• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1736-1746
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• 2021

The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion-fission reactor.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.262-268
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• 2022

This paper introduces a novel concept of the pulsed neutron facility (PNF) for maximizing the production of the thermal neutrons and its application to medical use based on prompt gamma neutron activation analysis (PGNAA) using Monte Carlo simulations. The PNF consists of a compact D-T neutron generator, a graphite pile, and a detection system using Cadmium telluride (CdTe) detector arrays. The configuration of fuel pins in the graphite monolith and the design and materials for the moderating layer were studied to optimize the thermal neutron yields. Biological samples - normal and cancerous breast tissues - including chlorine, a trace element, were used to investigate the sensitivity of the characteristic γ-rays by neutron-trace material interactions and the detector responses of multiple particles. Around 90 % of neutrons emitted from a deuterium-tritium (D-T) neutron generator thermalized as they passed through the graphite stockpile. The thermal neutrons captured the chlorines in the samples, then the characteristic γ-rays with specific energy levels of 6.12, 7.80 and 8.58 MeV were emitted. Since the concentration of chlorine in the cancerous tissue is twice that in the normal tissue, the count ratio of the characteristic g-rays of the cancerous tissue over the normal tissue is approximately 2.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.402-409
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• 2024

High energy neutron irradiations impact on structural and electrical properties of alumina are studied with particular emphasis on real time in-situ radiation induced conductivity measurement in low flux region. Polycrystalline Al₂O₃ samples are subjected to high energy neutrons produced from D-T neutron generator and Am-Be neutron source. 14 MeV neutrons from D-T generator are chosen to study the role of fast neutron irradiation in the structural modification of samples. Real time in-situ electrical measurement is performed to investigate the change in insulation resistance of Al₂O₃ due to radiation induced conductivity at low flux regime. During neutron irradiation, a significant transient decrease in insulation resistance is observed which recovers relative higher value just after neutron exposure is switched off. XRD results of 14 MeV neutron irradiated samples suggest annealing effect. Impact of relatively low energy neutrons on the structural properties is also studied using Am-Be neutrons. In this case, clustering is observed on the sample surface after prolonged neutron exposure. The structural characterizations of pristine and irradiated Al₂O₃ samples are performed using XRD, SEM, and EDX. The results from these characterizations are analysed and interpreted in the manuscript.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2678-2686
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• 2023

This work demonstrates the results of full-scale numerical experiments of a hybrid thorium-containing fuel plant operating in a state close to critical due to a controlled source of D-T neutrons. The proposed facility represented a level of generated power (~10-100 MW_t) in a small pilot. In this work, the simulation of the D-T neutron plasma source operation in conjunction with the facility blanket was performed. The fission of fuel nuclei and the formation of spatial-energy release were studied in this simulation, in pulsed and stationary modes of the facility operation. The optimization results of neutronic and fluid dynamics studies to level the emerging offsets of the radial energy formed in the volume of the facility multiplying part due to the pulsed operation of the D-T neutron plasma source were presented. The results will be useful in improving the power control-based subcriticality monitoring method in coupled systems of the "pulsed neutron source-subcritical fuel assembly" type.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.299-304
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• 2008

A detection system was set up to measure the neutron generation rate of a recently developed D-D neutron generator. The system is composed of a Si detector, He-3 detector, and electronics for pulse height analysis. The neutron generation rate was measured by counting protons using the Si detector, and the data was crosschecked by counting neutrons with the He-3 detector. The efficiencies of the Si and He-3 detectors were calibrated independently by using a standard alpha particle source $^{241}Am$ and a bare isotopic neutron source $^{252}Cf$, respectively. The effect of the cross-sectional difference between the D(d,p)T and $D(d,n)^3He$ reactions was evaluated for the case of a thick target. The neutron generation rate was theoretically corrected for the anisotropic emission of protons and neutrons in the D-D reactions. The attenuations of neutron on the path to the He-3 detector by the target assembly and vacuum flange of the neutron generator were considered by the Monte Carlo method using the MCNP 4C2 code. As a result, the neutron generation rate based on the Si detector measurement was determined with a relative uncertainty of ${\pm}5%$, and the two rates measured by both detectors corroborated within 20%.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.167-175
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• 2001

Semiconductor diode detectors coated with neutron reactive material are presently under investigation for various uses, such as remote sensing of thermal neutrons, fast neutron counting, and thermal neutron radiography. Theory indicates that single-coated devices can yield thermal neutron efficiencies from 4% to 11 %, which is supported by experimental evidence. Radiation endurance measurements indicate that the devices function well up to a limiting thermal neutron fluence of $10^{13}/cm^2$, beyond which noticeable degradation occurs. Thermal neutron contrast images of step wedges and simple phantoms, taken with dual in-line pixel devices, show promise for thermal neutron imaging detectors.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2662-2669
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• 2023

For the precise measurements of special nuclear materials (SNM) including Pu and Am isotopes, we have used phoswich detector combination of two single crystal scintillators of Gd₃Ga₃Al₂O₁₂:Ce and CsI:Tl. High detection efficiency and sensitivity along with high figure of merit for the discrimination of these phoswich detectors ensures the detection and discrimination of thermal neutrons and gammas from spontaneous fission of Pu and other isotopes in presence of high gamma background. Using this detector, the low energy gammas, which is stopped completely in 1mm thick disc of GGAG, can be also discriminated from high energies gamma and shows linearity in wide range of sample quantities. By changing only the appropriate shielding, the similar setup was used for thermal neutron detection and shows a very good linearity over wide range. The quantity of a test sample was also calculated accurately by using the measured calibrated plot.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4002-4018
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• 2023

An integrated neutron interrogation system has been developed for non-destructive assay of highly-radioactive special nuclear materials, to accumulate knowledge of the method through developing and using it. The system combines a differential die-away (DDA) measurement system for the quantification of nuclear materials and a prompt gamma-ray analysis (PGA) system for the detection of neutron poisons which disturb the DDA measurements; a common D-T neutron generator is used. A special care has been taken for the selection of materials to reduce the background gamma rays produced by the interrogation neutrons. A series of measurements were performed to test the basic performance of the system. The results show that the DDA system can quantify plutonium of as small as 20 mg and it is not affected by intense neutron background up to 1.57 × 10⁷ s^-1 and gamma ray of 4.43 × 10¹⁰ s^-1. The gamma-ray background counting rate at the PGA detector was reduced down to 3.9 × 10³ s^-1 even with the use of the D-T neutron generator. The test measurements show that the PGA system is capable of detecting 0.783 g of boron and about 86.8 g of gadolinium in 30 min.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.92-96
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• 1986

The $^{93}Nb(n,n\alpha)^{89m}Y$, $^{93}Nb(n,{\alpha})^{90m}Y$ and $^{93}Nb(n,2n)^{92m}Nb$ cross sections at a neutron energy of 14.6 MeV have been measured relative to the $^{27}Al(n,p)^{27}Mg$ and $^{27}Al(n,{\alpha})^{24}Na$ cross sections. A small accelerator utilizing $T(D,n)^4He$ reaction was used as a neutron source and the neutron energy spread is about 0.4MeV at the sample. All induced activities were measured with a 70cc HPGe detector in the same geometry.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1444-1450
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• 2019

The fast-neutron shielding behaviour (FNSB) of two clay-materials (Ball clay and Kaolin)of Southwestern Nigeria ($7.49^{\circ}N$, $4.55^{\circ}E$) have been investigated using effective removal cross section, ${\Sigma}_R(cm^{-1})$, mass removal cross section, ${\Sigma}_{R/{\rho}}(cm^2g^{-1})$ and Mean free path, ${\lambda}$ (cm). These parameters decide neutron shielding behaviour of any material. A computer program - WinNC-Toolkit has been used for computation of these parameters. The toolkit evaluates these parameters by using elemental compositions and densities of samples. The proficiency of WinNC-Toolkit code was probe by using MCNPX and GEANT4 to model fast neutron transmission of the samples under narrow beam geometry, intending to represent the actual experimental setup. Direct calculation of effective removal cross section ($cm^{-1}$) of the samples was also carried out. The results from each of the methods for each types of the studied clay-materials (Ball clay and Kaolin) shows similar trend. The trend might be the fingerprint of water content retained in each of the samples being baked at different temperature. The compositions of each sample have been obtained by Particle-Induced X-ray Emission (PIXE) technique (Tandem Pelletron Accelerator: 1.7 MV, Model 5SDH). The FNSB of the selected clay-materials have been compared with standard concrete. The cognizance of various factors such as availability, thermo-chemical stability and water retaining ability by the clay-samples can be analyzed for efficacy of the material for their FNSB.