MeV Neutrons Research Articles

Double differential cross section calculations of (n,p) reactions on 27Al and 24Mg targets

In structural fusion material research, double differential sections are necessary to determine heating and damage due to secondary particles that may occur in structural materials. Therefore, in this study, the double differential proton emission cross sections of 27Al and 24Mg target nuclei were calculated theoretically with the TALYS nuclear reaction code at 14 MeV neutron energy. The calculated values were compared with existing experimental data in the EXFOR library. Additionally, the contribution of direct, compound and preequilibrium reactions in theoretical calculations was investigated.

Reconstruction of neutron spectrum emitted by portable 14 MeV neutron generator by a combination of different methods

The presented paper is a continuation of the Tikhonov Regularisation, Minimum Fisher Information, Maximum Entropy, and Maximum Likelihood methods application to unfolding the neutron spectra emitted by the fusion devices from activation measurement. The aim of the analysis is the evaluation of the possible uncertainty improvement. The developed algorithms and methodology were validated with 14 MeV portable neutron generators before application to the analysis of the neutron spectrum emitted from tokamak. The analysis concerns plasma-target generator did not provide satisfactory uncertainty of the results. The modification of the irradiation foils’ set and the improvement in the activity measurement precision progress the results. The sealed tube portable neutron generator irradiated the Al, Fe, Ni, Zr, Au, Mg, and Nb foils. The HPGe spectrometer measured the induced activity with an uncertainty of about 10 %. The synthetic data analysis was made to establish the ratio between the error of the spectrum reconstruction and the discrepancy between the input data and activities reconstructed for this spectrum by the FISPACT-II inventory code. Two bin structures were considered. One was compatible with the FISPACT-II code, but the energy resolution was not regular. The second was characterised by constant bin spacing connected with the modification requirement before simulation and additional uncertainty of about 5 %. The best results were obtained by calculation of the mean value from all methods. The corresponding reconstruction uncertainty is equal to 11 %. The reconstructed spectrum has a dominant deuterium tritium peak slightly shifted to the higher energies due to the beam-target reaction kinematics. The evaluation of the impact of the deuterium-deuterium reaction due to the long-term operation was not reachable due to the measured high intensity from scattered neutrons.

Monte Carlo Simulation and N-XCOM Software Calculation of the Neutron Shielding Parameters for the NCRP Report 144 Recommended Conventional Concretes

Purpose: Radiation shielding requires deep knowledge about the shielding materials properties. Additionally, the interaction between the radiation and materials should be well understood.
 Materials and Methods: Monte Carlo (MC) simulation, NXCOM, and WinXCOM computational programs were utilized for the concrete shielding properties against 1.5 MeV neutron beam and 137Cs emitted γ-ray. In a simulated “good geometry” using MCNP5 MC code, NXCOM and WinXCOM, radiation attenuation factor (µ), microscopic neutron removal cross-section ( Half and Tenth Value Layers (HVL and TVL) of the studied concretes were derived. Obtained results by the methods were compared and discussed.
 Results: For 137Cs emitted γ-ray, mass attenuation factor (µ/ρ) obtained as 0.026 cm2/g, 0.025 cm2/g, and 0.025 cm2/g for the Serpentine concrete as the minimum factors by MCNP5 code, WinXCOM and NXCOM software, respectively. Good agreement was seen in the results derived by the use of the applied calculation methods. Maximum values for the µ/ρ were calculated as 0.03 cm2/g, 0.029 cm2/g and 0.029 cm2/g by MCNP5 code, WinXCOM and NXCOM, respectively. For the neutron attenuation factor, calculations were conducted for the concretes and the highest and lowest ΣR/ρ were derived for Serpentine and Ordinary concretes. MCNP5 MC code was calculated ΣR/ρ for the Serpentine and Ordinary concretes as 0.039 cm2/g and 0.030 cm2/g, respectively. ΣR/ρ for the Serpentine and Ordinary concretes as 0.039 cm2/g and 0.030 cm2/g, respectively.
 Conclusion: It was concluded that the calculated results showed N-XCOM program can be applied for the shielding calculations for the conventional concretes studied in this work.

Study of trends in cross-section values of primary reactions of most abundant isotope of elements with Z= 6 to 116 induced by 14 MeV neutrons using TALYS code and the EXFOR database

Study was carried out to understand the trends of variation of cross-section values of all possible primary reactions induced by 14 MeV neutrons incident on the most abundant isotope of elements from atomic number 6 to 116. The available nuclear reaction channels and cross-section values of all possible primary reactions were obtained using the theoretical values calculated using TALYS nuclear code version 1.96. The trends were explained on the basis of nuclear binding energy curve, the pairing effect and shell effect. Most of the trends show a regular variation except at some points, which give interesting information for the nucleus where trend is irregular. The variation of the available experimental cross-section values obtained from the EXFOR database, if any, was also obtained and compared with the trends obtained from the theoretical values. The discrepancies in the trends of theoretical and experimental cross-section values of various possible reactions were explained on the basis of the possible errors in experiments and the assumptions of the theoretical model of the TALYS nuclear code.

Effects of correlations in uncertainties of total cross section and elastic angular distribution for a deep penetration of 14-MeV neutrons in Cu

ABSTRACT Uncertainty in neutron reaction rates after penetrating a thick copper benchmark assembly was estimated based on two different kinds of Total Monte Carlo methods under random sampling methodology. Five hundred random nuclear data files were generated for 63Cu by Bayesian Monte Carlo method by perturbing the underlying nuclear model parameters. In the first method, these files were used directly to generate processed libraries, preserving all the correlations among different physical quantities. In the second method, the random files generated in the first method were used, but the angular distributions of elastic scattering were kept fixed to those of the non-perturbed nominal one. It was found that the two methods have the same neutron reaction rate after 608 mm penetration of copper. However, the uncertainty of the first method was smaller than that of the second method. It shows that the correlation between angular distribution of elastic scattering at 0 degrees and total cross section, which stems from Wick’s inequality, affects uncertainty of the calculated neutron reaction rate. It could be concluded that the uncertainty obtained by using the covariance files given in the ENDF-6 format may not give correct results for the uncertainty of neutron penetration calculations.

A CFD analysis of the rotating target holder of the 14-MeV neutron generator

An accelerator based 14-MeV neutron generator is installed at Institute for Plasma Research, Gandhinagar to generate high energy neutrons by using a water-cooled rotating tritium target. In this device, neutrons would be generated from the nuclear reaction 3H(D, n)4He by bombarding accelerated deuterium ions on a solid tritium target. The rotating target holder mainly consists of a rotating copper disc water-cooled case mounted on the rotating shaft. The high-energy deuterium ion beam impacted on solid TiT target fitted on the copper case imparting its energy to the case. This energy is taken away from the system through the water passing through the annulus cavity of the shaft. This paper presents details of the engineering CFD (computed fluid dynamics) analysis carried out for optimizing two important parameters of the rotating target. Firstly optimization of the location of flow parameters of the water as mass flow rate and Secondly optimization of speed of the rotating target and the design of fins inside the copper case.

Measurement of yields and angular distributions of γ-quanta from the interaction of 14.1 MeV neutrons with oxygen, phosphorus, and sulfur* *Supported by RSCF (23-12-00239)

A study of the inelastic scattering of neutrons with an energy of MeV on the nuclei of oxygen, phosphorus, and sulfur was performed at the TANGRA facility at JINR (Dubna). The experiment aimed to refine existing data and obtain new data on the yields and angular distributions of γ-quanta emitted by the studied nuclei due to neutron-induced nuclear reactions using the tagged neutron method. Two types of detector systems were used to register γ-quanta. The γ-ray yields were measured using a high-purity germanium (HPGe) detector. The angular distributions of γ-rays were obtained using a system of 18 scintillation detectors based on bismuth germanite Bi Ge O (BGO) located around the sample. The performed experiments measured the yields of two transitions for the reaction of tagged neutrons with O, nine transitions for the reaction with P, and nine transitions for the reaction with S for the first time. The angular anisotropy of the γ-radiation accompanying the inelastic scattering of neutrons with an energy of MeV on P nuclei was also measured for the first time.

Challenge to Performance Evaluation of Fusion Reactor Materials by Fission Neutron and Charged Particle Irradiation

Since there are no fusion reactors generating high-flux 14 MeV neutrons, it is necessary to evaluate materials’ performance in fusion reactors based on a correlation of fission neutron and charged particle irradiations. However, the irradiation tests involve various issues which prevent simple correlation and evaluation. In this paper, the issues related to irradiation temperature control and dose rate effects are pointed out and analyzed, and proposals regarding future irradiation tests are given.

Radioactivity study of EAST structural materials based on neutron activation system

The plasma operation in the Experimental Advanced Superconducting Tokamak (EAST) would generate 2.45 MeV neutrons and a small fraction of 14 MeV neutrons. Neutrons react with structural materials, leading to the production of radioactive isotopes. The identification of nuclides and the determination of radioactive activity are of great significance for radiation protection and fusion material applications. In this study, structural materials including stainless steel 316L(SS316L), stainless steel 304L(SS304L), aluminum(Al), tungsten(W) and copper(Cu) were transported to the activation terminal for fusion neutron irradiation. The gamma ray spectra were measured using high-purity germanium (HPGe) detector, and the main radioactive nuclides of SS316L and SS304L are 56Mn, while the main radioactive nuclide of Al is 28Al. For Cu and W, the main radionuclides resulting from neutron reactions are 66Cu and 187W, respectively. The uncertainty of the radioactivity is 2.27∼7.94 % for W, 1.95∼5.84 % for SS316L, 2.10∼6.07 % for SS304L, 0.68∼1.69 % for Al and 2.59∼6.75 % for Cu. The specific activity of radionuclides is proportional to neutron yield, so the material activation is the main source of radiation dose during irradiation with high-yield neutron. The neutron activation coefficients of SS316L, SS304L, Al, Cu and W are 2.74538 × 10−14, 3.6411 × 10−14, 2.50711 × 10−12, 1.76506 × 10−12 and 3.51725 × 10−14, respectively. The coefficient represents the radioactive activity generated by each fusion neutron. In addition, the predicted activity was calculated using FISPACT program and compared with experiments.

Measurement and simulation of the leakage neutron spectra from Fe spheres bombarded with 14 MeV neutrons

Measurement and simulation of the leakage neutron spectra from Fe spheres bombarded with 14 MeV neutrons

Reexamination of a 40Ca(n,α)37Ar cross section measurement from existing literature

Background 37Ar, produced from the neutron activation of 40Ca in sub-surface rock and soil, is important for nuclear explosion monitoring. Most of the cross section measurements done for the 40Ca(n,α) 37 Ar reaction are in the fast neutron region. A thermal cross section measurement was attempted in 1989, with the assumption that the 4-8 MeV neutrons would be completely thermalized with 8-cm of water. Methods The experimental setup was modeled using Monte Carlo N-Particle code to see if the assumptions for the neutron distribution were accurate. A model of a 3.0 MeV deuteron source was used to recreate the neutron source from the experiment. Additional modeling was done using SCALE to compare the predictions of 37Ar yields when using the measured cross section from 1989 compared to current nuclear data models. Results Modeling showed the neutron distribution was closer to 10% thermal and 90% fast as opposed to entirely thermalized. The additional modeling showed how the overestimation of the thermal cross section would affect the predicted yield of 37Ar. Conclusions Based on the modeling results, the assumption that all the 4-8 MeV neutrons are thermalized with 8-cm of water is not accurate. The high fraction of fast neutrons interacting with the Ca will result in an overestimation of the thermal cross section.

The single crystal diamond-based diagnostic suite of the JET tokamak for 14 MeV neutron counting and spectroscopy measurements in DT plasmas

The Joint European Torus (JET) has recently conducted its second deuterium–tritium (DT) experimental campaign DTE2, providing unique opportunity for studying both physics and engineering aspects of nuclear fusion plasmas. This also allowed the exploitation of new diagnostics and technologies that were not available during the first JET DT campaign held in 1997. Among these new instruments, the enhancement projects of the JET nuclear diagnostics lead to the development and installation of synthetic single crystal diamond detectors along different collimated line of sights. This paper describes the single crystal diamond-based diagnostic suite of the JET tokamak and the enhanced 14 MeV neutron diagnostic capabilities in terms of neutron yield and high resolution neutron spectroscopy. The diamond characterization measurements and the calibration procedure at JET are shown, together with performance of the diamond based neutron spectrometer as 14 MeV neutron yield monitor which allows the separation of 2.5 MeV and 14 MeV neutrons in trace tritium plasmas. The first high-resolution 14 MeV neutron spectroscopy measurements in neutral beam injection-heated DT plasmas are presented, allowing thermal and non-thermal neutron component separation. Prospects for the diagnose of DT burning plasmas such as ITER and SPARC will be presented.

Partially depleted operation of 250 μm-thick silicon carbide neutron detectors

Fusion experiments put many challenges on the neutron detection methods, since they require fast neutron counters and spectrometers with good energy resolution capable of operating in harsh environments under a large range of neutron flux intensities. Recently, SiC detectors were proposed for these roles due to their outstanding properties. Among those, the operation of SiC with partial polarization is of interest since it might allow for a fast neutron detector with efficiency tunable online, which would allow a good functionality under a large range of flux intensities.This paper describes the characterization of a series of SiC detectors operated with a fraction of the polarization needed to achieve full depletion. The characterization is performed through I/V measurements, alpha irradiation in vacuum and neutron irradiation. A functionality with excellent detection characteristics is demonstrated, reaching an energy resolution of 1.05 % for 5.5 MeV alphas and 2.05 % for 14 MeV neutrons. Some limitations on the maximum bias will be also discussed, pointing out some challenges for the manufacturing of thick SiCs in the future.

Design and analysis of novel high-performance thick epitaxial SiC radiation detector with low operating voltage by multiple p-type buried layers

In this work, epitaxial silicon carbide (SiC) radiation detectors (E-SRD) were modeled, and SiC PiN diodes with multiple p-type buried layers (PBL), named MPBL PiN, were proposed as E-SRD to address the problem that detection performance of E-SRD using convention PiN diodes (Cov. PiN) with thick epilayer is limited by its high fully depleted voltage (VFD). The charge collection efficiency (CCE), pulse height and relevant physical mechanism affecting detection performance of E-SRD were investigated. When both Cov. PiN and MPBL PiN have a 100 μm epilayer with n-type doping concentration of 3 × 1014 cm−3, the VFD of Cov. PiN reaches 2800 V, while that for MPBL PiN is only 550 V. 14 MeV neutron detection simulation results showed that, compared to Cov. PiN, the CCE, pulse height and radiation resistance of MPBL PiN improve significantly under low operating voltage within 1000 V, meanwhile its lower near-surface electric field is conducive to reducing leakage. Further, since the VFD of MPBL PiN is proportional to the thickness of the epilayer (HE) instead of to the square of the HE like Cov. PiN, the VFD of MPBL PiN with ultra-thick epilayer is much lower than that of Cov. PiN. Therefore, under commonly used operating voltage within 1000V, the sensitivity and pulse height of MPBL PiN can be proportional to the HE, while that of Cov. PiN hardly improved when HE is above 100 μm. So MPBL PiN can give full play to the high-performance detection advantages of ultra-thick epilayer, which is expected to promote the development and application of ultra-thick epilayer E-SRD with low operating voltage.

Measurement of neutron induced reaction cross-section of 99Mo

In the present work, we have measured 98Mo(n,γ)99Mo reaction cross-section using a 7Li(p,n)7Be neutron source at 1.67 ± 0.14, 2.06 ± 0.14 and 2.66 ± 0.16 MeV neutron energies. We have employed offline γ-ray spectroscopy to measure the induced activity of the sample. The 115In(n,n’γ)115mIn reaction was used as a monitor reaction. Different attributes propagating the uncertainty in the total result, measured cross-sections with their uncertainties and correlation coefficients are given in detail in the present study. The result is compared with the data libraries, EXFOR database and theoretical model outcome from different level density models.

A Portable Three-Layer Compton Camera for Wide-Energy-Range Gamma-ray Imaging: Design, Simulation and Preliminary Testing

(1) Background: The imaging energy range of a typical Compton camera is limited due to the fact that scattered gamma photons are seldom fully absorbed when the incident energies are above 3 MeV. Further improving the upper energy limit of gamma-ray imaging has important application significance in the active interrogation of special nuclear materials and chemical warfare agents, as well as range verification of proton therapy. (2) Methods: To realize gamma-ray imaging in a wide energy range of 0.3~7 MeV, a principle prototype, named a portable three-layer Compton camera, is developed using the scintillation detector that consists of an silicon photomultiplier array coupled with a Gd3Al2Ga3O12:Ce pixelated scintillator array. Implemented in a list-mode maximum likelihood expectation maximization algorithm, a far-field energy-domain imaging method based on the two interaction events is applied to estimate the initial energy and spatial distribution of gamma-ray sources. The simulation model of the detectors is established based on the Monte Carlo simulation toolkit Geant4. The reconstructed images of a 133Ba, a 137Cs and a 60Co point-like sources have been successfully obtained with our prototype in laboratory tests and compared with simulation studies. (3) Results: The proportion of effective imaging events accounts for about 2%, which allows our prototype to realize the reconstruction of the distribution of a 0.05 μSv/h 137Cs source in 10 s. The angular resolution for resolving two 137Cs point-like sources is 15°. Additional simulated imaging of the 6.13 MeV gamma-rays from 14.1 MeV neutron scattering with water preliminarily demonstrates the imaging capability for high incident energy. (4) Conclusions: We conclude that the prototype has a good imaging performance in a wide energy range (0.3~7 MeV), which shows potential in several MeV gamma-ray imaging applications.

Influence of 0.1–10 MeV neutron-induced SEUs on estimation of terrestrial SER in a nano-scale SRAM

The impact of 0.1–10 MeV neutron-induced single-event upsets (NSEUs) on terrestrial SER (soft error rate) prediction is analyzed for several broad spectrum neutron sources based on 0.1–200 MeV NSEU energetic-dependent cross-sectional data of a 40 nm SRAM. The results show that for test facilities with abundant <10 MeV neutrons, such as ISIS ChipIR, J-PARC BL10, CSNS Back-n, and CSNS BL09, the terrestrial SER is overestimated. The overestimation could reach 78%–287%. By reducing the cutoff energy (Emin) from 10 to 6 MeV, the overestimation can be decreased. However, the overestimation can still reach up to 164% in this case, significantly larger than the results reported by previous studies. By analyzing the contribution of neutrons in different energy bins to NSEUs, the discrepancy is proved to be attributed to the sensitivity of NSEUs to 0.1–1 MeV neutrons, along with the overpopulation of 0.1–1 MeV neutrons for certain test facilities, such as ISIS ChipIR and CSNS BL09. Therefore, it is necessary to take 0.1–1 MeV NSEU sensitivity into account when studying the prediction method using broad spectrum neutron sources.

Development of a D–D neutron spectrum detector based on helium-3 proportional counter

Deuterium–deuterium (D–D) neutron spectrum diagnostics in tokamaks are a challenging task with current technologies. To address this issue, we designed and tested a fast and compact helium-3 proportional counter with a diameter of 2.5 cm and an effective length of 15 cm and using Kr as a stopping gas. The detector achieved a resolution of 96 keV for 2.406 MeV neutrons with a pulse shaping of 2 µs. Test results indicate that this detector has the potential to form a D–D neutron spectrometer for tokamaks, composed of detector arrays.

A coded aperture with sub-mean free-path thickness for neutron implosion geometry imaging on inertial confinement fusion and inertial fusion energy experiments.

Inertial confinement fusion and inertial fusion energy experiments diagnose the geometry of the fusion region through imaging of the neutrons released through fusion reactions. Pinhole arrays typically used for such imaging require thick substrates to obtain high contrast along with a small pinhole diameter to obtain high resolution capability, resulting in pinholes that have large aspect ratios. This leads to expensive pinhole arrays that have small solid angles and are difficult to align. Here, we propose a coded aperture with scatter and partial attenuation (CASPA) for fusion neutron imaging that relaxes the thick substrate requirement for good image contrast. These coded apertures are expected to scale to larger solid angles and are easier to align without sacrificing imaging resolution or throughput. We use Monte Carlo simulations (Geant4) to explore a coded aperture design to measure neutron implosion asymmetries on fusion experiments at the National Ignition Facility (NIF) and discuss the viability of this technique, matching the current nominal resolution of 10 µm. The results show that a 10mm thick tungsten CASPA can image NIF implosions with neutron yields above 1014 with quality comparable to unprocessed data from a current NIF neutron imaging aperture. This CASPA substrate is 20 times thinner than the current aperture arrays for fusion neutron imaging and less than one mean free-path of 14.1MeV neutrons through the substrate. Since the resolution, solid angle, and throughput are decoupled in coded aperture imaging, the resolution and solid angle achievable with future designs will be limited primarily by manufacturing capability.

Evaluation of planar silicon pixel sensors with the RD53A readout chip for the Phase-2 Upgrade of the CMS Inner Tracker

The Large Hadron Collider at CERN will undergo an upgrade in order to increase its luminosity to 7.5 × 1034 cm-2s-1. The increased luminosity during this High-Luminosity running phase, starting around 2029, means a higher rate of proton-proton interactions, hence a larger ionizing dose and particle fluence for the detectors. The current tracking system of the CMS experiment will be fully replaced in order to cope with the new operating conditions. Prototype planar pixel sensors for the CMS Inner Tracker with square 50 μm × 50 μm and rectangular 100 μm × 25 μm pixels read out by the RD53A chip were characterized in the lab and at the DESY-II testbeam facility in order to identify designs that meet the requirements of CMS during the High-Luminosity running phase. A spatial resolution of approximately 3.4 μm (2 μm) is obtained using the modules with 50 μm × 50 μm (100 μm × 25 μm) pixels at the optimal angle of incidence before irradiation. After irradiation to a 1 MeV neutron equivalent fluence of Φeq = 5.3 × 1015 cm-2, a resolution of 9.4 μm is achieved at a bias voltage of 800 V using a module with 50 μm × 50 μm pixel size. All modules retain a hit efficiency in excess of 99% after irradiation to fluences up to 2.1 × 1016 cm-2. Further studies of the electrical properties of the modules, especially crosstalk, are also presented in this paper.