Neutron Energy Spectrum Research Articles

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.

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.

Effect of magnetron sputtering power on structure and properties of Ta films on Cu substrate of neutron production target for accelerator-based boron neutron capture therapy (AB-BNCT)

Tantalum, with a high hydrogen diffusion coefficient, can be used as an interlayer for neutron production target in Accelerator-based Boron Neutron Capture Therapy (AB-BNCT) to prevent hydrogen embrittlement of Cu substrate. Magnetron sputtering technique was employed to deposit Ta films on the Cu Substrate, with sputtering power being a key parameter. This study investigated the effect of sputtering power on the properties of Ta films deposited on Cu substrate through several methods and Monte Carlo simulations (PHITS and SRIM). A turning point was observed at 250 W, beyond and below which increasing sputtering power led to the increase of Ta film thickness, potentially due to changes in the crystalline structure during deposition. The Ta film exhibited superior surface performance at sputtering powers of 300 W and 350 W. The chemical state of Ta was not affected by sputtering power. Thermal conductivity was observed to decrease after Ta deposition. The thickness of Ta film in this study had a negligible effect on neutron energy spectra and angular distribution. Moreover, a Ta film thickness of 20 μm prevented proton deposition and hydrogen embrittlement in the Cu substrate. These findings have important implications for designing and fabricating neutron production target in AB-BNCT.

A Novel Technique of Extracting UCN Decay Lifetime from Storage Chamber Measurements Dominated by Scattering Losses

The neutron’s lifetime is a critical parameter in the standard model. Its measurements, particularly measurements using both beamline and ultracold neutron storage techniques, have revealed significant tension. In this work, we review the status of the tension between various measurements, especially in light of the insights provided by the β-decay correlation measurements. We revisit the lifetime measurement in a material storage chamber, dominated by losses from scattering off the walls of the storage chamber. The neutron energy spectra and associated uncertainties were, for the first time, well-characterized using storage data alone. Such models have applications in the extraction of the mean time between wall bounces, which is a key parameter for neutron storage disappearance experiments in search of neutron oscillation. A comparison between the loss model and the number of neutrons stored in a single chamber allowed us to extract a neutron lifetime of τn*=880(+158/−78)stat.(+230/−114)sys.s (68.3% C.I.). Though the uncertainty of this lifetime is not competitive with currently available measurements, the highlight of this work is that we precisely identified the systematic sources of uncertainty that contribute to the neutron lifetime measurements in material storage bottles, namely from the uncertainty in the energy spectra, as well as from the storage chamber surface parameters of the Fermi potential and loss per bounce. In doing so, we highlight the underestimation of the uncertainties in the previous Monte Carlo simulations of experiments using the technique of ultracold neutron storage in material bottles.

Measuring higher-order moments of neutron-time-of-flight data for cryogenic inertial confinement fusion implosions on OMEGA

Ion temperatures serve as an important diagnostic for inertial confinement fusion (ICF) implosions. In direct-drive ICF experiments on OMEGA, neutron-time-of-flight (nTOF) data are used to infer the ion temperature of the fusing plasma produced in the implosion experiment. The analysis of the nTOF data requires an assumption about the shape of the underlying source signal. Since the source nTOF signal is a near-replica of the neutron energy spectrum, an ideal Gaussian shape, corresponding to the neutron energy spectrum of a uniform temperature plasma, is routinely employed. However, spatial and temporal variations of the ion temperature in the plasma give rise to higher-order moments, which were first described by Munro [Nucl. Fusion 56, 036001 (2016)]. In this work, we show a simpler alternative analysis to derive moments of the neutron energy spectrum for a plasma with variations in ion temperature. We also present a revised analysis of measured nTOF signals that uses a model with an additional degree of freedom to take into account the effect of ion temperature variations on the shape of the spectrum. Compared to presently used nTOF analysis, the revised analysis yields on average ≈2× more accurate fits to the data and up to 15% higher ion temperatures for cryogenic experiments. Furthermore, we quantify the ion temperature inflation caused by radially symmetric fluid flows, which are present even in a symmetric implosion, and which serve as a lower bound on the ion temperature inflation in real implosions.

Analyzing the neutron and γ-ray emission properties of an americium–beryllium tagged neutron source

Americium–beryllium (AmBe), a well-known tagged neutron source, is commonly used for evaluating the neutron detection efficiency of detectors used in ultralow background particle physics experiments, such as reactor neutrino and diffuse supernova neutrino background experiments. In particular, AmBe sources are used to calibrate neutron tagging by selecting the 4438 keV γ-ray signal, which is simultaneously emitted with a neutron signal. Therefore, analyzing the neutron and γ-ray emission properties of AmBe sources is crucial. In this study, we used the theoretical shape of a neutron energy spectrum, which was divided into three parts, to develop models of the energy spectrum and verify the results using experimental data. We used an AmBe source to measure the energy spectra of simultaneously emitted neutrons and γ-rays and determine the emission ratio of the neutrons with and without γ-ray emission. The measured spectrum was consistent with that obtained from the simulated result, whereas the measured emission ratio was significantly different from the corresponding simulated result. Here, we also discuss the feasibility of determining the neutron emission rates from the spectra divided into three parts.

Macroscopic and microscopic investigation of maximum effectiveness of proton-boron capture therapy using Monte Carlo simulation

Proton-boron capture therapy (PBCT) has been proposed as a new method to enhance effectiveness of proton therapy. There is, however, no consensus on its potential effectiveness. Most theoretical studies have been done at a macroscopic level. At cellular dimensions, however, the dose ratio with PBCT can be strongly favourable due to proton-boron reactions. The goal of this study was to investigate the theoretical maximum dose enhancement with PBCT, and its radiation physics mechanisms, through both macroscopic and microscopic simulations. The Monte Carlo simulations involved two stages due to model limitations. In the macroscopic calculations, variations in Bragg-peak height and depth caused by a 10 mm thick boron region were examined for different proton energies (62, 80 and 100 MeV) using MCNPX. Percentage depth doses and energy spectra of protons, alpha particles, neutrons and photons were also computed. The second stage used Geant4 for dose calculation in cellular dimensions. Macroscopically, the energy deposition peak shifted towards the surface by about 3.1 mm and the Bragg peak dose increased by up to 53.1%, 53.4%, 53.9% for proton energies 62, 80 and 100 MeV, respectively, compared to the proton-only (no boron) case. Microscopically, the calculated total equivalent doses for these energies showed 5%, 23% and 50% enhancements, respectively. The results suggest increased effectiveness with PBCT due to production of short-range alpha-particles, especially at higher incident proton energies. Also, by adding boron, the mean proton energy is reduced thereby increasing the probability of local dose deposition. The theoretical maximum possible improvement in dose equivalent in the target region at the cellular scale was shown to slightly increase with beam energy, reaching about 54% at the highest studied energy. This combined approach has provided further insight into the potential of PBCT. Further studies, particularly at higher proton energies, are indicated.

Neutron spectroscopy with TENIS using an artificial neural network

In this research, a ThErmal Neutron Imaging System (TENIS) consisting of two perpendicular sets of plastic scintillator arrays for boron neutron capture therapy (BNCT) application has been investigated in a completely different approach for neutron energy spectrum unfolding. TENIS provides a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from 1H(nth, γ)2D reactions, but in the present study, the 70-pixel thermal neutron images have been used as input data for unfolding the energy spectrum of incident neutrons. Having generated the thermal neutron images for 109 incident mono-energetic neutrons, a 70 × 109 response matrix has been generated using the MCNPX2.6 code for feeding into the artificial neural network tools of MATLAB. The errors of the final results for mono-energetic neutron sources are less than 10% and the root mean square error (RMSE) for the unfolded neutron spectrum of 252Cf is about 0.01. The agreement of the unfolding results for mono-energetic and 252Cf neutron sources confirms the performance of the TENIS system as a neutron spectrometer.

Study on neutron energy spectrum unfolding algorithm with EJ309 liquid scintillation detector

Proton recoil method can be used to experimentally measure fast neutron energy spectrum of non-pulsed neutron sources. The neutron energy spectrum unfolding algorithms based on the MLEM method, the GOLD deconvolution method, the Direct-D method, have been developed by using the EJ309 liquid scintillation detector. The degree of iteration by the mean square error (MSE) is proposed as a judgment criterion by according to the iterative accuracy, convergence speed and iteration efficiency. The developed neutron energy spectrum unfolding algorithms can unfolding the standard simulated mono-energetic neutron spectrum (2.5 MeV), 252Cf neutron spectrum, Am-Be neutron spectrum and the experimentally measured D-D neutron spectrum with higher precision as well as fewer iterations. The unfolded neutron spectra are in good agreement with the standard simulated neutron spectra and evaluated D-D neutron spectrum, which is revealed that the developed unfolding algorithms can unfolding neutron energy spectrum with reasonable accuracy.

Core design and neutronic analysis of a long-life LBE-cooled fast reactor NCLFR-Oil

The exploitation of heavy oil reservoirs calls for the implementation of thermal recovery techniques. To provide the heat source for oil recovery energy system, NCLFR-Oil, a Natural Circulation Lead-bismuth Fast Reactor was designed for heavy oil extraction, with a designed power of 40 MW and a lifetime of at least 15 years. In order to achieve the designed object for a reactor core that is durable, compact, transportable, and exhibits strong passive and active safety features, the selection of core materials such as fuel and absorber, evaluation of reactivity control system and corresponding neutronic performance have been optimized by iterative design in this paper with the reference of many small modular reactors. The depletion calculation confirms that the designed reactor can operate for 19 years without refueling. Neutronic performance were analyzed for multiplication factor within lifetime, isotopes conversion, neutron energy spectrum, radial and axial power distributions, nuclear fission reaction rates, reactivity feedback coefficients, and control rod worth with shutdown margins. The core analysis shows that the reactor has good power and safety characteristics, which can fully meet the energy demand of heavy oil exploitation.