Neutron moderation effects in phc contaiminated soil samples

Prompt gamma ray analysis of soil samples contaminated with petroleum hydrocarbons(PHC) like benzene was carried out using 14 MeV neutron beams. Intensities of silicon, carbon, oxygen, and hydrogen gamma rays were measured for soil samples containing 2.20–10.4 wt% benzene. With increasing benzene concentration, the intensity of the C gamma rays increased while those of Si, H, and O gamma rays decreased. The reduction in Si, H, and O gamma ray intensities may be due to increasing neutron moderation effect in benzene-contaminated soil samples. The experimental intensities of gamma rays are in good agreement with the calculated intensities.The neutron moderation effects in benzene contaminated soil samples are about 26% weaker than those reported for soil samples containing moisture. From the slopes of silicon gamma ray intensity as a function of benzene concentration as well as moisture concentration, a simple scheme has been suggested to correct for the loss in carbon counts caused by neutron moderation from PHC and moisture in the soil samples.     

Utilization of pneumatic carrier facility of Dhruva reactor for trace element determination by neutron activation analysis

The pneumatic carrier facility (PCF) of Dhruva reactor is being extensively used for neutron activation analysis (NAA) studies pertaining to research work as well as routine sample analysis. It is useful for the determination of trace elements using short and medium half-lives radioisotopes produced in neutron activation with available higher neutron flux (~5 × 10¹³ cm^?1 s^?1). Solid samples placed in high density polypropylene capsule, are irradiated for 1 min duration and radioactive assay is carried out by high resolution gamma ray spectrometry. Design aspects of PCF and various applications to samples of diverse matrices using NAA are presented.     

Study of naturally occurring radioactive material (NORM) in top soil of Punjab State from the North Western part of India

The concentration levels of ²³⁸U, ²³² Th, ⁴⁰K and ¹³⁷Cs in top soils of State of Punjab located in the North Western part of India were measured using conventional low background gamma ray spectrometric setup as well as Compton suppressed gamma ray spectrometric setup. The radioactivity level of ²³⁸U and ²³²Th was found to vary between 15 Bq/kg and 27 Bq/kg and between 16 Bq/kg and 57 Bq/kg respectively. The radioactivity level of ⁴⁰K was found to vary between 266 Bq/kg and 799 Bq/kg. The mean radioactivity level of the NORM in general was found to be similar to what is expected as a result of their normal abundance.     

Identifying and quantifying short-lived fission products from thermal fission of HEU using portable HPGe detectors

Due to the emerging potential for trafficking of special nuclear material, research programs are investigating current capabilities of commercially available portable gamma ray detection systems. Presented in this paper are the results of three different portable high-purity germanium (HPGe) detectors used to identify short-lived fission products generated from thermal neutron interrogation of small samples of highly enriched uranium. Samples were irradiated at the Washington State University Nuclear Radiation Center’s 1 MW TRIGA reactor. The three portable, HPGe detectors used were the ORTEC MicroDetective [1], the ORTEC Detective [2], and the Canberra Falcon [3]. Canberra’s GENIE-2000 software was used to analyze the spectral data collected from each detector. Ultimately, these three portable detectors were able to identify a large range of fission products showing potential for material discrimination.     

Design,construction and characterization of a prompt gamma activation analysis facility at the Oregon State University TRIGA<Superscript>®</Superscript> reactor

A prompt gamma neutron activation analysis facility has been designed, built, and characterized at the Oregon State University TRIGA^® reactor. This facility was designed for versatile multi-elemental analyses. The facility utilizes the leakage neutrons originating from beam port #4 of the Oregon State University TRIGA^® reactor. The neutrons are collimated through a series of lead and Boral^® collimators, and filtered through both a bismuth filter and single-crystal sapphire. Samples are irradiated in a sample chamber outside the biological shielding of the reactor, and the resulting gamma radiation produced from neutron interactions within the sample is monitored using a high-purity germanium detector (HPGe). The thermal and epithermal neutron fluxes were measured using gold-foil irradiations and found to be 2.81 × 10⁷ and 1.70 × 10⁴ cm^?2 s^?1, respectively. The resulting cadmium ratio was 106. Measured detection limits for boron, chlorine, and potassium in a NIST SRM 1571 orchard leaf were 5.6 × 10^?4 mg/g, 8.2 × 10^?2 mg/g, and 1.0 mg/g, respectively. Detection limits for additional elements and samples are presented.     

Determination of zinc in geological samples using Compton suppression with thermal and epithermal instrumental neutron activation analysis

The determination of Zn in geological samples using instrumental neutron activation analysis is usually done using the ⁶⁴Zn(n,γ)⁶⁵Zn reaction and its 244 day half-life. However this analysis has proven to be potentially difficult. This is due to its relatively low neutron absorption cross section and gamma ray intensity, and the relatively high neutron absorption cross section and gamma intensity of ⁴⁶Sc, which has an energy peak that is only 5 keV greater than ⁶⁵Zn. The use of a high resolution detector makes it possible to differentiate between the ⁶⁵Zn and ⁴⁶Sc photopeaks peaks. However, the dominating ⁴⁶Sc gamma ray can even make peak fitting routines unsuccessful in the proper determination of ⁶⁵Zn. The use of a Compton suppression system suppresses the ⁴⁶Sc peak, which has two coincident gamma-rays, and this greatly improves the ratio of the height of the ⁴⁶Sc 1120.5 keV photopeak to the ⁶⁵Zn 1115.4 keV photopeak. Irradiating the sample with epithermal neutrons also improves the measurement since ⁶⁵Zn has a higher cross section for epithermal neutrons rather than thermal neutrons, whereas ⁴⁶Sc has a higher thermal cross section. Another technique to determine zinc is the use of ⁶⁸Zn(n,γ)^69mZn reaction with its 13 h half-life using epithermal neutrons and Compton suppression INAA. However, the 438 keV gamma ray of ^69mZn has no interference with any adjoining photopeak. A critical comparison of these two methods is given.     

Evaluation of an alternative convenient irradiation system for determination of emission rate of radio-isotopic neutron sources

In present work, an alternative irradiation system based on a symmetric cylindrical tank filled with a moderator containing hydrogen, which was equipped with a NaI(Tl) scintillation detector, was proposed for using in determination of neutron flux. This irradiation system was designed by MCNP4C code, with considering a ²⁴¹Am–Be neutron source in several volumes and different materials. When the neutron is captured by hydrogen, a 2.22 MeV prompt gamma-ray is emitted. The gamma pulse-height spectrum shows a photo-peak around 2.22 MeV whose net area is proportional to the total emission rate of neutron. The simulation result showed that a cylindrical tank with 110 cm diameter and height filled with water can be a suitable system for neutron source strength calibration. Furthermore, a proper two-layer shielding must be placed between the source and detector for preventing neutrons and gamma rays to directly enter the detector.     

Evaluation of Gd,Sm and Eu in Egyptian monazite samples using the prompt neutron capture gamma-ray technique

Three Egyptian monazite samples from the beach black sand at Abou Khashba near Rosetta, submitted by the laboratories of the nuclear materials authority of Egypt were analyzed for elemental contents, using Prompt gamma ray neutron activation analysis (PGNAA) as an effective analysis technique especially for rare earth elements. An integrated system for PGNAA, located at hot Lab. Center, has proven useful for the analytical purpose using 252Cf neutron source with neutron flux of 6.16E7 n/cm² s, that system has been designed and calibrated for PGNAA. Sensitivity curves using different standard concentration values of artificial liquid standards samples at different concentrations ranging from 1,000 to 5,000 mg/L in case of Sm and Gd, while from 100 to 1,000 mg/L for Eu, have been constructed. The average concentration values for Gd, Sm and Eu in the samples under investigation was at 1.46E4 ± 0.21, 6.683E4 ± 0.14 and 440 ± 0.18 ppm, respectively. A comparative study of the obtained results and the results of ICP-Ms was given.     

Analysis of SMELS by k 0-based IM-NAA method using PFTS position of KAMINI reactor for quality control exercise

As a part of QA/QC of k ₀-based internal monostandard neutron activation analysis (IM-NAA), three types of synthetic multielement standards (SMELS) were analyzed using pneumatic fast transfer system irradiation position of KAMINI reactor, IGCAR. Radioactive assay of activation products was carried out by high resolution gamma ray spectrometry. IM-NAA was used to determine relative concentration ratios of 22 elements with respect to gold internal monostandard. Absolute concentrations were calculated using assigned concentration of Au in all the types of SMELS. Z-score values within ±1 at 95.5 % confidence level and percentage deviations within ±5 % indicated good quality of the results by IM-NAA in most of the cases. Using this methodology, an ilmenite mineral sample was analyzed and concentrations of 14 elements were determined using Sc as monostandard.     

Application of k 0-based internal monostandard NAA for large sample analysis of clay pottery: as a part of inter comparison exercise

As a part of inter comparison exercise of an IAEA Coordinated Research Project on large sample neutron activation analysis, a large size and non standard geometry size pottery replica (obtained from Peru) was analyzed by k ₀-based internal monostandard neutron activation analysis (IM-NAA). Two large size sub samples (0.40 and 0.25 kg) were irradiated at graphite reflector position of AHWR Critical Facility in BARC, Trombay, Mumbai, India. Small samples (100–200 mg) were also analyzed by IM-NAA for comparison purpose. Radioactive assay was carried out using a 40 % relative efficiency HPGe detector. To examine homogeneity of the sample, counting was also carried out using X–Z rotary scanning unit. In situ relative detection efficiency was evaluated using gamma rays of the activation products in the irradiated sample in the energy range of 122–2,754 keV. Elemental concentration ratios with respect to Na of small size (100 mg mass) as well as large size (15 and 400 g) samples were used to check the homogeneity of the samples. Concentration ratios of 18 elements such as K, Sc, Cr, Mn, Fe, Co, Zn, As, Rb, Cs, La, Ce, Sm, Eu, Yb, Lu, Hf and Th with respect to Na (internal mono standard) were calculated using IM-NAA. Absolute concentrations were arrived at for both large and small samples using Na concentration, obtained from relative method of NAA. The percentage combined uncertainties at ±1 s confidence limit on the determined values were in the range of 3–9 %. Two IAEA reference materials SL-1 and SL-3 were analyzed by IM-NAA to evaluate accuracy of the method.     

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.     

Performance comparison of an 241Am-Be neutron source-based PGNAA setup with the KFUPM PGNAA setup

Monte Carlo calculations have been carried out to compare the performance of an ²⁴¹Am-Be neutron source-based prompt gamma neutron activation analysis (PGNAA) setup with that of the 2.8 MeV neutron-based PGNAA setup at King Fahd University of Petroleum and Minerals (KFUPM) to analyze Portland cement samples. This work is a part of a wide Monte Carlo studies being conducted at KFUPM in search of a more efficient neutron source for its 2.8 MeV neutrons, from the D(d,n) reaction, based PGNAA facility. In this study an ²⁴¹Am-Be neutron source-based PGNAA setup was simulated. For comparison, the diameter of a cylindrical external moderator of the ²⁴¹Am-Be neutron source, based PGNAA setup, was assumed to be similar to that used in the KFUPM PGNAA setup. The results of this study revealed that although the optimum geometry of the ²⁴¹Am-Be neutron source-based setup is similar to that of the KFUPM PGNAA facility, the performance of the ²⁴¹Am-Be neutron source-based setup is slightly poorer than that of the 2.8 MeV neutron-based setup. This revised version was published online in July 2006 with corrections to the Cover Date.     

Performance improvement of a PGNAA setup due to change in moderator design: A Monte Carlo study

Effect of neutron source-moderator geometry has been studied on the performance of a thermal neutron capture-based prompt gamma neutron activation analysis (PGNAA) setup. In the study prompt gamma-ray and thermal neutron yield was calculated for various positions of the neutron source inside 3–10 cm long high density polyethylene moderators. The study has been carried out for a Portland cement sample using Monte Carlo calculations. The maximum yields of the thermal neutrons and prompt gamma-rays have been observed for a neutron source at a distance of 1 cm from the sample. The maximum yield of the 1.94 and 6.42 MeV prompt gamma-ray from a Portland cement sample has been observed for moderators having length equal to or greater than 7 cm. The yield of both gamma rays is 2.57 times higher than the previously reported value for a PGNAA setup with the source placed outside a 5 cm thick moderator. The higher yield of gamma rays will result in higher sensitivity of the PGNAA setup.     

Determination of trace concentrations of thorium in uranium oxide matrix by epithermal instrumental neutron activation analysis

An epithermal instrumental neutron activation analysis (EINAA) method using cadmium filter was standardized to determine trace concentrations of thorium in four samples of uranium oxide (U₃O₈) samples. Samples and thorium standards, wrapped with cadmium foil, were irradiated at a reactor neutron flux of about 10¹² cm^?2 s^?1. Radioactive assay was carried out using a Compton suppressed anticoincidence gamma ray spectrometer consisting of HPGe-BGO detectors coupled to MCA. Concentrations of thorium in these samples were found to be in the range of 15–72 mg kg^?1. EINAA results were validated by determining thorium concentrations in uranium matrix by standard addition method. EINAA results were compared with those obtained by two wet chemical methods namely ion chromatography (IC) and inductively coupled plasma atomic emission spectrometry (ICP-AES). The results obtained by the three methods were found to be in good agreement, indicating further validity of the proposed EINAA method.     

Determining trace amounts of nickel in plant samples by neutron activation analysis

Neutron activation analysis of plant samples for trace amounts of nickel has applications in various fields. Landsberger and Robinson (Trans Am Nucl Soc 102:187–188, 2010) found their measurement of the concentrations of nickel for different NIST reference materials to be significantly greater than the certified values when measuring nickel from the 810.8 keV gamma ray that comes from ⁵⁸Co from the ⁵⁸Ni(n,p) reaction. They determined that this overestimation was due to a significant interference by the presence of ¹⁵²Eu at 810.5 keV, and presented a method for correcting this interference. Their method involved a long thermal irradiation and correction of gamma ray counts based on the 1,408 keV belonging to ¹⁵²Eu. This paper presents an alternative approach, which involves irradiating the samples with epithermal neutrons, which suppressed the ¹⁵²Eu to the point of being negligible for samples with low levels of europium. Both methods were determined to work well for the identification of nickel concentrations by neutron activation analysis.     

Non-destructive evaluation of concrete samples using PGNAA technique

Non-destructive evaluation of concrete is an important task of quality control in building construction industry. The quality evaluation of concrete samples was carried out using prompt gamma ray activation analysis (PGNAA) technique. In this study calcium and silicon contents of six concrete samples were determined through yield of 1.94 MeV gamma-rays from calcium and 3.54 MeV gamma-rays from silicon. The concrete samples were prepared by mixing cement, coarse and fine aggregates in different proportions. A linear correlation has been observed between the experimental yield of the prompt gamma rays and the actual calcium and silicon concentration in the concrete samples. Results of this study have demonstrated successful use of an accelerator-based PGNAA setup in non-destructive analysis of the concrete samples.     

Design, testing and optimization of a neutron radiography system based on a Deuterium–Deuterium (D–D) neutron generator

Simulations show that significant improvement in imaging performance can be achieved through collimator design for thermal and fast neutron radiography with a laboratory neutron generator. The radiography facility used in the measurements and simulations employs a fully high-voltage-shielded, axial D–D neutron generator with a radio frequency driven ion source. The maximum yield of such generators is about 10¹⁰ fast neutrons per seconds (E = 2.45 MeV). Both fast and thermal neutron images were acquired with the generator and a Charge Coupled Devices camera. To shorten the imaging time and decrease the noise from gamma radiation, various collimator designs were proposed and simulated using Monte Carlo N-Particle Transport Code (MCNPX 2.7.0). Design considerations included the choice of material, thickness, position and aperture for the collimator. The simulation results and optimal configurations are presented.     

Implementation of a methodology to analyse cylindrical 5-g sample by neutron activation technique,k 0 method,at CDTN/CNEN,Belo Horizonte,Brazil

Neutron activation analysis (NAA) is routinely applied to geometrical point-source or small samples because there are technical and theoretical difficulties to analyse larger samples weighing more than 0.5 g. The analysis of larger samples is very advantageous, because the analytical procedure will be less time consuming, it may be possible to reach lower detection limit for several elements, it decreases cost and overcomes the difficulties related to the representativeness of the sample when dealing with inhomogeneous volume or several small samples. Thus, increasing the amount of sample is a way to compensate for low flux of neutrons. This paper is about the establishment of a method at Laboratory for Neutron Activation Analysis, CDTN/CNEN, to determine the elemental concentrations in 5 g-samples, 25 times larger than usual samples analysed by neutron activation, k ₀ method, keeping the current irradiation and gamma spectrometry facilities. To develop this method, several aspects were evaluated such as detector efficiency over the volume source, neutron self-shielding during neutron irradiation, axial neutron flux gradient and gamma ray attenuation within the sample during counting. The results suggest that if an appropriate adjustment of the above mentioned parameters is done, the k ₀ method of NAA can provide satisfactory results also for larger samples than the samples typically used in NAA. The KayWin software proved to be a robust program analysing the larger samples weighing 5 g and cylindrical geometry as if it were a small cylindrical sample, producing reliable results. It was successfully implemented at Belo Horizonte, Brazil, fully following the basic principles of the k ₀ standardization method.     

Complexation of Eu3+ with cone conformers of three substituted calix[4]arenes in nitrobenzene saturated with water

From extraction experiments and $ \gamma Accurate knowledge of the trace elemental concentrations in wheat and its products is of great importance from a nutritional point of view. In this study, six wheat samples were prepared from the agriculture research center of Arak named Sardari, Amir, MV-17, Batava, Karaj-2 and Alvand; they were analyzed by neutron activation method (NAA). In this method, Isfahan miniature reactor as a neutron source and relative NAA method has been used as the analysis type. In this design in order to record gamma spectrum the MCA system and high purity germanium detector were used. Finally, the concentration of the trace elements such as Br, K and Na value was determined for the Sardari, Amir, Alvand, MV-17, Batava, Karaj-2 wheat samples. The average concentration of trace elements in all wheat samples in the studied area are 2.41(0.8597?6.1175) mg kg^?1 for Br, 13.42(8.7063?24.696) mg kg^?1 for Na and 463.30(434.22?505.45) mg kg^?1 for K, respectively. These were compared with other reports results. This study has been conducted as the first time for this region.     

Moderator for neutron activation with the photoneutrons produced by a LINAC

Using Monte Carlo methods a polyethylene moderator has been designed to induce activation using the photoneutrons field of a 15 MV linear accelerator for radiotherapy. In the calculations the photoneutron spectrum at 1 m from the isocenter was used as a source term and the neutron spectra were calculated in the center of different size cylindrical moderators. The best size was selected defining the thermal-to-fast-neutron ratios as a figure of merit. The moderator was built and its performance was evaluated by inducing the activation of Mn dissolved in water, silver coins and souvenir coin. The thermal neutron fluence rate was determined with the Mn samples being 9.96 × 10⁵ cm^?2 Gy _x ^?1 .