Neutron Energy Spectrum Research Articles

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.

Neutronic optimization of thorium-based fuel configurations for minimizing slightly used nuclear fuel and radiotoxicity in small modular reactors

Effective management of slightly used nuclear fuel (SUNF) is crucial for both technical and public acceptance reasons. SUNF management, radiotoxicity risk, and associated financial investment and technological capabilities are major concerns in nuclear power production. Reducing the volume of SUNF can simplify its management, and one possible solution is utilizing small modular reactors (SMR) and advanced fuel designs like those with thorium. This research focuses on studying the neutronic performance and radionuclide inventory of three different thorium fuel configurations. The mass of fissile material in thorium-based fuel significantly impacts Kinf, burn-up, and neutron energy spectrum. Compared to uranium, thorium as a fuel produces far fewer transuranic elements and less long-lived fission products (LLFPs) at the end of the core cycle (EOC). However, certain fission product elements produced from thorium-based fuel exhibit higher radioactivity at the beginning of the core cycle (BOC). Physical separation of thorium and uranium in the fuel block, like seed-and-blanket units (SBU) and duplex fuel designs, generate less radioactive waste with lower radioactivity and longer cycle lengths than homogeneous or mixed thorium-uranium fuel. Furthermore, the SBU and duplex feel designs exhibit comparable neutron spectra, leading to negligible differences in SUNF production between the two.

Characterisation of neutron fields at the n-lab, a fast neutron facility at the University of Cape Town

The n-lab is a fast neutron facility based in the Department of Physics, University of Cape Town, offering collimated neutron beams produced by an MP 320 deuterium–tritium sealed tube neutron generator, and a 220 GBq americium–beryllium radioisotopic source. Characterisations of the spatial and energy distributions of the fast neutron beams were performed using an EJ-301 organic liquid scintillator detector and digital data acquisition system. Neutron energy spectra were obtained through unfolding analyses with MAXED, and a Monte Carlo approach to the propagation of uncertainties was implemented. Measurements of fluence and neutron emission rates were determined through the neutron activation of copper foils and subsequently used to validate the scaling of the unfolded neutron energy spectra.

TRU utilization and MA transmutation in thorium-based fluorinated molten salt fast reactor

In this paper, a thorium-based fluorinated molten salt fast neutron reactor is taken as the research object, and the utilization of thorium resources, reuse of transuranic waste (TRU), and transmutation of minor actinide (MA) in fast reactor are the research purpose. The fast reactor adopts a multi-layer core structure, in which the fuel is divided into an inner fission fuel zone and an outer breeding fuel layer, respectively loaded with fluorinated molten salts LiF+ThF4+XF4 (where X is 233U, TRU, or 233U+MA) and LiF+ThF4. In addition to serving as nuclear fuel, these two molten salts also act as coolants in the primary circuits.Through the cross sections of nuclear reactions, transmutation chain of actinide nuclides, and the particle population density formula of actinide nuclides in nuclear reaction equilibrium, the possible nuclear reactions of actinide nuclides in thorium-based fast reactors are analyzed. The calculated results show that the thorium-based fast reactor has the following characteristics: fast neutron energy spectrum, negative temperature reactivity coefficient, effective utilization of thorium resources, the ability to breed fissle fuel such as 233U, production of high-purity radioisotope 238Pu material, and the ability to incinerate Pu and MAs in the TRU spent fuel.In the 233U fuel scheme, the 233U fuel is breeded, breeding ratio of fissile fuels is 1.21, the reaction products contain very little Np, and almost no Pu, Am, and Cm long-lived radionuclides. In the TRU fuel scheme, the annual capacity to incinerate MAs and Pu is 1072.6 kg, while producing a large amount of 233U and 233Pa fuel (together 1162.5 kg/year).In the 233U+2.0 wt% TRU-MAs fuel scheme, 127.35 kg of MA waste can be incinerated annually, and their transmutation rate is 15.95%, which is equivalent to the annual production of MA nuclear waste from 4.9 conventional 1000 MWe light water reactors. Here, TRU-MAs is the mixtured MA fuels after separation of Pu from the TRU fuel. At the same time, radioactive fuel 238Pu, which can be used for space nuclear power, is produced with an annual yield of 87.89 kg and an isotopic abundance of 87.57%. If the added TRU-MAs fuel is replaced with 2.0 wt% 237Np fluorinated salt, the annual yield and isotope abundance of 238Pu are further increased to 130.24 kg and 95.4%.

The effect of dominance ratio on the statistical convergence of sensitivity in Monte Carlo codes

Sensitivity computation in Monte Carlo-based codes is widely used and has been enhanced significantly. Although sensitivity converges in most reactor configurations, it encounters difficulties in large core designs. We suspect that this issue might be linked to the dominance ratio. To test this hypothesis, we develop a simple benchmark to validate whether statistical convergence of sensitivity depends on dominance ratio, and if it may be linked to other factors, such as neutron energy spectrum. The benchmark’s simplicity enables us to calculate the dominance ratio analytically and use eigenmodes decomposition to compute sensitivity to total fission neutron yields, substantially lowering computational costs. This method results in a good match when compared to the direct method which serves as a reference. We clearly see regular sensitivity convergence speed behavior linking it with the number of latent generations and the dominance ratio. Therefore, we establish a formula to recommend the number of latent generations required for sensitivity to converge, thus significantly saving computational resources.

Measurement and analysis of neutron spectrum responses of ST401 scintillators with different thickness

In the measurement of pulsed neutrons in the MeV energy range, plastic scintillators are one of the most widely used materials, and their neutron energy spectrum responses are key data of pulsed neutron energy spectrum measurement. The neutron energy spectrum responses of ST401 plastic scintillators with 5 different thickness values ranging from 0.5 to 10 mm in an energy range from 0.5 MeV to 100 MeV are measured by using the time-of-flight (TOF) method on the white neutron source (WNS) beamline of the China Spallation Neutron Source (CSNS). The effects of in-beam gamma rays, the gamma flash produced slow component of scintillator, and the pulse width of the neutron source on the measurement of neutron spectrum response are analyzed. Owing to the boundary effect of the finite volume of the scintillator, the neutron energy spectrum response curves of ST401 with different thickness values present approximately logarithmic shape, and proton escape is the main reason for the deviation of the curve from linearity. The thicker the scintillator, the higher the neutron energy deviates from linearity.

Radiation characterization summary of the NETL beam port 1/5 free-field environment at the 128-inch core centerline adjacent location

The characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the University of Texas at Austin Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor for the beam port (BP) 1/5 free-field environment at the 128-inch location adjacent to the core centerline has been accomplished. NETL is being explored as an auxiliary neutron test facility for the Sandia National Laboratories radiation effects sciences research and development campaigns. The NETL reactor is a TRIGA Mark-II pulse and steady-state, above-ground pool-type reactor. NETL is intended as a university research reactor typically used to perform irradiation experiments for students and customers, radioisotope production, as well as a training reactor. Initial criticality of the NETL TRIGA reactor was achieved on March 12, 1992, making it one of the newest test reactor facilities in the US. The neutron energy spectra, uncertainties, and covariance matrices are presented as well as a neutron fluence map of the experiment area of the cavity. For an unmoderated condition, the neutron fluence at the center of BP 1/5, at the adjacent core axial centerline, is about 8.2×1012 n/cm2 per MJ of reactor energy. About 67% of the neutron fluence is below 1 keV and 22% above 100 keV. The 1-MeV Damage-Equivalent Silicon (DES) fluence is roughly 1.6×1012 n/cm2 per MJ of reactor energy.

Radiation characterization summary for the WSMR fast burst reactor environment at the 6-inch location

The characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields for the White Sands Missile Range (WSMR) Fast Burst Reactor, also known as molybdenum-alloy Godiva (Molly-G) has been assessed at the 6-inch irradiation location. The neutron energy spectra, uncertainties, and common radiation metrics are presented. Code-dependent recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. The Molly-G core was designed and configured similarly to Godiva II, as an unreflected, unmoderated, cylindrical annulus of uranium-molybdenum-alloy fuel with molybdenum loading of 10%. At the 6-inch position, the axial fluence maximum is about 2.4×1013 n/cm2 per MJ of reactor energy; about 0.1% of the neutron fluence is below 1 keV and 96% is above 100 keV. The 1-MeV Damage-Equivalent Silicon (DES) fluence is estimated at 2.2×1013 n/cm2 per MJ of reactor energy. The prompt gamma-ray dose is roughly 2.5E+03 rad(Si) per MJ and the delayed gamma-ray dose is about 1.3E+03 rad(Si) per MJ.

Exploration of genetic algorithms to build a balanced neutron spectra dataset useful to train unfolding techniques based on artificial neural networks

Diverse domains need neutrons unfolding technics to assess the incident neutron energy spectrum. Examples are radiation protection, nuclear reactor physics or criticality safety. Traditionally, methods based on the Bayesian approach requires an initial guess of the solution which may significantly impact the unfolding result. This work proposes a novel method for neutron spectrum reconstruction using machine learning (ML) techniques trained on a large dataset. To ensure the ML algorithm to perform on a large domain of application particular attention has been paid to the dataset creation. We propose a comparison of two methods of building large dataset where the most adequate solution is obtained using a dynamic genetic algorithm (GA). This GA targets optimal combinations of 48 parameters to generate a variety of neutron spectra. The resulting dataset is then used to train a new convolutional neural network architecture for unfolding neutron spectra. Obtained performance metrics of the tested architecture show high efficiency and emphasize the added value of the built dataset.

The computer model of a neutron fluxes forming system on a linear electron accelerator

A computer model of the system for generating neutron fluxes at the output of a linear electron accelerator has been developed in the Geant4 and PhysList QGSP BIC HP programming environment. With the help of the model, a number of virtual experiments on 107 primary neutrons were carried out. The dependence of the ratio of the number of neutrons, incident on the detector with a reflector to the number of neutrons without a reflector on the radius of curvature of the hemispherical dome of the reflector is determined, which makes it possible to establish the optimal dimensions of the working area. The energy spectra of neutrons incident on the detector are determined. It has been established that when using a graphite reflector 30 × 30 × 30 cm in size with a radius of curvature of the hemispherical dome of 5 cm, the number of neutrons at the location of the detector increases by 16.9 %, and the neutron background in the accelerator bunker decreases by 2.5 times, which is in good agreement with the real experiment made. Such a decrease in the radiation background, according to sanitary standards, will make it possible to increase the accelerator current by 2.5 times and increase the neutron flux.

Neutronics analysis for conceptual design of target system based on a deuteron accelerator-driven fusion neutron source

A Breeding Blanket (BB), producing tritium and heat energy converted to electricity, is a critical component for the successful Fusion Demonstration Reactor (DEMO). A dedicated neutron source is required to validate the performance of the DEMO BB in a continuous long-term fusion-like environment. Therefore, a pre-conceptual design study has been carried out for the Integrated Breeding Test Facility (IBTF) to evaluate the DEMO BB performance. The IBTF is based on a 40 MeV deuteron accelerator-driven system with a maximum current of 10 mA. A solid Beryllium (Be) target with a 20 cm width and height is used to irradiate the tritium breeding unit (TBU).This study conducted a neutronics analysis of the Be target system of the IBTF. The produced neutron energy spectrum and yield were evaluated on a beam energy of 30 to 50 MeV. The analysis results show that the forward direction neutron yield with the 40 MeV deuteron is the highest at 4.92E + 15 n/s, and the neutron ratio of 10–20 MeV is also the highest. A Blistering Mitigation Layer (BML) was added to prevent the blistering effect caused by the deuteron beam within the Be target. A comparison was done for the performance evaluation of the BML candidates vanadium, palladium, tantalum, and niobium. There were no significant differences in the forward direction neutron yield for the BML options. Vanadium was the most beneficial based on the activation analysis results. Therefore, vanadium with a 0.3 mm thickness was added behind the Be target with a 5 mm thickness. In addition, neutronics analysis was conducted for the test cell shielding designs. Concrete 3.5 m thick was required to satisfy a dose of less than 100 uSv/h during the beam operation. When evaluating the dose rate during maintenance, a dose along the cooling pipes was observed due to the activation of the pipes. The neutron flux with the cooling pipes was reduced by using B4C in the maintenance cell shielding, the dose rate was below 100 uSv/h in the test cell.Deuteron and neutron transport calculations were carried out using MCNP6 with the TENDL2019 cross-section library. Activity analysis and gamma source production were performed using FISPACT-Ⅱ with the TENDL2017 cross-section library.

DEVELOPMENT AND TESTING OF A COMPUTER MODEL OF A SYSTEM FOR FORMING NEUTRON FLUXES ON A LINEAR ELECTRON ACCELERATOR

A computer model of the system for forming neutron fluxes at the output of a linear electron accelerator has been developed in the Geant4 programming environment. The system for forming neutron fluxes consists of a monoenergetic neutron source, a graphite reflector, a polyethylene moderator, lead protection and detector. With the help of the model, a number of virtual experiments on 107 primary neutrons were carried out for the energy range of the neutron source from 0.1 to 1.1 MeV. The dependence of the ratio of the number of neutrons falling on the detector with a reflector to the number of neutrons without a reflector is determined, on the energy of the neutron source. Energy spectra of neutrons falling on the detector are determined. It was established that the graphite reflector not only increases the number of particles falling on the detector, but also moderated the energy spectrum of neutrons. The neutron background in the accelerator bunker when using a reflector decreases from 2 to 7 times, depending on the energy of the neutron source.

A simulation study of a collimating shield structure in a vertical neutron camera on HL-3

A shield structure with five collimating tubes has been designed in a vertical neutron camera (VNC) on HL-3. Polyethylene is used to screen scattered neutrons, and the lead layer is applied to obstruct the stray X-rays and γ-rays in the environment, as well as secondary γ-rays produced by neutrons interacting with materials. The shielding effect of the structure is studied by using GEANT4. The energy spectra of the incident neutrons and the rate of direct neutrons at the end of five channels are obtained, which demonstrates that the design of shield achieve the criteria as we excepted.

The Computer Model of a Thermal Delayed Neutron Fluxes Forming System for Nuclear Medicine

In the work the computer model of a cell of a system for generating fluxes of therapeutic beams of delayed neutrons, based on the use of delayed fission neutrons, was developed in the Geant 4 environment. The principle of such a neutron source is that when a powerful electron beam interacts with a combined tungsten target and a target containing fissile material, a fission reaction occurs; as a result of which neutrons are emitted. If we move a target activated in this way several tens of meters into a neutron flux generation system consisting of a heater, protection, collimator and reflector, we will obtain a compact neutron source for nuclear medicine. A significant advantage of such a neutron source is the absence of gamma background from the electron accelerator and the combined target, and a bulky protection system is not required. In the Geant 4 environment, the geometry of this cell was developed and a series of experiments were carried out with 107 neutrons. The QGSP BIC HP physical sheet was used. A study of neutron energy spectra showed that more than half of the neutrons whose fluxes are formed using such a cell of the formation system have an energy <100 keV, which is suitable for use for therapeutic purposes. Analysis of the data obtained in a computer experiment made it possible to develop a modified cell of the system for generating streams of therapeutic beams of delayed neutrons, which differs from the basic one by the presence of a solid polyethylene moderator with holes for activated targets and a graphite reflector. Analysis of the data obtained showed that in this case the number of thermal neutrons hitting the detector increases 10 times compared to the base cell, and the energy of 80% of the particles does not exceed 5 keV, which is much better suited for therapeutic purposes.

Measurement of neutron spectra for various thicknesses of concrete and steel shielding at 24-GeV/c proton beam facility using Bonner sphere spectrometer

ABSTRACT Neutron energy spectra down to thermal energy were measured using a Bonner sphere spectrometer (BSS) for various thicknesses of concrete and steel shielding at the CERN/CHARM facility, where high-energy neutrons were produced by 24-GeV/c protons incident on a thick copper target. The thicknesses of the concrete and steel shielding blocks ranged from 40 cm to 200 cm and from 20 cm to 80 cm, respectively. The BSS consisted of a spherical 3He proportional counter and five polyethylene moderators with diameters of 7.62 cm, 10.2 cm, 12.7 cm, 17.8 cm, and 24.1 cm, respectively. In addition, polyethylene moderators combined with a lead or copper inner shell were used to increase the sensitivity to high-energy neutrons. The neutron energy spectra were deduced using an unfolding method. The initial guesses were obtained using the PHITS code for each experimental geometry. The response function for the BSS was determined using the MCNP6.2 code with JENDL-4.0/HE. The neutron energy spectra over the entire energy region from 10−4 eV to 10 GeV were successfully obtained for the different shielding conditions. The validity of the response function and the contribution of each moderator are discussed referring to previous studies and tests at the standard neutron fields of AIST.

Irradiation device development with changing neutron spectrum

This paper presents a sketch drawing of the developed design of an irradiating device, which allows to regulate the neutron energy spectrum in it, as well as the results of neutron-physical calculations with predicted values of the range of neutron flux density variation in different energy groups. It is shown that the fraction of thermal and over-thermal neutrons in the dry irradiation channel increases with increasing beryllium filling packing fraction. While the fraction of fast neutrons (> 1.15 MeV) decreases. In the wet irradiation channel, the dependence of thermal and over-thermal neutrons is similar to that in the dry channel, but the effect is much smaller.