Neutron beam preparation with Am–Be source for analysis of biological samples with PGNAA method

Material analysis with prompt gamma neutron activation analysis (PGNAA) requires a proper geometrical arrangement for equipments in laboratory. Application of PGNAA in analysis of biological samples, due to small size of sample, needs attention to the dimension of neutron beam. In our work, neutron source has been made of ²⁴¹Am–Be type. Activity of ²⁴¹Am was 20 Ci which lead to neutron source strength of 4.4 × 10⁷ neutrons per second. Water has been considered as the basic shielding material for the neutron source. The effect of various concentration of boric acid in the reduction of intensity of fast and thermal components of the neutron beam and gamma ray has been investigated. Gamma ray is produced by (α, n) reaction in Am–Be source (4.483 MeV), neutron capture by hydrogen (2.224 MeV), and neutron capture by boron (0.483 MeV). Various types of neutron and gamma ray dosimeters have been employed including BF₃ and NE-213 detectors to detect fast and thermal neutrons. BGO scintillation detector has been used for gamma ray spectroscopy. It is shown that the gamma and neutron radiation dose due to direct beam is of the same magnitude as the dose due to radiation scattered in the laboratory ambient. It is concluded that 14 kg boric acid dissolved in 1,000 kg water is the optimum solution to surround the neutron source. The experimental results have been compared with Monte Carlo simulation.     

Investigation of fast neutron shielding characteristics depending on boron percentages of MgB<Subscript>2</Subscript>, NaBH<Subscript>4</Subscript> and KBH<Subscript>4</Subscript>

The macroscopic cross-section Σ and average neutron fluence in matter Φ are usable factors to comment neutron shielding property of samples. In this paper, we have used MgB₂, NaBH₄ and KBH₄ samples including different percentages of boron. Neutron macroscopic cross-section measurements of them have been done by using a source of mono-energetic neutrons (E _eff = 4.5 MeV ²⁴¹Am–Be). Average neutron fluence values and double differential fast neutron flux distributions of each samples calculated by using FLUKA Monte Carlo code. Also half value layers (HVLs) of samples are compared to paraffin which is one of the most neutron moderators. As a result, growing boron concentration can raise neutron shielding property of materials.     

Prompt gamma-ray neutron activation analysis methodology for determination of boron from trace to major contents

Prompt gamma ray neutron activation analysis methodologies were standardized using a reflected neutron beam and Compton suppressed γ-ray spectrometer to quantify boron from trace to major concentrations. Neutron self-shielding correction factors for higher boron contents (0.2–10 mg) in samples were obtained from the sensitivity of chlorine by irradiating KCl with and without boron. This method was validated by determining boron concentrations in six boron compounds and applied to three borosilicate glass samples with boron contents in the range of 1–10 mg. Low concentrations of boron (10–58 mg kg⁻¹) were also determined in two samples and five reference materials from NIST and IAEA.     

10B quantitative determination in ppb range by particle tracks reading for boron neutron capture therapy

Quantitative determination of ppb-order ¹⁰B was carried out in a small number of cultured tumor cells by particle tracks by the ₁₀B(n,)⁷Li reactions using solid state nuclear track detector (SSNTD) for boron neutron capture therapy research. The detection limit of ¹⁰B concentration in tumor cells is less than 1·10^-3 ppm under our experimental conditions which permits the quantitative determination of ¹⁰B in a small number of boronated tumor cells (10⁵/ml = ca. 0.1 mg). Boron quantities in tumor cells determined were well compatible with the tumor cell killing effects of boron neutron capture therapy.     

Gamma-ray analysis of renal cadmium by pulsed thermal neutron capture

A system has been developed for measuring the amount of cadmium in the left kidney. The in-vivo technique utilises prompt γ-rays arising from neutron capture by¹¹³Cd. A plused neutron beam is produced by bombarding a beryllium target with 10 MeV protons. The sensitivity of the measurements has a limit of 3.25 mg (2SE) for a localised dose of 2.0 rem.     

Measurements of fast neutron capture cross sections on <Superscript>63</Superscript>Cu and <Superscript>186</Superscript>W

Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were measured by fast neutron activation method at neutron energies from 1 to 2 MeV. Monoenergetic fast neutrons were produced by ³H(p,n)³He reaction. Neutron energy spread by target thickness, which was assumed to be the main factor of neutron energy spread, was estimated to be 1.5% at neutron energy of 2.077 MeV. Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were calculated by reference comparison method on those of ¹⁹⁷Au(n,γ). Not only statistical errors of gamma-counts from samples but also systematic errors in the counting efficiency for HP Ge detector and the uncertainty of areal density of samples were considered in calculating neutron capture cross section. Estimated neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were also compared with ENDF-6 data.     

Prompt gamma neutron activation analysis of boron with a 241Am-Be neutron source

A prompt gamma neutron activation analysis (PGAA) setup installed at ANRTC has been used to analyze boron. It consists of a 22.6% REGe detector and a 740 GBq ²⁴¹Am-Be neutron source moderated with water and paraffin. At the sample irradiation position, the thermal neutron fluence rate measured was 2.36·10⁴ n·m^–2· s^–1 and the corresponding Cd-ratio was 22 for gold monitor. The absolute detection efficiency in the range of 120–1500 keV was determined using ¹⁵²Eu standard solution. The sensitivity and detection limit for standard boric acid samples has been determined. The boron content in boric acid prepared from Turkish borate ores is measured to be 15.91±0.46% wt.     

Prompt gamma neutron activation analysis of boron with a 241Am-Be neutron source

A prompt gamma neutron activation analysis (PGAA) setup installed at ANRTC has been used to analyze boron. It consists of a 22.6% REGe detector and a 740 GBq ²⁴¹Am-Be neutron source moderated with water and paraffin. At the sample irradiation position, the thermal neutron fluence rate measured was 2.36·10⁴ n·m^–2· s^–1 and the corresponding Cd-ratio was 22 for gold monitor. The absolute detection efficiency in the range of 120–1500 keV was determined using ¹⁵²Eu standard solution. The sensitivity and detection limit for standard boric acid samples has been determined. The boron content in boric acid prepared from Turkish borate ores is measured to be 15.91±0.46% wt.     

Measurement of low concentrations of hydrogen in small samples by intermediate neutron moderation

An improved method based on the moderation of intermediate neutrons for the measurement of hydrogen in small samples is described. With the aid of boron and cadmium filters, a space shielded from slow neutrons is set up close to an isotopic neutron source shrouded by water moderator. A BF₃ proportional counter enclosed with a sample cell is placed in this space. The neutron count rate of the counter increases when a hydrogen-containing material is introduced into the cell, due to the moderation of intermediate neutrons passing through the filters. With a 1.3 μg²⁵²Cf neutron source, the lower limit of hydrogen detection for 200 ml samples in 10 min count time is 0.01 wt.%. This method is suitable for measuring the H₂O content of heavy water.     

Neutron spectrum and doses in a 18 MV LINAC

Bonner sphere spectrometer with TLDs pairs has been utilized to measure the neutron spectrum 100 cm from isocenter of a 18 MV LINAC, simultaneously the ambient dose equivalent due to neutrons and photons was measured in the control room area with neutron and gamma-ray area monitors. Measurements were carried out when the LINAC was delivering a dose of 600 MU at the isocentre that was located at 5 cm depth of a head phantom. Undesired neutron field in the treatment room produce activation reactions with nuclei in different materials of LINAC, couch, air, and phantom. To determine the dose due to decay of activation nuclei the ambient dose due to gamma-rays was measured inside the treatment room immediately after dose was delivered. Measured spectrum has two peaks, one between 0.1–1 MeV and other in the thermal region, the ambient dose equivalent in the control room are 3.1 and 0.93 μSv h⁻¹ for photons and neutrons, respectively. In the treatment room the ambient dose equivalent due to photons produced during decay of activation nuclei varies from 6 to 26.1 μSv h⁻¹.     

Measurement of the 238U(n,γ)239U reaction cross sections in the energy range of 2.0–5.0 MeV by using a neutron activation technique

The neutron capture cross-sections of ²³⁸U at the neutron energies of 4.38?±?0.05 MeV, 3.02?±?0.49 MeV and 2.04?±?0.26 MeV have been measured using the activation method and off-line gamma-ray spectrometric technique. The effects of neutron flux fluctuation, multiple scattering, flux self-shielding and gamma-ray self-absorption were corrected. The excitation function of the ²³⁸U(n,γ)²³⁹U reaction was also calculated using the TALYS-1.9 code. The experimental results were compared with the evaluated data, the theoretical data and the previous experimental data.

     

Estimation of the possibilities of a neutron generator-based source of thermalized neutrons for activation analysis

Characteristics of a source of thermal neutrons based on an evacuated NG-400 neutron generator with the maximum flux (Φ_f) 2 × 10¹¹ neutron/s for 14 MeV neutrons and 2 × 10⁹ neutrons/s for 3 meV neutrons have been investigated. The possibilities of its application for neutron activation analysis have been estimated. The distribution, composition, and density (φ_T) values of the thermal neutron flux have been measured in the inner cavity of the moderator using activation detectors. φ_T was 2 × 10⁸ and 2 × 10⁶ neutrons/cm² s for thermalized neutrons with energies of 14 and 3 MeV, respectively. The possibilities of the apparatus have been estimated theoretically and experimentally for the cases of thermalized neutrons of 14 MeV and 3 MeV.     

A low cost neutron capture prompt gamma-ray analysis facility at a research reactor

A low cost neutron capture prompt gamma activation analysis facility has been constructed at The University of Michigan's Pheonix Memorial Laboratory. Although the neutron beam used has a fairly large epithermal component (Cd ratio 7.1), background levels are low enough to result in satisfactory measurement of over 16 different elements. For the elements of greatest sensitivity (samarium, boron, gadolinium, and cadmium) minimum detectable levels of 3.6·10⁻⁵ to 1.4·10⁻⁵ gram for a one hour measurement are possible. The fast neutrons incident to the detector were found to be minimal. Estimates of up to 3 years of continuous operation before measurable damage is expected.     

Neutron damage of hexagonal boron nitride: h-BN

Hexagonal boron nitride (h-BN) was neutron damaged at an integral flux of 2.40 × 10¹² n cm⁻² s⁻¹ for 1, 2, 3 and 4 h. The h-BN samples undergo a transition from sp² to sp³ hybridization as a consequence of the neutron induced damage with the formation of cubic boron nitride (c-BN) spots, as suggested both by FT–IR and Raman spectroscopy. In addition to c-BN, also a certain degree of amorphization is achieved by h-BN already at the lowest neutron fluence of 8.64 × 10¹⁵ n cm⁻² as clearly evidenced by Raman spectroscopy. The Wigner or stored energy to the radiation-damaged h-BN samples was studied by DSC and also in this case there was a clear evidence that the neutron damage was partly irreversible and insensitive to the thermal annealing up to 630 °C. Electron spin resonance (ESR) was employed to further study the structural defects induced by the neutron bombardment of h-BN. Two kinds of paramagnetic defective structures centered on ¹¹B atoms were identified.

     

Experimental sensitivities for 3 MeV neutron activation analysis

The experimental sensitivity for 72 different elements using 3 MeV neutron activation has been investigated. Using a 200 kV Cockcroft-Walton neutron generator with a 3 MeV neutron flux of about 1.5·10⁶n·cm⁻²·sec⁻¹, γ-ray spectra of 51 elements were obtained with a sufficient number of photopeak counts for sensitivity calculations using a photopeak integration method. A useful table summarizing the sensitivity results is given. That 3 MeV neutron activation analysis is practical, is demonstrated by the experimental sensitivities obtained. Guest worker from the Institute of Nuclear Techniques, Academy of Mining and Metallurgy, Krakow, Poland, at the National Bureau of Standards, 1968–1969.     

Design and construction of a facility for neutron activation analysis using the 14 MeV neutron generator at HMS Sultan

The potential for using a small, sealed tube, DT neutron generator for neutron activation analysis has been well documented but not well demonstrated, except for 14 MeV activation analysis. This paper describes the design, construction and characterization of a neutron irradiation facility incorporating a small sealed tube DT neutron generator producing 14 MeV neutrons with fluence rates of 2·10⁸ s⁻¹ in 4π (steady state) and 10¹¹ s⁻¹ in 4π (pulsed). Monte Carlo modeling using MCNP4c and McBend9 has been used to optimize the design of this facility, including the location of a thermal irradiation facility for conventional neutron activation analysis. A significant factor in designing the facility has been the requirement to conform with Ionising Radiation Regulations and the design has been optimized to keep potential radiation doses to less that 1 μSv/h at the external walls of the facility. Activation of gold foils has been used for flux characterization and the experimental results agree well with the modeling.     

A borehole sonde using neutron activation technique

A simulated borehole sonde has been assembled, with an aluminium casing of 70 cm in length and 12 cm in inner diameter. It contains a 5 Ci Pu–Be source with a neutron yield of about 5.45·10⁶+10% n·cm^–2·s^–1, a shadow shield, and a Hp Ge detector of 15% efficiency and 2 keV FWHM for the 1.33 MeV line of⁶⁰Co. Evaluations of the assembly through the dependence of thermal neutron flux on water content, matrix composition and borehole configuration have been performed. Accordingly, thermal neutron flux distributions have been measured around the sonde and inside the ore in both the simulated dry and water filled borehole. From these measurements one could estimate that the effective moderating thickness of water is about 4 cm, while the volume matrix of the ore to be investigated by this assembly is a slab of about 8 cm width and a height of about 15 cm. It also follows that the uranium-thorium ore analysis method described in this work may become important as a field neutron activation technique.This work was financially supported by the IAEA under research project No. 3534/R.B.     

Determination of silicon in aluminium by 14 MeV neutron activation analysis

A method has been developed for determining silicon in aluminium by fast neutron activation. It is based on the separation of two gamma lines by a Ge(Li) detector: the 1.73 MeV line from the product of²⁷Al(n, α)²⁴Na and the 1.78 MeV line from the²⁸Si(n, p)²⁸Al reaction. In the case of aluminium-silicon alloys 100 μg silicon can be determined, with an error of 10% in an aluminium sample of 1 g. This work was supported in part by the International Atomic Energy Agency.     

Gamma ray spectra and sensitivities for 14 MeV neutron activation analysis

The purpose of this study was to define experimentally the sensitivity of determination for 63 different elements by 14 MeV neutron activation, with a 150 kV Cockroft-Walton accelerator at a neutron flux of 2·10⁸ n·cm⁻²·sec⁻¹ on the sample. The obtained gamma ray spectra are given, and the origin of the photopeaks observed are explained. A maximum irradiation time of five minutes was used as a convenient experimental limit to obtain the maximum sensitivity, considering, however, that the tritium target life is limited, and that the time to perform an analysis has to be reasonable. The practical use of 14 MeV neutron activation analysis is demonstrated by the detection limits obtained.     

Activation cross section measurements of 16O(n,t) above 18.1?MeV up to 33.1?MeV

We have measured the cross sections of the ¹⁶O(n,t) reactions above 18.1 up to 33.1 MeV in an neutron activation method. H₂O (water) as an ¹⁶O target was irradiated with semi-monoenergetic neutrons generated from the ⁹Be(p,n)⁹B reaction with 25–35 MeV protons. The neutron flux was obtained with the aid of previous study by Uwamino et al. (Nucl Instr Methods A 271:546, 1988). The tritium activities were measured by using the liquid scintillation counting (LSC) method. The present value for the cross section of ¹⁶O(n,t) reaction agrees with previous values measured by using the same LSC method at similar neutron energy ranging from 18.1 up to 33.1 MeV.