MeV Neutrons Research Articles

A multimodal neutron-based technique for the elemental analysis of materials in bulk

Neutrons and gamma rays are effective probes for the elemental characterisation of bulk samples. Neutrons interactions with matter are characteristic of the nature of the target nucleus, and the incident neutron energy. Most nuclides exhibit distinctive structure in their total and differential cross sections, thus elements can be differentiated via their total, elastic and inelastic scattering cross sections, and the energies of prompt gamma rays produced in inelastic collisions. In this work we report on the use of the fast neutron transmission analysis technique for the analysis of 12C in graphite. We also present results of prompt gamma ray neutron activation analysis of graphite using 14 MeV neutrons, which was explored for the purpose of developing a multimodal neutron-based technique for elemental analysis of materials in bulk.

Effect of ionizing radiation on thermal and mechanical properties of filled-epoxy adhesives

Stave assemblies at the Large Hadron Collider and other high-energy physics experiments require radiation-resistant adhesives composed of light elements with high thermal conductivity and low interfacial resistance. Porous carbon foams are currently bonded to carbon fiber faceplates with graphite-filled epoxy. Low-viscosity epoxies are used to increase the amount of filler in order to double the thermal conductivity without impacting the properties of the material under irradiation. Promising new adhesive formulations showed no degradation in thermal conductivity after exposure to a fluence of 1016 n/cm2 (>0.1 MeV neutrons) or 10 Gy gamma radiation, but showed variable degradation, compared to the currently used adhesive, after exposure to a fluence of 1015 p/cm2 (1 MGy, 67.5 MeV protons). Adhesives with more than double the irradiated thermal conductivity of the baseline material showed marked improvement in thermal transport in graphite/epoxy/foam/epoxy/graphite structures based on thermal diffusivity measurements. Silicon carbide particles can be added to graphite to stiffen the modulus of the filled epoxy. When spherical AlN was used as a filler, the degradation in thermal conductivity under proton irradiation varied based on the size of the ceramic filler. Options for improving the thermal performance of irradiated adhesives for high-energy physics experiments are discussed.

Determination of boron content in Tinkal ore samples with Geant4 using MCLLS method

A new Monte Carlo-Library Least Squares (MCLLS) method was developed for the neutron inelastic-scattering and thermal-capture analysis (NITA) techniques in real-time online analysis systems to determine boron content in Tinkal ore samples. The Geant4 toolkit was used to simulate the Prompt Gamma-ray Neutron Activation Analysis (PGNAA) system for the NITA technique. Tinkal ore samples were bombarded with 14 MeV neutron beams in order to represent neutrons released from the Deuterium-Tritium (DT) neutron generator. After the simulation of shooting Tinkal ore samples with 14 MeV neutron beams was completed, histograms were obtained from Geant4. Once the histograms are broadened according to the shape calibration of the BGO scintillation detector, the spectrum (broadened histograms) are ready to be used in the MCLLS process. A single element library was created by broadened histograms of the nine Tinkal ore samples. In addition, set #10 was selected to be an unknown sample and was used for validating MCLLS. After analysing the unknown sample spectrum, boron content was estimated by the single element library created by the MCLLS method. The estimated value of the boron content by MCLLS was consistent with the laboratory result.

Mechanical Behaviour of the Rotating Target SORGENTINA-RF

The SORGENTINA-RF project aims at developing a 14 MeV neutron source based on a deuterium/tritium ion accelerator and a rotating target where fusion reactions take place. Among the different research fields, the most interesting and promising is medical radioisotope production. In this framework, intense research work on the design of the rotating target has been carried out. More in detail, to define the optimal design configuration, a sensitivity analysis on the mechanical performance of the rotating target considering the influence of some relevant parameters, such as material type, thickness, and presence of internal stiffeners, was carried out. Among the materials analyzed, aluminium alloy represents the best compromise to efficiently address all the critical requirements in the design phase. One of the most demanding project requirements that the conceptual design must fulfill is the ability of the target to dissipate a thermal power of 250 kW without precluding its mechanical properties and resistance. To investigate the performance of the rotating target, some thermo-mechanical analyses were undertaken with finite element method under some thermal transients and mechanical loads able to simulate the working conditions of the system. The numerical results emphasise the target’s ability to withstand operating conditions. The main outcomes of the present study have been implemented as engineering solutions in the project design.

Single-Event Upsets for Single-Port and Two-Port SRAM Cells at the 5-nm FinFET Technology

Single-event cross sections of single-port and two-port SRAM designs are presented at the 5-nm node for alpha particles, 14-MeV neutrons, thermal neutrons, and heavy ions for reduced and nominal supply voltages. Single-port SRAM shows higher susceptibility with low LET particles at nominal operating voltages. However, circuit and parasitic differences in designs lead to the two-port design having higher susceptibility to high LET particles. The stronger dependence of two-port SRAM on particle LET is due to the increased contribution of diffusion current to overall charge collection. The increased number of fins connected to a storage node for the two-port design significantly increases diffusion charge collection and SEU cross-section, despite increased critical-charge and cell size.

Development of Directional 14 MeV-Fusion Neutron Detector Using Liquid-Scintillator-Filled Capillaries

Triton burnup studies have been performed in fusion experimental devices to evaluate the confinement performance of 3.5 MeV alpha particles. For this purpose, the generation rate of deuterium-tritium (D-T)-born 14 MeV neutrons has been measured using a scintillating fiber detector. The directionality of this detector provides excellent selectivity for 14 MeV neutrons; however, the lack of a pulse–shape discrimination (PSD) capability limits high-energy gamma-ray reduction. In this study, we developed a new 14 MeV neutron detector with directionality that can discriminate neutrons and gamma rays based on the PSD technique by filling capillaries with a liquid scintillator. The performance of this detector was evaluated at the FNL (Tohoku University, Japan) and OKTAVIAN (Osaka University, Japan). The detector response was modeled using the particle and heavy ion transport code system (PHITS). The experimental and simulation results demonstrated that the detector has a directional response to fast neutrons and excellent PSD capability.

A study on (n, α) reaction cross sections using a new empirical systematic

In this article, we report a new empirical formula for quick calculation of cross sections of (n, α) reactions with 14–15 MeV neutrons. Cross sections are analysed in terms of the compound nucleus model. A systematic trend for 14–15 MeV neutrons is found in the variation of (n, α) reaction cross sections with the parameters (N − Z + 1), (En+Q)0.5 and N/Z. The empirical relation between the cross sections and these parameters has been obtained, which give fairly good fits with the experimental data. We have also investigated the odd-even effects on (n, α) cross sections considering binding energy systematic of the shell model. The present formula is very useful in predicting of the (n, α) cross sections, where the measurements are not available as well as in testing new experimental data.

Measurement of the [formula omitted] and [formula omitted] reaction cross-sections and isomeric ratios induced by the 14.8 MeV neutrons

The precise measurement of the (n, 2n) reaction cross-sections of xenon isotopes is of great significance for the diagnosis of the Inertial Confinement Fusion fuel area density ρR. The preparation of xenon samples and how they can be efficiently adsorbed to obtain an optimal reaction cross section are extremely difficult in existing experiments. This work aims to use the 93Nb(n,2n)92mNb reaction cross section as the standard to calculate the cross-sections and isomeric ratio for 134Xe(n,2n)133m,gXe and 136Xe(n,2n)135m,gXe reactions. This work also aims to explore the feasibility of adsorbent selection, the method of adsorption sample and the feasibility of the method used in the laboratory. In this paper, the Geant4 program was used to simulate and calculate the adsorption factors of activated carbon and polymer for standard 152Eu multi-line sources, so as to determine the most suitable adsorption material. Next, the GammaVision analysis software was used to observe whether the characteristic γ-rays released by other isotope reaction products of xenon in the sample and packaging material had an effect on the characteristic γ-rays of the target product. Then, the basic principle of activation method and the relative measurement principle are used to calculate the cross-sections and isomeric ratio for 134Xe(n,2n)133Xe and 136Xe(n,2n)135Xe reactions. Finally, the experimental results are compared and analyzed with the existing data in EXFOR database to verify whether the experimental method is correct.

Measurement of Cumulative fission product yields on 235U induced by 2.8 MeV neutrons

Off-line gamma-ray spectrometry was used to accurately measure the Cumulative fission product yields (CFPYs) of fission products in the 235U (n, f) reaction induced by 2.8 MeV neutrons. The 2.8 MeV quasi-monoenergetic neutron beam was produced by the CPNG-600 Cockcroft Walton accelerator at the China Institute of Atomic Energy （CIAE）and the gamma spectra were measured by the HPGe γ-ray Spectrometer. After fully considering and revising the sources of uncertainty, high-precision CFPYs of 4 fission products were obtained. This study has important applications in reactor design and operation and is conducive to the establishment of an evaluated nuclear database.

Development of a compact fast-neutron spectrometer for nuclear emergency response applications

We have developed a Compact Fast Neutron Spectrometer (CFNS) for passive assay of special nuclear material (SNM) through the observation of fast neutrons. The CFNS consists of eight organic glass scintillators (OGS) coupled to silicon photomultipliers and a waveform digitizer, which are integrated within a human-portable box. The CFNS determines the neutron energy profile by spectrum unfolding using the Maximum-Likelihood Expectation Maximization method. The detector acquisition system was optimized to have a dynamic range of up to 10 MeV neutron energy. Bulk special nuclear material (SNM) measurements from the National Criticality Experiments Research Center were analyzed for SNM validation/examination. The results show that the CFNS can be used to distinguish between fission and (α, n) neutron emitters, regardless of intervening material type (Cu and polyethylene) and thickness, by taking the ratio of neutron counts at different regions in the unfolded energy spectrum. Additionally, by fitting an exponential curve to the unfolded energy spectrum of PuO2 and Pu neutron emitters, the CFNS showed the ability of distinguishing between pure Pu oxide, pure Pu metal and mixed oxide-metal configurations.

76Se(n,2n)75Se reaction cross section measurement at 14.77 MeV neutron energy

The cross section of 76Se(n,2n)75Se reaction at 14.77 ± 0.17 MeV neutron energy was measured with offline gamma spectrometry. Furthermore, covariance analysis for the measured cross section with respect to the other neutron induced reaction cross sections for the same target was reported. The experimental results were compared with reported experimental data from the EXFOR database and evaluated data from the ENDF/B-VIII.0, JENDL-5, and TENDL-2021 libraries. Statistical model calculations with the TALYS-1.96, and EMPIRE-3.2.3 codes were carried out for optimized parameters corresponding to experimental trends. The measured cross section is in good agreement with the data from the EXFOR database, and the evaluated data libraries.

Carbon Detection Using an HPGe Detector

Fast neutron activation analysis has been applied to bulk carbon detection. DT-neutron generator emitting 14.1 MeV neutrons was used as a neutron source. Prompt 4.44 MeV gammas from the inelastic neutron scattering on carbon were detected using a 30% HPGe semiconductor gamma spectrometer shielded from the primary neutron beam. The sensitivity of the experimental setup is discussed. Keywords: Activation analysis; Fast neutrons; Gamma spectrometry.

Impacts of 14 MeV neutron irradiation on electrical and spectral properties of flexible GaInP/GaAs/InGaAs solar cells

Impacts of 14 MeV neutron irradiation on electrical and spectral properties of flexible GaInP/GaAs/InGaAs solar cells

Simultaneous detection of fast and thermal neutrons with a stilbene-6Li glass composite scintillator

For simultaneous detection of fast and thermal neutrons, a composite scintillator consisting of stilbene crystal and 6Li glass was designed. Energy calibration was finished with a 152Eu γ source through Compton coincidence measurement. A 241Am-Be neutron source with polyethylene moderator was used to test it. The detection efficiency was examined through Monte Carlo simulation. The measured results show that the composite scintillator has excellent triple pulse shape discrimination performance. With the threshold of 100 keVee, figure of merits for γ-fast neutron and fast neutron-thermal neutron are 1.571 ± 0.009 and 2.376 ± 0.010 respectively. Two thermal neutron peaks emerge at 0.245 MeVee and 0.291 MeVee in the energy spectra. 79.24% of the incident thermal neutrons can be detected by 6Li glass. 39.56% of 1 MeV neutrons can be detected by stilbene.

Fission cross-section measurement for [formula omitted] and [formula omitted] in [formula omitted] reaction induced by 14MeV D-T neutrons

In this work, the independent fission cross-sections of U(n,f)238Xe135g and U(n,f)238Xe135m reactions induced by neutrons with the energies of 14.1 MeV, 14.5 MeV, and 14.7 MeV were measured by activation methods. The neutrons from T(d,n)He4 were used in the experiments and their energies were determined by the ratio of Zr(n,2n)90Zr89 and Nb(n,2n)93Nb92m reaction cross-sections. Aluminum films were chosen as reference samples to measure the neutron fluence relative to the cross section of the Al(n,α)27Na24 reaction. The effect of self-absorption, geometry, and cascade coincidence were also considered during the data analysis. Besides, the increase in the daughter nuclide yield due to the decay of the parent nuclides in the same decay chain was deducted. As a result, the independent fission cross-sections of the U(n,f)238Xe135g reaction are 2.54 ± 0.14 mb, 3.05 ± 0.19 mb and 2.94 ± 0.19 mb, while the cross-sections of U(n,f)238Xe135m reaction are 2.11 ± 0.16 mb, 2.47 ± 0.18 mb and 2.34 ± 0.21 mb for 14.1 MeV, 14.5 MeV and 14.7 MeV neutrons, respectively. This work provides experimental data for the database of nuclear fission reactions.

Measurement of dosimetrical cross sections with 14.05 MeV neutrons from compact neutron generator

This paper deals with the measurement of the differential dosimetry cross sections using a small compact D-T neutron generator with 14.05 ± 0.08 MeV neutron emission (1 × 108 n/s into 4π). In spite of low neutron flux, it is possible to get measurable gamma-ray activities by irradiating a larger amount of target material placed in close geometry. The experimentally determined cross sections are in good agreement with the cross sections in the IRDFF-II dosimetry library. The comparison with other nuclear data libraries was performed as well. It is worth noting that the mean standard deviation in IRDFF-II library is about 4 %, while in the case of other data libraries they are from 5.5 % to 7.5 %. This result can be understood as a validation of IRDFF-II using 14.05 MeV neutrons and also a confirmation of the applicability of small compact generators in the measurement of activation cross sections.

TDCpix pixel detector ASIC with 100 ps time stamping

The TDCpix pixel read-out ASIC contains 1800 pixels arranged in 40 columns and 45 rows with the dimension of 300μm x 300μm. Each pixel contains a preamplifier and shaper circuit with a dynamic range of 0.8 to 10 fC and a rise time of 5 ns, followed by a Leading-Edge discriminator with Time-over-Threshold correction. The discriminator outputs of each pixel are connected to time-to-digital converters (TDC) measuring the time when the input signal has exceeded the threshold and the pulse width with a binning of 97 ps. The electronics noise of pre-amplifier/shaper is 170 e− or 2.7 mV rms with a gain of 65 mV/fC. The jitter of the entire processing chain for an electrical input signal of 2.4 fC is lower than 60 ps rms. The ASIC has been designed to work in radiated environments of 6 * 104 Gray per year and 2 x 1014 1 MeV neutron equivalent cm−2.

Semiconductor Detector Study for Detecting Fusion Neutrons using Geant4 Simulations

Accurate neutron flux measurements in fusion reactors are essential, in order to determine the feasibility and progress of the reaction as well as for safety issues. Semiconductor neutron detectors exhibit promising characteristics for operation in the extreme environmental conditions of fusion reactors. Silicon, Diamond and Silicon Carbide are the most studied and anticipated materials for constructing detectors with high efficiency and irradiation resistance. The ITER fusion reactor is expected to run D-D plasma measurements in the near future, so the detection of 2.45 MeV neutrons with appropriate detectors is of great and immediate importance. In the present work the study of 2.45 MeV neutrons interactions with a silicon, diamond and silicon carbide detector was made, using GEANT4 [1] simulations, in order to compare their response. An experimental study will follow at the neutron production facility of the TANDEM accelerator of the I.N.P.P. of the NCSR “Demokritos”, with detectors provided by CIVIDEC Instrumentation GmbH, so the geometry of the simulations was built accordingly. A quasi-monoenergetic neutron beam of 2.45 MeV was produced through 3H(p,n) reactions in a TiT target. Due to the low cross section of the reaction, biasing techniques were implemented in the simulation to increase the counting rate and thus producing realistic results. These biasing techniques were studied, with various tests and the parameters affecting the choice of the biasing factor are shown and discussed.

Radiation Biology of Structurally Different <i>Drosophila melanogaster</i> Genes. Message 9. General Regularities and Locus-Specific Features of Radiomutability of Sex-Linked and Autosomal Genes

The results of large-scale radiation, genetic, cytogenetic and molecular genetic studies on the nature and frequency of heritable mutations of two sex-linked and three autosomal genes of Drosophila melanogaster after exposure to 60Co γ-rays and 0.85 MeV neutrons are presented. Regularities were established for studied genes and radiations in the induction of five different types of recessively inherited mutations, which can be combined into two main classes - the class of chromosomal mutations in one way or another affecting the gene under study (changes leading to sterility of F1 mutants, multilocus deletions, inversion or translocation breakpoint chromosome in the region of gene localization) and a class of “point” gene mutations with a complex spectrum of DNA changes detected by PCR and sequencing. In this report, the results of classical genetic and cytogenetic analysis of these classes of mutations are considered in detail and the frequency of their induction by γ-rays or neutrons is estimated per locus per dose unit (1 Gy). An important and unexpected result of the assessment was the fact that the frequency of “point” mutations turned out to be invariant (on average 1.15 E-06/locus/Gy) for the studied genes and radiations, in contrast to chromosomal mutations, where a pronounced locus-specificity is observed for individual types of mutations. At the same time, neutrons are two or more times more effective than γ-rays in inducing this class of mutations. It is significant that the ave-rage frequency of “point” mutations induction correlates with the average cDNA length of these genes (1.62 kb), but not with their average size (6.07 kb), indicating that the target for " point mutations is, obviously, not the entire DNA of the gene, but only the informational part of its exons. The dependence of the frequency of chromosomal mutations of one type or another on the position of the gene on the chromosome and in the three-dimensional space of the genome is discussed.

Towards the use of SDD as an absolute detector for high-energy neutron measurements

As of today, the standard method employed in tokamaks for the absolute measurement of the neutron flux (thus of the nuclear fusion power) is based on activation foils, being the most robust and unbiased technique for the absolute determination of neutron fluence. However, this technique is not able to provide real-time data useful for the control of future fusion plants like DEMO. In this paper, we present some preliminary results about the R&D activity aimed at developing the Single-crystal Diamond Detectors (SDD) used for fast neutron measurements into an absolute neutron flux monitor. Tests have been conducted at the new NILE neutron source of the Rutherford-Appleton Laboratory, a facility with compact neutron generators with a maximum yield of 109 n/s and 1010 n/s for 2.5 MeV and 14 MeV neutrons, respectively. A series of neutron spectra and flux measurements have been taken with different SDD and associated DAQ. Comparisons with standard activation foils (and namely Fe, Zr, Al and Nb foils for 14 MeV neutrons and In for 2.5 MeV neutrons) and with other reference detectors are presented and discussed. Also discussed is the stability of the SDD over time when employed at high neutron rates in realistic neutron environment, and the effects of neutron irradiation on both the counting rate and detector resolution.