Portable Neutron Research Articles

Pillar-structured neutron detector based multiplicity system

This work demonstrates the potential of silicon pillars filled with boron-10 as a sensor technology for a compact and portable neutron multiplicity system. Solid-state, semiconductor based neutron detectors may enable completely new detector form factors, offer an alternate approach to helium-3 based systems, and reduce detector weight and volume requirements. Thirty-two pillar-structured neutron detectors were assembled into a system with an active area of over 20 cm2 and were used in this work to demonstrate the feasibility of this sensor technology as a potential replacement for helium-3 based gas detectors. Multiplicity measurements were successfully carried out using a californium-252 neutron source, in which the source mass, system efficiency, and die-away time were determined. This demonstration shows that these solid-state detectors could allow for a more compact and portable system that could be used for special nuclear material identification in the field.

Design of a novel nondestructive portable mobile neutron activation system

A new non-destructive portable mobile system was designed to be used for neutron activation assay according to the Safeguards purposes. The sample is irradiated in a neutron field and its composition is determined by identifying the characteristic induced gamma radiation emitted by the fission and activation products. For the design of a neutron activation system, many aspects must be considered such as choosing the best materials to assure the protection against radiation and estimating the thickness required to shield the neutron source based on its flux or strength. The polyethylene and paraffin wax were selected as neutron shielding material, lead being chosen as gamma shielding material. Several calculations were performed using the Monte Carlo code MCNP5 to estimate the required shield thickness and distribution of thermal neutron flux at different irradiation position inside the shield. The available 252Cf neutron source with 12µCi of activity was used. The novel neutron activation analysis system was proposed, designed and fabricated to meet the maximum flexibility for neutron activation analysis, at minimum cost and with adequate optimized shielding and portability.

Optimization of Beam Shaping Assembly design for Boron Neutron Capture Therapy based on a transportable proton accelerator

The optimization of a Beam Shaping Assembly design for Boron Neutron Capture Therapy (BNCT), based on a commercial available proton accelerator, has been simulated. The primary goal is to improve the flux of the epithermal neutrons, which is the major drawback in all designed BNCT facilities with portable neutron source, satisfying simultaneously all the other relevant parameters.A BNCT facility has been simulated using the MCNP4B Monte Carlo code. A wide set of material were examined as moderators, while three different materials are used as fast neutron, gamma ray and thermal neutrons filters.Fast neutrons from 9Be(p,n)9B reaction are moderated through the 7LiF and TiF3 moderators and further reduced using 60Ni as fast neutron filter. The undesired gamma rays and thermal neutrons are removed using Bi and Cd filters respectively. The therapeutic efficacy of the proposed facility was calculated through the dosimetric evaluation in a head phantom.According to the results of the calculations the proposed unit satisfies the recommended by IAEA parameters. Only the flux of the epithermal neutrons remains below the recommended threshold, but is significantly improved.

Development of a method for on-line determination of chlorine impurity in crude oil by using fast neutrons

An important challenge in the petroleum industry is to reduce the salt content to acceptable levels in crude oil. Continuous monitoring of crude oil salinity is thus an important aspect. However, development of safe and reliable on-line and non-intrusive monitoring systems is still a challenging task. In this work, a method for non-intrusive, on-line monitoring of the chlorine impurity in crude oil based on prompt gamma-ray neutron activation analysis (PGNAA) in conjunction with a portable pulsed neutron generator producing 14MeV neutrons in deuterium-tritium (DT) nuclear reactions is proposed and tested. Simulations with Monte Carlo N-Particle version 6 (MCNP6) transport code combined with Center for Engineering Applications of Radioisotopes Detector Response Function (CEARDRF) and a specific purpose Monte Carlo code called Center for Engineering Applications of Radioisotopes Coincidence Prompt Gamma-Ray (CEARCPG) were used for the design optimization of the experiments. Preliminary results indicate that a minimum detection limit for chlorine impurity of about 71mg/L is achievable for an integration time of 30min.

Development of lightweight portable neutron survey meter

ABSTRACTWe have developed a lightweight portable neutron survey meter comprising a proportional gas counter containing a mixed gas of methane gas and nitrogen gas for measuring the ambient neutron dose equivalent H*(10) up to about 20 MeV neutrons intended for use in nuclear power plants, and accelerator facilities. Since no heavy polyethylene moderator is used, the survey meter is only about 2.2 kg in weight causing a weight reduction of 70% or more compared to that of the conventional moderated-type survey meter. The spectrum-weight function, G(E) is adopted for dose conversion.In order to evaluate the energy characteristics of the developed survey meter, mono-energetic neutron reference fields and continuous energy neutron reference fields were used. The evaluation was also carried out by the PHITS calculation. Although the neutron energy response to the mono-energetic neutron fields was highly deviated from the ambient dose equivalent H*(10) in keV energy region, the response to H*(10) showed very good agreement only within 25% to the continuous energy neutron fields which are close to the actual work-place neutron fields. The neutron energy response was also investigated for high energy quasi-mono-energetic neutron fields and showed much better results compared with those of conventional moderated-type neutron survey meters.

Real-time neutron source localization and identification with a hand-held, volumetrically-sensitive, moderating-type neutron spectrometer

Measuring source-dependent properties of free neutrons over a large neutron energy range, with hand-portable instrumentation, continues to push the frontier of neutron detection instrumentation design and analysis techniques. Building on prior work – C.B. Hoshor, et al., A portable and wide energy range semiconductor-based neutron spectrometer, Nucl. Instrum. Methods Phys. Res. A 803 (2015) 68–81 – which focused on demonstrating one-dimensional-based energy-dependent neutron measurement and analysis with a new class of solid-state moderating-type spectrometer, this work introduces two “core” algorithmic methodologies that expand the analysis of neutron thermalization measurements to three spatial dimensions to determine the location and identity of neutron radiation sources in real time. Two extensions of these core methodologies are then proposed to further improve both the accuracy and reliability of source location and identity determinations with this new class of hand-held instrumentation. In 432 preliminary simulation tests, these method extensions are shown to decrease the average source location error by 64% and provide correct identity determinations in all test cases.

Compensation scheme for online neutron detection using a Gd-covered CdZnTe sensor

The development of portable and personal neutron dosimeters requires compact and efficient radiation sensors. Gd-157, Gd-155 and Cd-113 nuclei present the highest cross-sections for thermal neutron capture among natural isotopes. In order to allow for the exploitation of the low and medium-energy radiative signature of the said captures, the contribution of gamma background radiation, falling into the same energy range, needs to be cancelled out. This paper introduces a thermal neutron detector based on a twin-dense semiconductor scheme. The neutron-sensitive channel takes the form of a Gd-covered CdZnTe crystal, a high density and effective atomic number detection medium. The background compensation will be carried out by means of an identical CdZnTe sensor with a Tb cover. The setting of a hypothesis test aims at discriminating the signal generated by the signature of thermal neutron captures in Gd from statistical fluctuations over the compensation of both independent channels. The measurement campaign conducted with an integrated single-channel chain and two metal Gd and Tb covers, under Cs-137 and Cf-252 irradiations, provides first quantitative results on gamma-rejection and neutron sensitivity. The described study of concept gives grounds for a portable, online-compatible device, operable in conventional to controlled environments.

Experimental Investigation of Ion Sources of Gas-Filled Neutron Tubes

The results of experimental studies of discharge generation regimes in ion sources of gas-filled neutron tubes are presented. The objects of investigation are neutron tubes developed at the Dukhov All-Russia Research Institute of Automatics (VNIIA) for portable pulsed neutron generators. The experiments showed that the conditions for igniting a discharge depend on the pulse-periodic powering system of the ion source and the conditions for heating the gas generator of the neutron tubes. The obtained data show that the efficiency and stability of gas-filled neutron tubes can be increased by optimizing the operating regimes of the powering systems and creating the appropriate ignition conditions for a discharge.

Multiple Gamma-Ray Detection Capability of a CeBr3 Detector for Gamma Spectroscopy

The newly developed cerium tribromide (CeBr3) detector has reduced intrinsic gamma-ray activity with gamma energy restricted to 1400–2200 keV energy range. This narrower region of background gamma rays allows the CeBr3 detector to detect more than one gamma ray to analyze the gamma-ray spectrum. Use of multiple gamma-ray intensities in elemental analysis instead of a single one improves the accuracy of the estimated results. Multigamma-ray detection capability of a cylindrical 75 mm × 75 mm (diameter × height) CeBr3 detector has been tested by analyzing the chlorine concentration in water samples using eight chlorine prompt gamma rays over 517 to 8578 keV energies utilizing a D-D portable neutron generator-based PGNAA setup and measuring the corresponding minimum detection limit (MDC) of chlorine. The measured MDC of chlorine for gamma rays with 517–8578 keV energies varies from 0.07 ± 0.02 wt% to 0.80 ± 0.24. The best value of MDC was measured to be 0.07 ± 0.02 wt% for 788 keV gamma rays. The experimental results are in good agreement with Monte Carlo calculations. The study has shown excellent detection capabilities of the CeBr3 detector for eight prompt gamma rays over 517–8578 keV energy range without significant background interference.

Formation of the Neutron Energy Spectrum Near the Target of a Portable NG-24M Neutron Generator

The spectrum of neutrons near the target of a portable NG-24M neutron generator is measured by the neutron activation method. The recovered neutron spectrum is represented analytically as a superposition of known physically substantiated spectra.

Eastern Mediterranean Economic Exchange during the Iron Age: Portable X-Ray Fluorescence and Neutron Activation Analysis of Cypriot-Style Pottery in the Amuq Valley, Turkey.

Two markers of regional exchange in the Eastern Mediterranean during the first millennium BCE are the White Painted and Bichrome Wares from Cyprus’s Cypro-Geometric and Cypro-Archaic periods. Although these ceramics are often assumed to be imports from Cyprus, excavations in southern Turkey at sites such as Tarsus-Gözlükule, Kilise Tepe, Sirkeli Höyük, and Kinet Höyük suggest that at least some of this pottery was produced locally, requiring a major revision of our understanding of economic interaction in the Eastern Mediterranean. We employ a combination of portable x-ray fluorescence and neutron activation analysis to investigate the White Painted and Bichrome Wares recovered from Tell Tayinat, Çatal Höyük, and Tell Judaidah, three sites in the Amuq Valley of southeastern Anatolia. Our results demonstrate that a clear geochemical distinction exists between imported and local versions of this pottery. Through comparison with legacy datasets, we locate the likely origin of the imported pottery in the Circum-Troodos sediments of central and southern Cyprus. The secondary and tertiary settlements of Çatal Höyük and Tell Judaidah had access only to this imported material. In contrast, the inhabitants of Tell Tayinat, capital city of the region, consumed both imported and locally produced White Painted and Bichrome Wares. This pattern cannot be explained in purely economic terms whereby the frequency of imports decreases as distance from the point of production increases. Instead, we suggest that elite feasting practices drove demand, resulting in either local potters producing Cypriot-style pottery or Cypriot potters settling in the vicinity of Tell Tayinat. These findings offer new insights into the relationship between historically attested Iron Age kingdoms in southern Turkey and Cyprus and complicate our understanding of exchange in the Eastern Mediterranean during the Iron Age.

Angular distribution of 4.43-MeV γ-rays produced in inelastic scattering of 14.1-MeV neutrons by 12C nuclei

The work is devoted to measuring the angular distribution of 4.43-MeV γ-rays produced in inelastic scattering of 14.1-MeV neutrons by 12C nuclei. A portable ING-27 neutron generator (designed and fabricated at VNIIA, Moscow) with a built-in 64-pixel silicon α-detector was used as a source of tagged neutrons. The γ-rays of characteristic nuclear radiation from 12C were detected with a spectrometric system that consisted of 22 γ-detectors based on NaI(Tl) crystals arranged around the carbon target. The measured angular distribution of 4.43-MeV γ-rays is analyzed and compared with the results of other published experimental works.

Portable spectroscopic fast neutron probe and 3He detector dead-time measurements

This paper presents dead-time calculations for the Portable Spectroscopic Fast Neutron Probe (N-Probe) using a combination of the attenuation law, MCNP (Monte Carlo N-particle Code) simulations and the assumption of ideal paralyzing and non-paralyzing dead-time models. The N-Probe contains an NE-213 liquid scintillator detector and a spherical 3He detector. For the fast neutron probe, non-paralyzing dead-time values were higher than paralyzing dead-time values, as expected. Paralyzing dead-time was calculated to be 37.6 μs and non-paralyzing dead-time was calculated to be 43.7 μs for the N-Probe liquid scintillator detector. Dead-time value for Canberra 3He neutron detector (0.5NH1/1K) was also estimated using a combination of subcritical assembly measurements and MCNP simulations. The paralyzing dead-time was estimated to be 14.5 μs, and the non-paralyzing dead-time was estimated to be 16.4 μs for 3He gas filled detector. These results are consistent with the dead-time values reported for helium detectors.

Energy resolution measurements of CeBr3 and LaCl3:Ce detectors

The energy resolution of LaCl3:Ce and CeBr3 detectors were measured from gamma rays produced due to thermal neutron capture in CeBr3 and LaCl3:Ce detector’s material using a portable DD neutron generator-based prompt gamma ray neutron activation analysis (PGNAA) setup. The energy resolution of CeBr3 detector varies from 6.64 ± 0.04 to 2.75 ± 0.04 % over 368–2223 keV gamma rays while for LaCl3:Ce it varies from 7.09 ± 0.02 to 0.90 ± 0.01 % over 272–8579 keV gamma rays. The energy resolution data of both detectors exhibit 1/√E γ dependence.

SU‐F‐T‐657: In‐Room Neutron Dose From High Energy Photon Beams

Purpose:To estimate neutron dose inside the treatment room from photodisintegration events in high energy photon beams using Monte Carlo simulations and experimental measurements.Methods:The Monte Carlo code MCNP6 was used for the simulations. An Eberline ESP‐1 Smart Portable Neutron Detector was used to measure neutron dose. A water phantom was centered at isocenter on the treatment couch, and the detector was placed near the phantom. A Varian 2100EX linear accelerator delivered an 18MV open field photon beam to the phantom at 400MU/min, and a camera captured the detector readings. The experimental setup was modeled in the Monte Carlo simulation. The source was modeled for two extreme cases: a) hemispherical photon source emitting from the target and b) cone source with an angle of the primary collimator cone. The model includes the target, primary collimator, flattening filter, secondary collimators, water phantom, detector and concrete walls. Energy deposition tallies were measured for neutrons in the detector and for photons at the center of the phantom.Results:For an 18MV beam with an open 10cm by 10cm field and the gantry at 180°, the Monte Carlo simulations predict the neutron dose in the detector to be 0.11% of the photon dose in the water phantom for case a) and 0.01% for case b). The measured neutron dose is 0.04% of the photon dose. Considering the range of neutron dose predicted by Monte Carlo simulations, the calculated results are in good agreement with measurements.Conclusion:We calculated in‐room neutron dose by using Monte Carlo techniques, and the predicted neutron dose is confirmed by experimental measurements. If we remodel the source as an electron beam hitting the target for a more accurate representation of the bremsstrahlung fluence, it is feasible that the Monte Carlo simulations can be used to help in shielding designs.

New generation non-stationary portable neutron generators for biophysical applications of Neutron Activation Analysis

BackgroundNeutron sources are increasingly employed in a wide range of research fields. For some specific purposes an alternative to existing large-scale neutron scattering facilities, can be offered by the new generation of portable neutron devices. Scope of reviewThis review reports an overview for such recently available neutron generators mainly addressed to biophysics applications with specific reference to portable non-stationary neutron generators applied in Neutron Activation Analysis (NAA). Major conclusionsThe review reports a description of a typical portable neutron generator set-up addressed to biophysics applications. General significanceNew generation portable neutron devices, for some specific applications, can constitute an alternative to existing large-scale neutron scattering facilities. Deuterium-Deuterium pulsed neutron sources able to generate 2.5MeV neutrons, with a neutron yield of 1.0×106n/s, a pulse rate of 250Hz to 20kHz and a duty factor varying from 5% to 100%, when combined with solid-state photon detectors, show that this kind of compact devices allow rapid and user-friendly elemental analysis. "This article is part of a Special Issue entitled "Science for Life" Guest Editor: Dr. Austen Angell, Dr. Salvatore Magazù and Dr. Federica Migliardo".

Performance tests of a large volume cerium tribromide (CeBr3) scintillation detector

The response of a large cylindrical 76mm×76mm (height×diameter) cerium tribromide (CeBr3) detector was measured for prompt gamma rays. The total intrinsic activity of the CeBr3 detector, which was measured over 0.33–3.33MeV range, was found to be 0.022±0.001 counts/s/cm3. The partial intrinsic activity ( due to 227Ac contamination), was measured over a energy range of 1.22–2.20MeV energy, was found to be 0.007±0.001 counts/s/cm3. Compared to intrinsic activities of LaBr3:Ce and LaCl3:Ce detectors of equivalent volume, the CeBr3 detector has 7–8 times less total intrinsic activity.The detector response for low energy prompt gamma rays was measured over 0.3–0.6MeVgamma energy range using a portable neutron generator-based Prompt Gamma Neutron Activation Analysis (PGNAA) setup. The experimental yield of boron, cadmium and mercury prompt gamma-rays was measured from water samples contaminated with 0.75–2.5wt% mercury, 0.31–2.50wt% boron, and 0.0625–0.500wt% cadmium, respectively. An excellent agreement has been observed between the calculated and experimental yields of the gamma rays. Also minimum detection limit (MDC) of the CeBr3 detector was measured for boron, cadmium and mercury samples. The CeBr3 detector has 23% smaller value of MDCB and 18% larger value of MDCCd than those of a LaBr3:Ce detector of equivalent size.This study has shown that CeBr3 detector has an excellent response for the low energy prompt gamma-rays with almost an order of magnitude low intrinsic activity as compared to LaCl3:Ce and LaBr3:Ce detectors of equivalent volume.

Test facility for nuclear planetology instruments

An experimental facility for testing and calibrating nuclear planetology instruments has been constructed in partnership between the Space Research Institute (Moscow) and the Joint Institute for Nuclear Research. A model of Martian soil with a size of 3.82 × 3.21 m2 and an overall mass of about 30 t is assembled from silicate glass. Glass is chosen in order to imitate absolutely dry soil close in composition to the Martian one. The heterogeneous model allows one to imitate the average elemental composition of Martian soil in the best possible way by adding layers of the necessary materials to it. Near-surface water ice is simulated by polyethylene layers buried at different depths within the glass model. A portable neutron generator is used as the neutron source for testing active neutron and gamma spectrometers. The facility is radiation-hazardous and is thus equipped with interlock and radiation monitoring systems in accordance with the effective regulations.

The comparison between simple and advanced shielding materials for the shield of portable neutron sources

Background : Monte Carlo simulaons play a vital role in the calculaon of the necessary shielding both for neutrons and photons. Advanced and simple shielding materials against neutron and gamma rays were compared by simulaon using the MCNB4B Monte Carlo code. The simulaons were carried out for the three common neutron sources, namely the 252 Cf, the 241 Am/Be and the DD neutron generator which are suitable for transportable facilies. Materials and Methods : The source has been simulated as sphere with 3 cm diameter while the necessary shielding is designed in the form of a sphere around the neutron source. The materials considered were chosen according to the EU Direcve 2002/95/EC, hence excluding lead and cadmium. Results: In the case of DD neutron generator the thickness, the weight and the volume of the shield can decrease up to 41.3, 44, and 78.4% correspondingly. With regard to the 252 Cf neutron source the use of advanced shielding materials can reduce the corresponding parameters up to 32.7, 40.7, and 68.4% respecvely. As regards the 241 Am/Be neutron source, based on advanced shielding materials the thickness, the mass and the volume of the shield can decrease by 33.8, 49.5, and 70% respecvely. Conclusion: The obtained results showed that the use of advanced shielding materials has led to reduce greatly the weight and the volume of the necessary shield.

A portable and wide energy range semiconductor-based neutron spectrometer

Hand-held instruments that can be used to passively detect and identify sources of neutron radiation—either bare or obscured by neutron moderating and/or absorbing material(s)—in real time are of interest in a variety of nuclear non-proliferation and health physics applications. Such an instrument must provide a means to high intrinsic detection efficiency and energy-sensitive measurements of free neutron fields, for neutrons ranging from thermal energies to the top end of the evaporation spectrum. To address and overcome the challenges inherent to the aforementioned applications, four solid-state moderating-type neutron spectrometers of varying cost, weight, and complexity have been designed, fabricated, and tested. The motivation of this work is to introduce these novel human-portable instruments by discussing the fundamental theory of their operation, investigating and analyzing the principal considerations for optimal instrument design, and evaluating the capability of each of the four fabricated spectrometers to meet the application needs.