Neutron Energy Spectrum Research Articles

NeuSIM4: A comprehensive GEANT4 based neutron simulation code

A new neutron SIMulation program based on the versatile GEANT4 toolkit, neuSIM4, has been developed to describe interactions of neutrons in the NE213 liquid scintillator from 0.1 to 3000 MeV. neuSIM4 is designed to accommodate complicated modern detector geometry setups with multiple scintillator detectors, each of which can be outfitted with more than one photo-multiplier. To address a broad spectrum of neutron energies, two new neutron interaction physics models, KSCIN and NxQMD, have been implemented in GEANT4. For neutrons with energy below 110 MeV, we incorporate a total of eleven neutron induced reaction channels on hydrogen and carbon nuclei, including nine carbon inelastic reaction channels, into KSCIN. Beyond 110 MeV, we implement a neutron induced reaction model, NxQMD, in GEANT4. We use its results as reference to evaluate other neutron-interaction physics models in GEANT4. We find that results from an existing cascade physics model (INCL) in GEANT4 agree very well with the results from NxQMD, and results from both codes agree with new and existing light response data. To connect KSCIN to NxQMD or INCL, we introduce a transition region where the contribution of neuSIM4 linearly decreases with corresponding increased contributions from NxQMD or INCL. To demonstrate the application of the new code, we simulate the light response and performance of a 2 × 2 m2 neutron detector wall array consisting of 25 2m-long scintillation bars. We are able to compare the predicted light response functions to the shape of the experimental response functions and calculate the efficiency of the neutron detector array for neutron energies up to 200 MeV. These simulation results will be pivotal for understanding the performance of modern neutron arrays with intricate geometries, especially in the measurements of neutron energy spectra in heavy-ion reactions.

Machine Learning Based Compton Suppression for Nuclear Fusion Plasma Diagnostics

Diagnostics are critical on the path to commercial fusion reactors, since measurements and characterisation of the plasma is important for sustaining fusion reactions. Gamma spectroscopy is commonly used to provide information about the neutron energy spectrum from activation analysis, which can be used to calculate the neutron flux and fusion power. The detection limits for measuring nuclear dosimetry reactions used in such diagnostics are fundamentally related to Compton scattering events making up a background continuum in measured spectra. This background lies in the same energy region as peaks from low-energy gamma rays, leading to detection and characterisation limitations. This paper presents a digital machine learning Compton suppression algorithm (MLCSA), that uses state-of-the-art machine learning techniques to perform pulse shape discrimination for high purity germanium (HPGe) detectors. The MLCSA identifies key features of individual pulses to differentiate between those that are generated from photopeaks and Compton scatter events. Compton events are then rejected, reducing the low energy background. This novel suppression algorithm improves gamma spectroscopy results by lowering minimum detectable activity (MDA) limits and thus reducing the measurement time required to reach the desired detection limit. In this paper, the performance of the MLCSA is demonstrated using an HPGe detector, with a gamma spectrum containing americium-241 (Am-241) and cobalt-60 (Co-60). The MDA of Am-241 improved by 51% and the signal to background ratio improved by 49%, while the Co-60 peaks were partially preserved (reduced by 78%). The MLCSA requires no modelling of the specific detector and so has the potential to be detector agnostic, meaning the technique could be applied to a variety of detector types and applications.

Observation of energetic ion anisotropy using neutron diagnostics in the Large Helical Device

Energetic ion anisotropy was observed by tangential sightline compact neutron energy spectrometers (CNESs) in tangential neutral beam heated deuterium plasmas in Large Helical Device. Significant upper and lower energy shifts in D–D neutron energy from 2.45 MeV were measured according to the beam ion injection directions and CNES sightline using a conventional liquid scintillation detector with the unfolding technique and a novel Cs2LiYCl6:Ce with a 7Li-enrichment (CLYC7) scintillation detector without unfolding. The observed neutron energy spectrum was compared with that predicted by a numerical simulation based on orbit following models. Numerical simulation revealed that the Doppler shift in D–D neutron energy results from energetic ion anisotropy.

Investigation of Cs27LiYCl6:Ce scintillator energy response for D-D fusion neutron spectrometer

During plasma experiments on EAST (Experimental Advanced Superconducting Tokamak), the discharges heated with neutral beam injection (NBI) for plasma results in the production of 2.45 MeV deuterium-deuterium (D-D) fusion neutrons. In order to measure the energy spectrum of D-D fusion neutrons, a new Cs27LiYCl6:Ce (CLYC-7) crystal with 7Li-enrichment has been developed and fabricated into a CLYC-7 scintillation detector. Before utilizing this detector for studying neutron energy spectra, it is crucial to explore the response function of the detector within a relevant neutron energy range. Therefore, the detector has been experimented using the 4.5 MeV electrostatic accelerator at Peking University to investigate its response to fast neutrons, energy resolution, and pulse discrimination capability in the neutron energy range 2.07 MeV–5.51 MeV. The neutron reactions occurring within the CLYC-7 scintillator were simulated using MCNPX, focusing on the reactions 35Cl(n,p) and 35Cl(n,α), as well as the energy response curve and the contributions of different particles (protons, alphas) to the neutron spectrum. The experimental results were successfully validated on EAST, where the 2.45 MeV fusion neutron energy spectrum was measured. That demonstrates the ability of the CLYC-7 scintillation detector to diagnose the fusion neutron energy spectrum for deuterium plasma discharge.

Simultaneous measurements for fast neutron flux and tritium production rate using pulse shape discrimination and single crystal CVD diamond detector

This paper presents the development of a simultaneous measurement method for fast neutron energy spectra and tritium production rates within mixed radiation fields using a single crystal chemical vapour deposition diamond detector combined with a lithium fluoride (LiF) foil. The method involves the separation of pulses with rectangular shapes and the determination of the depth position within the single crystal diamond (SCD) struck by fast neutrons or nuclear reaction products including recoil tritons from the LiF foil based on pulse width, extracting pulse events occurred at the specific bulk region and the surface region of the SCD. Subsequently, unfolding techniques were employed to analyse the energy deposition spectrum of pulses at the specific bulk region which are induced only by fast neutrons, allowing the deduction of the fast neutron energy spectrum. To evaluate the tritium production rate, the energy deposition spectrum of pulses from events occurring at the SCD surface facing the LiF foil was analysed. By estimating the energy deposition spectrum solely induced by fast neutrons striking the SCD surface and subtracting it from the energy deposition spectrum of events at the SCD surface, the contribution of energetic ions, such as recoil tritons generated by the 6Li(n,α)3H reaction in the LiF foil, was determined. The fast neutron flux and tritium production rate obtained through this study were consistent with particle transport calculations, demonstrating the successful development of a method suitable for performance testing of fusion reactor blankets.

Shielding mass estimation model for gas-cooled space nuclear reactors

The shielding is expected to be “mass-effective” for space nuclear power systems (SNPs) because it accounts for a significant portion of the total weight. Hence, the shield mass should be estimated rapidly but accurately especially in the early stage of SNPs’ design. In this paper, a simple mass estimation model was developed based on the preliminarily optimized shield design for the Jupiter Icy Moons Orbiter (JIMO) reactor. In the model, the reactor-induced 1 MeV equivalent silicon neutron fluence and photon absorbed dose at the unshielded payload were predicted by a circular surface source hypothesis at the core bottom, and the proportional coefficients were determined by the Monte Carlo calculation. Moreover, the attenuation factors of the shield were derived by solving the neutron/gamma coupled transport process within the shield body, and the correlations were established between the attenuation effect and the shield thicknesses. Afterwards, when given the radiation limits at the payload, the shield geometry and associated mass could be easily determined. Several detailed calculations corresponding to the variation of shield design were used to verify the estimation model. Those variations include the thickness of the Be-B4C layers, the effective density of the B4C and the number of the shielding layers. Additionally, the potential changes of the reactor such as the neutron energy spectrum and the structural materials in the core were also considered to evaluate the applicability of the model. Totally, the estimation model agrees with the detailed calculation within 15 % in a wide range, which suggests the model can be incorporated into the system code to help optimize the overall performance of the SNPs.

Influence mechanism of lithium surface oxidation on neutron yield of lithium target for accelerator-based neutron source

Lithium is generally adopted as the target to generate epithermal neutrons based on 7Li(p,n)7Be nuclear reaction for the application of accelerator-based boron neutron capture therapy (AB-BNCT). The stability of neutron yields and neutron energy spectrum are key factors for the therapeutic effects of the AB-BNCT. Owing to the active chemical properties of lithium, the surface oxidation formation of lithium may influence the soundness of the target system and the stability of neutron yields. Experimental and simulation methods were performed for a better understanding of the passivation layer formation after air exposure and the influence of the passivation layer on the neutron yield and energy spectrum. On one hand, X-ray photoelectron spectroscopy was adopted to study the chemical states of lithium after air exposure. On the other hand, particle and heavy ion transport code system (PHITS) was used to evaluate the effect of surface chemical states variation on neutron yield and energy spectrum. The simulation results indicate that the formation of Li2O with a thickness of 2 μm could mainly influence the neutron yields at the neutron emission angle of 20–90° with a maximum reduction of ∼ 10.4%.

Optimization of photoneutron source for use in subcritical reactors

This work is an effort to maximize the number of output neutrons of a photoneutron target for 100 MeV incident electrons. The work steps were; to select the appropriate material, select the appropriate shape, and find the optimized dimensions. The simulations of this work were done with MCNPX2.6 simulation code. At first, U-235 was selected as the best material for the target, among some heavy atoms from the number of outlet neutron point of view. Then the best shape for the target was selected from geometric shapes that have been considered for the target and the dimensions of the target were optimized. After these parts, the energy spectrum of outlet neutrons was estimated and after that, the deposited energy of neutrons, electrons, and photons in the target was estimated and drowned with Tec plot.

Neutron spectroscopy for Boron Neutron Capture Therapy beams characterization

Boron Neutron Capture Therapy (BNCT) is a therapeutic treatment for malignant tumors that utilizes the nuclear reactions that happen when thermal neutrons are captured by boron-10 atoms to selectively destroy designated cells. Boron-10 atoms are biochemically accumulated inside the tumor target, which is then irradiated with thermal neutrons.In recent years, the possibility to obtain accelerator based intense neutron beams has given a boost to the diffusion of BNCT also in Europe, removing the need of nuclear reactors. In this contest, the monitoring and characterization of the epithermal neutron beams dedicated to BNCT becomes an important issue.The directional neutron spectrometer called NCT-WES (Neutron Capture Therapy Wide Energy Spectrometer) is a single moderator neutron spectrometer composed of a polyethylene cylinder embedding six semiconductor-based detectors positioned at different depths along the cylinder axes. The position of the six detectors is studied in order to maximize the response of each one in a selected neutron energy range. The unfolding of the six simultaneous readings allows to reconstruct the incoming neutron energy spectrum as in a parallelized Bonner Sphere System. A cylindrical collimator situated in the front of the spectrometer makes the instrument sensitive to neutrons coming only from a given direction, which allows to exclude the contribution of the room scattered radiation. The experimental validation of the spectrometer, obtained through several measuring campaigns, is reported and discussed.

Preliminary design and analysis activities of the deuteron accelerator target for tritium breeding unit test

A breeding blanket (BB) is a key component that multiplies neutrons and generates tritium in the fusion demonstration reactor. Evaluation of the performance and reliability of the BB is a very important engineering task, and it is going through the ITER TBM program. The Korean DEMO program needs to complement the ITER TBM program to validate the BB, and the test facility is necessary to validate the BB in the continuous wave operation. Based on this need, the Korea Institute of Fusion Energy has initiated a pre-conceptual study to verify the long-term performance and reliability of the BB.This study deals with neutronics and engineering analyses of a solid beryllium (Be) target that generates neutrons through a deuteron accelerator for the fusion neutron source. In the neutronics analysis, the yield and energy spectrum of neutrons generated according to the incident particles and energy were analyzed. As a result, in the case of 40 MeV deuteron, the yield of neutrons with 10∼20 MeV similar to the fusion neutron was the highest in the forward direction of 1.08E+15 n/s. In addition, vanadium (V) was selected as the blistering mitigation layer (BML), and the thickness of the Be target and the BML were optimized through neutronics analysis to minimize the blistering. In the engineering analysis, a preliminary conceptual design of the target to cool the nuclear heating caused by the nuclear reaction was performed. To improve the cooling capacity of the target composed of Be-V-CuCrZr with the pressurized water coolant, the design of the target was optimized and its thermal integrity was evaluated. Next, thermal stress was evaluated by coupling the temperature distribution of the target obtained through thermal analysis with structural analysis. Through this, the preliminary design concept of the target that satisfies thermal and mechanical integrity requirements was derived.

Unfolding neutron spectra using splines

Organic scintillators detect neutrons by scattering reactions with hydrogen nuclei (protons) in organic compounds. The recoil protons then produce the light output from the scintillator. The energy spectrum of neutrons impinging on the scintillator has to be determined from this light output using a so-called unfolding algorithm. However, this problem is ill-posed and therefore difficult to solve. In this study, a new method for unfolding neutron spectra is presented. This method uses B-splines approximation of neutron energy spectrum. Results of this method are shown for test data and experimental data and compared to the results of the previously developed unfolding method.

Thermal neutron dose measurements using TLD-100 detectors in the IPEN/MB-01 reactor core

Considerable experimental effort has been aimed at uncovering a reliable way to perform a dosimetric assessment in mixed radiation fields. In fields composed by gammas and neutrons, TLD dosimeters are usually applied to execute such measurements, although there is no consensus on the most favorable strategy to employ them. In this context, TLD-100 measurements within two different core configurations of the IPEN/MB-01 research reactor and Monte Carlo simulations have been used to investigate the behavior of those detectors in multiple mixed radiation fields, deriving a methodology to evaluate the dose deposition in the dosimeter by different gamma and neutron energy spectra and intensities. A surprising outcome is the linear neutron dose response shown by TLD-100 even irradiated by so distinct irradiation fields.

Study of accelerator-based epithermal neutron flux depending on Li and Be targets

The Li(p,n) and Be(p,n) reactions are widely used as accelerator-based neutron sources for boron neutron capture therapy (BNCT). Because the energy of the neutrons produced by these reactions is high for BNCT, a moderator is required to reduce the energy to the epithermal energy region. The neutron yields and energy spectra derived from the Li(p,n) and Be(p,n) reactions vary with the incident proton energy; a high proton energy results in a high neutron yield. However, a high proton energy can also increase the neutron energy, and in this case, a high neutron energy requires a thick moderator, which decreases the epithermal neutron flux. Notably, these tradeoffs are not completely clear. Furthermore, the upper limit of the epithermal neutron energy used in different BCNT facilities is not the same. Therefore, epithermal neutron flux estimation with applying an appropriate moderator is necessary for comparing epithernal neutron flux depending on the proton energy, target type, and epithermal neutron range definition.In this study, GEANT4 simulations were performed to determine effective moderator materials and thicknesses. After applying a moderator, we examined the epithermal neutron flux by changing the conditions that can influence the epithermal neutron flux, such as the proton energy, target species (Li and Be), and different definitions of the epithermal neutron region. The results showed that the epithermal neutron flux increases with the increasing proton energy until a certain level for both Li and Be targets. For proton energies above 3 MeV for the Li target and above 7 MeV for the Be target, the epithermal neutron flux decreases or its increase rate gradually decreases. And, when the upper energy limit of the epithermal neutron range is changed from 10 keV to 20 keV or 40 keV, the epithermal neutron flux is more than 109 cm−2 s−1 at a proton energy of 3 MeV for Li target and 8 MeV for Be target.

A compact scintillator-based detector for high energy neutron spectrometry

A compact neutron spectrometer consisting of a 6 mm × 6 mm × 120 mm prism of EJ-276 plastic scintillator capable of pulse shape discrimination coupled to a single silicon photomultiplier is described. In principle this device will be capable of spectroscopic measurements exceeding 100 MeV using unfolding analyses with detector response functions measured at neutron time-of-flight facilities. A characterisation of the device is described and illustrated with measurements of the detector response made at three fast neutron facilities with energies between 5.0 MeV and 63.5 MeV. Nearly mono-energetic detector response functions were used to produce neutron energy spectra by unfolding light output distributions measured with the detector in the energy range of interest.

Maxwell–Boltzmann-like neutron spectrum production for Maxwellian averaged cross sections measurements

Over the years, Maxwellian Averaged Cross Sections (MACS) have been measured by neutron activation, providing a neutron energy spectrum resembling the one found inside the stars. Recently, a new method has been proposed to produce stellar spectra at different stellar temperatures (a Maxwell–Boltzmann neutron energy distribution), employing the 7Li(p,n)7Be reaction. The method is based on the idea of shaping the proton beam energy to shape the neutron beam spectrum. This method was applied to obtain a well-reproduced Maxwell–Boltzmann neutron spectrum (MBNS) at kT=28 keV. An initial proton energy of 3170 keV and an aluminum foil as a proton energy shaper were employed. Differential angular neutron energy distributions from 0 to 90 degrees in 10∘ steps were measured to obtain the 0∘–90∘ integrated neutron spectrum over a neutron flight path of 50 cm. This manuscript reports on the measurement results, confirming the method’s capability, and suggests the approach for producing a high-quality MBNS at kT=28 keV.

Large-scale kinetic simulations of colliding plasmas within a hohlraum of indirect-drive inertial confinement fusion.

The National Ignition Facility has recently achieved successful burning plasma and ignition using the inertial confinement fusion (ICF) approach. However, there are still many fundamental physics phenomena that are not well understood, including the kinetic processes in the hohlraum. Shan etal. [Phys. Rev. Lett. 120, 195001 (2018)0031-900710.1103/PhysRevLett.120.195001] utilized the energy spectra of neutrons to investigate the kinetic colliding plasma in a hohlraum of indirect drive ICF. However, due to the typical large spatial-temporal scales, this experiment could not be well simulated by using available codes at that time. Utilizing our advanced high-order implicit PIC code, LAPINS, we were able to successfully reproduce the experiment on a large scale of both spatial and temporal dimensions, in which the original computational scale was increased by approximately seven to eight orders of magnitude. Not only is the validity of the explanation of the experiment confirmed by our simulations, i.e., the abnormally large width of neutron spectra comes from beam-target nuclear fusions, but also a different physical insight into the source of energetic deuterium ions is provided. The acceleration of deuterium ions can be categorized into two components: one is propelled by a sheath electric field created by the charge separation at the onset, while the other is a result of the reflection of the potential of the shock wave. The robustness of the acceleration mechanism is analyzed with varying initial conditions, e.g., temperatures, drifting velocity, and ion components. This paper might serve as a reference for benchmark simulations of upcoming simulation codes and may be relevant for future research on mixtures and entropy increments at plasma interfaces.

Novel 3D-Printed lead-free radiation protection apron in the medical X-ray and thermal neutron energy range

In this study, we employed 3D printing technology to fabricate poly lactic acid (PLA) polymer samples infused with gadolinium oxide nanoparticles at additive rates of 10% and 20%. The objective was to explore their potential as radiation shielding aprons within the medical X-ray and thermal neutron energy spectrum. To facilitate comparisons, a PLA polymer sample with no additive was also produced. The homogeneity and well-defined structures of the PLA samples were observed using SEM and EDS analyses. Additionally, the excellent thermal stability of the proposed test samples was reported. In terms of gamma-ray shielding, there is a remarkable consistency between experiment, theory and simulation outcomes with a maximum discrepancy of approximately 5%. P-PLA-Gd20 sample exhibits attenuation capabilities against X-rays to a level that could serve as an alternative to lead. Additionally, the thermal and fast neutron attenuation effectiveness of the prepared samples were determined. A shielding effectiveness of 100% against thermal neutrons was achieved using a 10 mm sample thickness and the P-PLA-Gd20 sample. The findings consistently highlight the efficacy of the proposed polymer sample with a 20% gadolinium oxide nanoparticle additive, positioning it as a viable and promising alternative to traditional lead aprons.

Design and performance evaluation of a compact thermal and fast neutron spectrometer

Measurement of the space neutron radiation environment has put forward new requirements for the miniaturization and lightweight design of neutron spectrometers. In this paper, a compact broad-spectrum neutron spectrometer was designed, in which the fast neutron detector was based on the EJ-254 scintillator-coupled with SiPMs array, and the thermal and epithermal neutron detectors were based on the NaI(Tl + Li) (NaIL) scintillator coupled with SiPMs array encapsulated in Al and Cd shells, respectively. The overall weight of the neutron spectrometer was only 2.4 kg, and its dimensions were 154 mm × 123 mm × 115 mm. The fast neutron energy spectrum was obtained based on the capture-gated method and the Maximum Likelihood Expectation Maximization (MLEM) unfolding algorithm; the counts of thermal and epithermal neutrons were obtained based on the optimized frequency-domain Pulse Shape Discrimination (PSD) method and cadmium difference method. The proposed neutron spectrometer with all applied methods was tested as a whole system with a252Cf and a241Am–Be source. The spectrum result was compared to an ISO standard source. The accuracy of fast part of 0.5–10 MeV was 5.9% in definition of Relative Root Mean Squared Error (RRMSE). The thermal and epithermal neutron counts were far higher than standard spectrum, and we proved it being effected by laboratory introduced neutron scattering through a simulation, thus the thermal and epithermal flux measurements was believed to be decent. The Figure of Merit (FoM) of the NaIL's PSD neutron gamma discrimination were 2.57 and 2.90 respectively.

Characterization and mapping of the neutron fields around Bruce Power’s 177Lu isotope production system

Bruce Power operates a first-of-its-kind isotope production system (IPS) that enables continuous production of 177Lu within Canada Deuterium Uranium (CANDU) commercial power reactors. Located on the reactivity mechanisms deck of Unit 7, just outside of reactor containment but in close proximity to the primary heat transport (PHT) pumps, this facility offers unique advantages for 177Lu production. However, employees working in this area encounter a radiation hazard which consists primarily of photoneutrons. These originate from the base of the PHT pumps and are only present when the reactor is operating. This study evaluates neutron exposure at Bruce Power’s IPS by using a nested neutron spectrometer (NNS) to determine the neutron energy spectra and absolute dosimetric quantities such as the ambient dose equivalent, H*(10). The results from the NNS are then compared to surveys performed by a portable neutron rem meter (Model NP-2 by Nuclear Research Corporation), routinely used by Bruce Power staff for workplace monitoring. While the Model NP-2 generally showed consistent results across locations, a 50% dose correction factor was identified when operators were harvesting 177Lu from the IPS. This finding highlights an opportunity to reduce the neutron dose that is assigned to operators when producing 177Lu.

Neutron spectra from photonuclear reactions: Performance testing of Monte-Carlo particle transport simulation codes

The production of neutrons in photon-induced nuclear reactions in the giant-dipole-resonance energy domain remains a topic of high interest for various applications, including the activation and decommissioning of electron accelerator facilities, the detection of illicit materials for homeland security, and the evaluation of neutron dose received by patients during radiotherapy treatments. General-purpose Monte-Carlo (MC) simulation codes for particle transport are intensively used to account for photoneutron production in these applications. However, due to the current scarcity of measured photoneutron energy spectra in the literature, experimental validation of MC-simulated photoneutron energy distributions is not always feasible. Therefore, a critical benchmark among simulation results from various MC codes presently appears as the only option to systematically assess their capabilities in accurately simulating photoneutron production for nuclear reactions of interest. In this work, neutron energy spectra from several targets under irradiation by 20 MeV photons are simulated, employing various state-of-the-art MC codes (FLUKA, Geant4, MCNP6, and PHITS) in their default or generally employed settings. A detailed analysis of the simulated neutron spectra allows one to not only assess the performance of various MC codes in applications such as those mentioned above, but also to partially gauge the incurred systematic uncertainty, and to highlight the present need for more comprehensive evaluated nuclear data in this domain. Ultimately, this work suggests that more prudence is required when using MC codes for applications where photonuclear reactions play a dominant role and where not only the production rate but also the energy spectrum of the emitted neutrons matters.