Performance comparison of 2.8 MeV and <Superscript>241</Superscript>Am-Be neutrons based moisture measurement setups

Performance of a ²⁴¹Am-Be neutron source-based and 2.8 MeV neutrons-based moisture measurement setups have been compared using Monte Carlo simulation. In the setup fast neutrons transmitted through the sample were detected by a fast neutron detector, which was placed behind a massive long double truncated collimator. The setup geometry was optimized to detect maximum effect of 1–7 wt.% moisture on the neutron intensity transmitted through the sample. The yield of neutrons transmitted through concrete, coal, wood and soil samples containing 1–7 wt.% moisture was calculated for 2.8 MeV neutrons and neutrons from an ²⁴¹Am-Be source. The slopes of the fast neutron intensities transmitted through the samples vs. their moisture contents are very sensitive to the neutron energy and the sample composition. Higher slopes have been observed for the samples with larger bulk density. The slopes of fast neutron yield show dependence on the incident neutron energy. Larger slopes have been observed for neutrons with samller energy. Due to the overall large slopes of the transmitted intensity data of the samples for 2.8 MeV neutrons, it is expected to achieve better sensitivity in moisture measurements for a 2.8 MeV neutrons based moisture setup.     

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.     

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.     

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.     

Characterization of a new external neutron beam facility at the Ohio State University

A thermal neutron beam facility has been designed and implemented at the Ohio State University Research Reactor. A project is underway to construct a large vacuum chamber such that the facility could have neutron depth profiling and neutron radiography capabilities as intended. The neutron beam is extracted from the reactor through a neutron collimator emplaced in Beam Port #2. The neutron spectrum entering the neutron collimator was unfolded from foil activation analysis results and also simulated with a full reactor core model in the MCNP Monte Carlo code. The neutron collimator uses polycrystalline bismuth as a gamma ray filter and single-crystal sapphire as a fast neutron filter. The beam is defined by multiple 3.0 cm diameter apertures made of borated aluminum. Characterization of the beam was performed using foil activation to find the flux and a low-budget neutron imaging apparatus to see the beam profile. The modulation transfer function was calculated to offer insight into the resolution of the imaging system and the collimation of the beam. The neutron collimator delivers the filtered thermal neutron beam with a ~4 cm diameter and a thermal equivalent flux of (1.27 ± 0.03) × 10⁷ n/(cm²s) at 450 kW power at the end of the collimator.     

Production of monoenergetic MeV-range neutrons by <Superscript>3</Superscript>H(p,n)<Superscript>3</Superscript>He reaction

Monochromatic MeV-energy neutron source for secondary reaction was developed utilizing tritium embedded titanium (Ti-³H) thin film via ³H(p,n)³He reaction. We have measured the neutron energies and the energy spread by resonance reactions of ¹²C(n,tot) and ²⁸Si(n,tot). The available energy was within the range from 0.6 to 2.6 MeV. Energy spread was 1.6% at energy of 2.077 MeV. The flux in the beam direction was determined to be 3.76·10⁷ n/s/sr by irradiating ¹⁹⁷Au by about 2 MeV neutrons. This source was shown to be useful for measurements of nuclear data by measuring the total cross sections of neutrons on Fe and comparing these data to the data of ENDF-6.     

The neutron flux of a252Cf neutron activation analysis device using the MCNP code

This report presents results from the application of the Monte Carlo N-Particle (MCNP) computer code to the²⁵²Cf neutron activation analysis (NAA) Device in the Technical Physics Institute of the Heilongjiang Science Academy of the People's Republic of China. The thermal and epithermal neutron flux at the sample positions and the neutron and photon fluxes on the surfaces of the device were calculated. A comparison between the calculated and experimental thermal and epithermal neutron fluxes at sample positions yield relative errors of less than 10% for the thermal neutron flux.     

Half-life of <Superscript>97</Superscript>Zr,re-measured

Summary The half-life of ⁹⁷Zr, used for the calculation of thermal/epithermal neutron flux ratio in k₀-NAA, is re-determined using three measurement systems with different pulse processing principles. The result of 16.755±0.013 hours clarifies the discrepancy between two widely used literature values, 16.744±0.011 and 16.90±0.05 hours. Different dead-time correction methods used on various measurement systems are evaluated. Factors influencing precise measurement of relative peak counting rates are discussed in time-series measurements over a dynamic range of 1000-fold radioactive intensities (10 half-lives).     

An241Am−Be source based thermal neutron activation analysis facility at KFUPM

An instrumental thermal neutron activation analysis facility based, on a 16 Ci²⁴¹ Am–Be source, a high resolution -ray spectrometry setup and a PC-based data acquisition system at KFUPM is described. The thermal neutron flux distribution was determined from the induced activities of high purity indium foils. The absolute thermal neutron flux was calculated from the activities of bare and cadmium-covered gold foils at a position of 3 cm from the soource at which the flux reaches a maximum. The facility tests were carried out with the determination of manganese concentrations in six types of industrially important steel samples. The result of 1.33% manganese in SS-304 steel sample was in excellent agreement with the literature value. The method is nondestructive, economical and ideal for bulk analysis.     

Measurement of <Superscript>99</Superscript>Mo production cross-section from the <Superscript>100</Superscript>Mo(n,2n) reaction with quasi monoenergetic neutron based on the <Superscript>9</Superscript>Be(p,n) reaction

The production cross-section of the medical isotope, ⁹⁹Mo from the enriched ¹⁰⁰Mo(n,2n) reaction with the average neutron energies of 21.9 and 26.5 MeV has been determined for the first time by using an off-line γ-ray spectrometric technique. The average neutron energies were generated by using the ⁹Be(p,n) reaction with the proton energies of 35 and 45 MeV from the MC50 cyclotron of the Korea Institute of Radiological and Medical Sciences (KIRAMS) at Seoul, South Korea. The ¹⁰⁰Mo(n,2n) reaction cross-section as a function of neutron energy was also calculated theoretically by using the computer code TALYS-1.8 and EMPIRE-3.2 Malta. The experimental results are in close agreement with the theoretical values from TALYS-1.8. However, the present data at the neutron energy of 21.9 MeV is slightly lower and at 26.5 MeV is higher than the values from EMPIRE-3.2 Malta.     

Design and evaluation of a TNA explosive-detection system to screen carry-on luggage

Thermal neutron analysis (TNA) technology has been used for the non-destructive detection of explosives. The system uses a relatively weak ²⁵²Cf neutron source (1.03·10⁷ n/s) and two 3"×3" NaI(Tl) detectors. The presence of explosives is confirmed via detection of the 10.83 MeV prompt gamma-ray associated with nitrogen decay. The MCNP4A code was used to simulate the neutron and gamma transport through the system. The thermal neutron flux in the activation position was measured using gold and indium foils. The measured thermal neutron flux was lower, by not more than 9.5%, than that of simulation. In this report the results of the preliminary tests on the system are described.     

Evaluation of imaging plate technique coupled with activation detector as the passive neutron monitor

The spatial distribution of neutrons was measured at the muon science laboratory of KEK by the activation detector method using an imaging plate for the radioactivity measurement. It was confirmed that this method is highly sensitive to detect the average neutron dose of 10 μSv/h. The distribution of thermal and epithermal neutrons was also measured in the experimental room. The cadmium ratio inside the experimental room is one except for the neutron leakage point. The spatial distribution of neutrons inside the concrete shield of KENS was measured by the same method. Aluminum and gold foils were used for the measurement of fast and thermal neutrons, respectively. Two dimensional change of the reaction rate of the ²⁷Al(n,α)²⁴Na reaction shows a good agreement with the results calculated by the Monte Carlo simulation using MARS14 code. Thermal and epithermal neutron flux ratio on the beam axis was measured by the cadmium ratio method. The flux ratios were about 30 and almost constant for every slot except for the surface of the shield, because the cadmium ratio is 2. This method was very useful to measure the activity of many pieces of detector simultaneously without any efficiency and decay correction. Wide dynamic range and high sensitivity are also the merit of this method.     

Monte Carlo simulation of PGNAA system for determining element content in the rock sample

The element content in rock sample can be determined by prompt gamma ray activation analysis technology. The neutron distributions under the conditions with different moderating materials, moderator size and distance from neutron source to lead-out hole were simulated using Monte Carlo method, and then the optimal structure parameters to get the highest thermal neutron flux was obtained. The PGNAA system with optimal parameters based on ²⁵²Cf neutron source was founded. In addition, the rock and element standard samples were irradiated by thermal neutron in this system. Moreover, the element content was calculated by processing gamma ray spectroscopy recorded, and it is in agreement with result of XRF method.     

Study of the <Superscript>241</Superscript>Am(n,2n)<Superscript>240</Superscript>Am reaction cross section in the energy range of <Emphasis Type="Italic">E</Emphasis><Subscript><Emphasis Type="Italic">n</Emphasis></Subscript> = 8.8−11.1 MeV

The measurement of the cross section of the reaction ²⁴¹Am(n,2n)²⁴⁰Am has been performed at neutron energies from 8.8 to 11.1 MeV, implementing the activation technique. The neutron beam was produced at the TANDEM accelerator of NCSR “Demokritos” by the ²H(d,n)³He reaction, using a deuterium gas target. During the 5-day long irradiation, the neutron beam fluctuations were monitored in 100 seconds intervals by a BF₃ counter connected with a multiscaling unit. The radioactive target consisted of a 37 GBq ²⁴¹Am source enclosed in a Pb container. A natural Au foil, a ²⁷Al foil and a ⁹³Nb foil were used as reference materials for the neutron flux determination. After the end of the irradiation the activity induced at the target and the reference foils, was measured off-line by a 56% HPGe detector.     

Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions.     

Cyclotron production and parameters calculation of <Superscript>48</Superscript>V Nitinol stent for renal arteries in brachytherapy

The Nitinol stent was bombarded in a cyclotron at a flux rate of 4 μA/cm² to produce ⁴⁸V via ⁴⁸Ti (p, n) ⁴⁸V reaction. In this study dose distribution of ⁴⁸V radioactive stent was investigated for renal arteries. Version X-2.6 of the MCNP Monte Carlo radiation transport system code was employed to calculate dose distribution around the stent. As ⁴⁸V is a mixed gamma and beta particle emitter, two separate runs of MCNP for both beta and gamma particles were performed and the total deposited dose was acquired by adding the two mentioned values. In order to verify the simulation, the calculated results have been compared with previous published data for the source. Calculated results show high dose gradient near the stent and the maximum amount of dose deposits at the vessel wall. According to (AAPM) TG-60/149 protocol, the dosimetric parameters, including geometry function, G(ρ,z), radial dose function, g _L (ρ), and anisotropy function, F(ρ,z), were also determined.     

Radioanalytical prediction of radiative capture in 99Mo production via transmutation adiabatic resonance crossing by cyclotron

In this study, the transmutation adiabatic resonance crossing (TARC) concept was estimated in ⁹⁹Mo radioisotope production via radiative capture reaction in two designs. The TARC method was composed of moderating neutrons in lead or a composition of lead and water. Additionally, the target was surrounded by a moderator assembly and a graphite reflector district. Produced neutrons were investigated by (p,xn) interactions with 30 MeV and 300 μA proton beam on tungsten, beryllium, and tantalum targets. The ⁹⁹Mo production yield was related to the moderator property, cross section, and sample positioning inside the distinct region of neutron storage as must be proper to achieve gains. Gathered thermal flux of neutrons can contribute to molybdenum isotope production. Moreover, the sample positioning to gain higher production yield was dependent on a greater flux in the length of thermal neutrons and region materials inside the moderator or reflector. When the sample radial distance from Be was 38 cm inside the graphite region using a lead moderator design, the production yield had the greatest value of activity, compared with the other regions, equal to 608.72 MBq/g. Comparison of the two designs using a Be target revealed that the maximum yield occurred inside the graphite region for the first design at 38 cm and inside the lead region for the second design at 10 cm. The results and modeling of the new neutron activator were very encouraging and seem to confirm that the TARC concept can be used for ⁹⁹Mo production in nuclear medicine.     

Design and simulation of neutron radiography system based on 241Am–Be source

Neutron imaging is extended rapidly as a means of non-destructive testing (NDT) of materials. Various effective parameters on the image quality are needed to be studied for neutron radiography system with good resolution. In the present study a portable system of neutron radiography has been designed using ²⁴¹Am–Be neutron source. The effective collimator parameters were calculated to obtain relatively pure, collimated and uniform neutron beam. All simulations were carried out in two stages using MCNPX Monte Carlo code. In the first stage, different collimator configurations were investigated and the appropriate design was selected based on maximum intensity and uniformity of neutron flux at the image plane in the outlet of collimator. Then, the overall system including source, collimator and sample was simulated for achieving radiographic images of standard samples. Normalized thermal neutron fluence of 2.61×10^?5 cm^?2 per source particle with n/γ ratio of 1.92×10⁵ cm^?2 μSv^?1 could be obtained at beam port of the designed collimator. Quality of images was assessed for two standard samples, using radiographic imaging capability in MCNPX. The collimated neutron beam in the designed system could be useful in a transportable exposure module for neutron radiography application.     

Development and implementation of <Emphasis Type="Italic">k</Emphasis><Subscript>0</Subscript>-INAA standardization at PINSTECH

The k ₀ method has been introduced at the 30 kW miniaturized neutron source reactor (MNSR) at the Pakistan Institute of Nuclear Science & Technology (PINSTECH). It involved the full energy peak efficiency calibration of the HPGe detector for different counting geometries and the characterization of the neutron flux at four inner irradiation channels. The latter involved the determination of the thermal to the epithermal flux ratio, epithermal flux shape factor, the modified spectral index, Westcott’s g-factor, the Maxwellian neutron temperature and the fast flux. The method was validated by analyzing IAEA-SL1 (Lake Sediment) and NIST-SRM-1572 (Citrus Leaves) reference materials. All calculations were performed in Excel, including the optimization step. The results revealed that most of the elements were estimated with less than 10% relative deviation from the certified value.     

Development of <Superscript>177</Superscript>Lu-DOTA-anti-CD20 for radioimmunotherapy

Rituximab was successively labeled with ¹⁷⁷Lu-lutetium chloride. ¹⁷⁷Lu chloride was obtained by thermal neutron flux (4 × 10¹³ n cm⁻² s⁻¹) of natural Lu₂O₃ sample with a specific activity of 2.6–3 GBq/mg. The macrocyclic bifunctional chelating agent, N-succinimidyl-1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid (DOTA-NHS) was prepared at 25 °C using DOTA, N-hydroxy succinimide (NHS) in CH₂Cl₂. DOTA-rituximab was obtained by the addition of 1 mL of a rituximab pharmaceutical solution (5 mg/mL, in phosphate buffer, pH 7.8) to a glass tube pre-coated with DOTA-NHS (0.01–0.1 mg) at 25 °C with continuous mild stirring for 15 h. Radiolabeling was performed at 37 °C in 24 h. Radio-thin layer chromatography showed an overall radiochemical purity of >98% at optimized conditions (specific activity = 444 MBq/mg, labeling efficacy; 82%). The final isotonic ¹⁷⁷Lu-DOTA-rituximab complex was checked by gel electrophoresis for structure integrity control. Radio-TLC was performed to ensure that only one species was present after filtration through a 0.22 μm filter. Preliminary biodistribution studies in normal rats were carried out to determine complex distribution of the radioimmunoconjugate up to 168 h. The biodistribution data were in accordance with other antiCD20 radioimmunoconjugates already reported.