND TESTING OF A SIMPLIFIED BONNER SPHERE SYSTEM FOR RAPID NEUTRON FIELD CHARACTERIZATION.

Personal neutron dosimetry requires workplace specific correction factors from detailed neutron field characterization. A Bonner sphere system (BSS) with a 3He proportional counter and a 3 inch an a 9 inch sphere was developed at SCK·CEN for rapid neutron field characterization. The BSS energy and angular responses were simulated with MCNPX and validated by irradiations with a thermal neutron beam at the Belgian Reactor 1 (BR1) and 252Cf and Am-Be sources at the SCK·CEN Laboratory for Nuclear Calibration (LNK). The BSS unfolding performance was tested using the FRUIT unfolding code for measurements of the scattered neutron field at the LNK. The unfolded spectra were in good agreement with MCNPX simulations.

A NOVEL TEPC FOR MICRODOSIMETRY AT NANOMETRIC LEVEL: RESPONSE AGAINST DIFFERENT NEUTRON FIELDS.

Tissue equivalent proportional counter (TEPC) is the most accurate device for measuring the microdosimetric properties of a particle beam, nevertheless no detailed information on the track structure of the impinging particles can be obtained, since the lower operation limit of common TEPCs is ~0.3 μm. On the other hand, the pattern of particle interactions at the nanometer level is measured by only three different nanodosimeters worldwide: practical instruments are not yet available. In order to partially fill the gap between microdosimetry and track-nanodosimetry, a low-pressure avalanche-confinement TEPC was recently designed and constructed for simulating tissue-equivalent sites down to the nanometric region. The present article aims at describing the response of this newly developed TEPC in the range 0.3 μm-25 nm against a fast neutron field from a 241Am-Be source and a quasi-monoenergetic neutron beam. The experimental results are in good agreement with Monte Carlo simulations carried out with the FLUKA code.

CHARACTERISATION OF THE GRAPHITE MODERATED THERMAL NEUTRON FIELD AT CMI

The graphite pile was set up at CMI to provide reference thermal neutron field for metrology and dosimetry purposes. It consists of three Pu-Be and three Am-Be radionuclide sources located in a 1.95 m (width) × 1.95 m (length) × 2.0 m (height) graphite moderator block. The neutron field in the volume of the experimental channel with dimensions of 40 cm × 40 cm × 135 cm (depth) was characterised experimentally and by Monte-Carlo calculations. Neutron fluence was determined by the activation of gold foils and Manganese tablets irradiated in nine different positions and by two 3He detectors of different construction irradiated in one position (bare and covered in three spherical PE moderators of 3, 3.5 and 4 inch diameter). Weighted mean value of the neutron fluence rate measured by all the abovementioned detectors in the pile centre was 2.91 × 104 cm-2 s-1 ± 1.8%. This value was used as a calibration factor for the Monte-Carlo model predictions. The neutron spectral fluence rate calculated by validated Monte-Carlo model was used to determine the conventionally true ambient and personal dose equivalent rates at different positions.

Calculation and Validation of the Response Matrix for a Neutron Multisphere Spectrometer with an Indium Central Detector.

Expansion of accelerator technologies into different application areas brings the need for neutron-field measurements in their vicinity. The measurements of neutron energy spectra and neutron dose rates can be done with the aid of neutron multisphere spectrometers with passive neutron detectors. In addition to the usually used thermoluminescent detectors and gold activation foils, another activation material-indium-can be considered. This paper presents a design of such an indium-based neutron multisphere spectrometer. Its response matrix was calculated with Monte Carlo code MCNP5 in the energy interval from 0.001 eV to about 20 MeV. The response matrix was validated with an Am-Be standard neutron source. MAXED code was used for neutron energy spectra unfolding. The values of the neutron fluence rate and neutron ambient dose equivalent rate in a reference point were calculated. The properties and advantages of indium for use in neutron multisphere spectrometry are discussed.

Fast-neutron and gamma-ray imaging with a capillary liquid xenon converter coupled to a gaseous photomultiplier

Gamma-ray and fast-neutron imaging was performed with a novel liquid xenon (LXe) scintillation detector read out by a Gaseous Photomultiplier (GPM). The 100 mm diameter detector prototype comprised a capillary-filled LXe converter/scintillator, coupled to a triple-THGEM imaging-GPM, with its first electrode coated by a CsI UV-photocathode, operated in Ne/5%CH4 at cryogenic temperatures. Radiation localization in 2D was derived from scintillation-induced photoelectron avalanches, measured on the GPM's segmented anode. The localization properties of 60Co gamma-rays and a mixed fast-neutron/gamma-ray field from an AmBe neutron source were derived from irradiation of a Pb edge absorber. Spatial resolutions of 12± 2 mm and 10± 2 mm (FWHM) were reached with 60Co and AmBe sources, respectively. The experimental results are in good agreement with GEANT4 simulations. The calculated ultimate expected resolutions for our application-relevant 4.4 and 15.1 MeV gamma-rays and 1–15 MeV neutrons are 2–4 mm and ∼ 2 mm (FWHM), respectively. These results indicate the potential applicability of the new detector concept to Fast-Neutron Resonance Radiography (FNRR) and Dual-Discrete-Energy Gamma Radiography (DDEGR) of large objects.

Light yield and energy resolution studies for SoLid phase 1

The SoLid experiment is searching for sterile neutrinos at a nuclear research reactor. It looks for inverse beta decays (producing a positron and a neutron in delayed coincidence) with a very segmented detector made of thousands of scintillating cubes. SoLid has a very innovative hybrid technology with two different scintillators which have different light emissions: polyvynil-toluene cubes (PVT) to detect the positrons and 6LiF:ZnS sheets on two faces of each PVT cube to detect the neutrons. It allows us to do an efficient pulse shape analysis to identify the signals from neutrons and positrons. The 288 kg detector prototype (SM1) took data in 2015. It demonstrated the detection principle and background rejection efficiency. The construction of SoLid phase I (∼ 1.5 t) has now started. To improve the energy resolution of SoLid phase I, we have tried to increase the light yield studying separately the two scintillators: PVT and ZnS. A test bench has been built to fully characterize and improve the neutron detection with the ZnS using an AmBe source. To study the positron light yield on the PVT, we have built another test bench with a 207Bi source. We have improved the design of the cubes, their wrapping or the type and the configuration of the fibers. We managed to increase the PVT light yield by about 66 % and improve the resolution of the positron energy on the test bench from 21 % to 16 % at 1 MeV.

Performance of the EDELWEISS-III experiment for direct dark matter searches

We present the results of measurements demonstrating the efficiency of the EDELWEISS-III array of cryogenic germanium detectors for direct dark matter searches. The experimental setup and the FID (Fully Inter-Digitized) detector array is described, as well as the efficiency of the double measurement of heat and ionization signals in background rejection. For the whole set of 24 FID detectors used for coincidence studies, the baseline resolutions for the fiducial ionization energy are mainly below 0.7 keVee (FHWM) whereas the baseline resolutions for heat energies are mainly below 1.5 keVee (FWHM). The response to nuclear recoils as well as the very good discrimination capability of the FID design has been measured with an AmBe source. The surface β- and α-decay rejection power of Rsurf < 4 × 10−5 per α at 90% C.L. has been determined with a 210Pb source, the rejection of bulk γ-ray events has been demonstrated using γ-calibrations with 133Ba sources leading to a value of Rγ −mis−fid < 2.5 × 10−6 at 90% C.L.. The current levels of natural radioactivity measured in the detector array are shown as the rate of single γ background. The fiducial volume fraction of the FID detectors has been measured to a weighted average value of (74.6 ± 0.4)% using the cosmogenic activation of the 65Zn and 68,71Ge isotopes. The stability and uniformity of the detector response is also discussed. The achieved resolutions, thresholds and background levels of the upgraded EDELWEISS-III detectors in their setup are thus well suited to the direct search of WIMP dark matter over a large mass range.

Preparation and characteristics of epoxy/clay/B4C nanocomposite at high concentration of boron carbide for neutron shielding application

In this research, the characteristics of the prepared samples in epoxy matrix by means of X-ray diffraction (XRD), energy dispersive X-ray spectroscopy (EDS), as well as scanning electron microscope (SEM) are evaluated. Meanwhile, the obtained mechanical properties of the specimen are investigated. Thermogravimetric analysis (TGA) is also employed to evaluate the thermal degradation of manufactured nanocomposites. The thermal neutron absorption properties of nanocomposites containing 3wt% of montmorillonite nanoclay (closite30B) have been studied experimentally, using an Am-Be point source. Mechanical tests reveal that the higher B4C concentrations, the more tensile strengths, but lower Young's modulus in all samples under consideration. TGA analysis also shows that thermal stability of the nanocomposite, increases in presence of B4C. Finally, neutron absorption analysis shows that increasing the B4C concentration leads to a nonlinearly build-up of neutron absorption cross section.

A new response matrix for a [formula omitted]LiI scintillator BSS system

A new response matrix was calculated for a Bonner Sphere Spectrometer (BSS) with a 6LiI(Eu) scintillator, using the Monte Carlo N-Particle radiation transport code MCNPX. Responses were calculated for 6 spheres and the bare detector, for energies varying from 1.059E(−9) MeV to 105.9 MeV, with 20 equal-log(E)-width bins per energy decade, totalizing 221 energy groups. A comparison was done among the responses obtained in this work and other published elsewhere, for the same detector model. The calculated response functions were inserted in the response input file of the MAXED code and used to unfold the total and direct neutron spectra generated by the 241Am-Be source of the Universidad Politécnica de Madrid (UPM). These spectra were compared with those obtained using the same unfolding code with the Mares and Schraube matrix response.

Experimental determination of the response functions of a Bonner sphere spectrometer to monoenergetic neutrons

A Bonner sphere spectrometer (BSS) plays an important role in characterizing neutron spectra and determining their neutron dose in a neutron-gamma mixed field. A BSS consisting of a set of nine polyethylene spheres with a 3He proportional counter was developed at Peking University to perform neutron spectrum and dosimetry measurements. Response functions (RFs) of the BSS were calculated with the general Monte Carlo code MCNP5 for the neutron energy range from thermal up to 20 MeV, and were experimentally calibrated with monoenergetic neutron beams from 144 keV to 14 MeV on a 4.5 MV Van de Graaff accelerator. The calculated RFs were corrected with the experimental values, and the whole response matrix was completely established. The spectrum of a 241Am-Be source was obtained after unfolding the measurement data of the BSS to the source and in fair agreement with the expected one. The integral ambient dose equivalent corresponding to the spectrum was 0.95 of the expected value. Results of the unfolded spectrum and the integral dose equivalent measured by the BSS verified that the RFs of the BSS were well established.