Continuous Neutron Research Articles

Correlated n−γ angular distributions from the Q=4.4398 MeV C12(n,n′γ) reaction for incident neutron energies from 6.5 MeV to 16.5 MeV

Neutron scattering cross sections and angular distributions represent one of the most glaring sources of uncertainty in calculations of nuclear systems. Even simple nuclei like $^{12}\mathrm{C}$ show indications of errors in nuclear databases for scattering reactions. Measurements of inelastic neutron scattering have historically measured either the scattered neutrons or the nuclear deexcitation $\ensuremath{\gamma}$ emission. Only a very small number of experiments attempted correlated measurements of both the neutron and $\ensuremath{\gamma}$ data simultaneously, even though these $n\text{\ensuremath{-}}\ensuremath{\gamma}$ correlations could be essential for understanding particle transport in nuclear systems. In this work we describe a measurement of the $n, \ensuremath{\gamma}$, and correlated $n\text{\ensuremath{-}}\ensuremath{\gamma}$ angular distributions from the $Q=4.4398 \mathrm{MeV}$ $^{12}\mathrm{C}(n,{n}^{\ensuremath{'}}\ensuremath{\gamma})$ reaction in a single experiment using an EJ-309 liquid scintillator detector array with wide angular coverage, and with a continuous incident neutron energy range from 6.5 to 16.5 MeV. We also provide a thorough covariance description of these results, including normalization of the probability distribution. While the measured $n$ distributions agree well with the relatively large number of available literature measurements, there are comparatively very few measurements of the $\ensuremath{\gamma}$ distributions from this reaction. However, our data support the presence of a nonzero ${a}_{4}$ Legendre polynomial component of the $\ensuremath{\gamma}$ angular distribution suggested in past measurements, which is currently not incorporated in the ENDF/B-VIII.0 library despite the use of these same literature data for evaluation of the $^{12}\mathrm{C}(n,{n}^{\ensuremath{'}}\ensuremath{\gamma})$ cross section. The correlated $n\text{\ensuremath{-}}\ensuremath{\gamma}$ distribution measurements are limited to three measurements at incident neutron energies near 14 MeV. Our results do not generally agree with any of these literature measurements. We observe clear indications of significant changes in the $n$ distribution for specific $\ensuremath{\gamma}$-detection angles and vice versa especially near thresholds for other reaction channels, which shows the potential for significant bias in experiments that, for example, tag on inelastic scattering using a single or small number of $\ensuremath{\gamma}$-detection angles and could impact particle transport calculations.

Neural Network Based Deep Learning Method for Multi-Dimensional Neutron Diffusion Problems with Novel Treatment to Boundary

In this paper, the artificial neural networks (ANN) based deep learning (DL) techniques were developed to solve the neutron diffusion problems for the continuous neutron flux distribution without domain discretization in advance. Due to its mesh-free property, the DL solution can easily be extended to complicated geometries. Two specific realizations of DL methods with different boundary treatments are developed and compared for accuracy and efficiency, including the boundary independent method (BIM) and boundary dependent method (BDM). The performance comparison on analytic benchmark indicates BDM being the preferred DL method. Novel constructions of trial function are proposed to generalize the application of BDM. For a more in-depth understanding of the BDM on diffusion problems, the influence of important hyper-parameters is further investigated. Numerical results indicate that the accuracy of BDM can reach hundreds of times higher than that of BIM on diffusion problems. This work can provide a new perspective for applying the DL method to nuclear reactor calculations.

Impact of Be7 breakup on the Li7(p,n) neutron spectrum

The formation of continuum neutron distribution in $^{7}\mathrm{Li}(p,n)$ has been identified as due to the coupling of the $^{7}\mathrm{Be}$ breakup levels to the final state of the reaction. The continuum neutron spectra produced by the $^{7}\mathrm{Li}(p,n)$ reaction has been estimated by measuring the double differential cross sections for continuum and resonant breakup of $^{7}\mathrm{Be}$, through the $^{7}\mathrm{Li}(p,n)^{7}\mathrm{Be}^{*}$ reaction at 21 MeV of proton energy. The breakup contributions from continuum and $5/{2}^{\ensuremath{-}}$, $7/{2}^{\ensuremath{-}}$ states of $^{7}\mathrm{Be}$ have been identified. The measured double differential cross sections have been reproduced through continuum-discretized coupled channel-coupled reaction channel calculations. The cross sections were projected to neutron spectrum using the Monte Carlo approach and validated using experimentally measured $^{3}\mathrm{He}$ gated neutron spectra. $^{7}\mathrm{Li}(p,n)$ neutron spectrum at 20 MeV incident proton energy measured by McNaughton et al. [Nucl. Instrum. Methods 130, 555 (1975)] has been reproduced by adapting estimated model parameters for the reaction.

Space to Air High-Altitude Region Adjoint Neutron Transport

Neutrons from an atmospheric nuclear explosion can be detected by sensors in orbit. Current tools for characterizing the neutron energy spectrum assume a known source and use forward transport to recreate the detector response. In realistic scenarios the true source is unknown, making this an inefficient, iterative approach. In contrast, the adjoint approach directly solves for the source spectrum, enabling source reconstruction. The time–energy fluence at the satellite and adjoint transport equation allow a Monte Carlo method to characterize the neutron source’s energy spectrum directly in a new model: the Space to High-Altitude Region Adjoint (SAHARA) model. A new adjoint source event estimator was developed in SAHARA to find feasible solutions to the neutron transport problem given the constraints of the adjoint environment. This work explores SAHARA’s development and performance for mono-energetic and continuous neutron energy sources. In general, the identified spectra were shifted towards energies approximately 5% lower than the true source spectra, but SAHARA was able to capture the correct spectral shapes. Continuous energy sources, including real-world sources Fat Man and Little Boy, resulted in identifiable spectra that could have been produced by the same distribution as the true sources as demonstrated by two-dimensional (2D) Kolmogorov–Smirnov tests.

Errors introduced in fission neutron spectrum measurements using a single reference

Measurements of physical quantities are often made relative to a reference quantity, simplifying the interpretation of the measurement and its uncertainties. However, if applied without careful considerations, this technique can carry with it a significant and complex systematic error that is usually ignored. For example, measurements of continuous neutron spectra relative to a reference spectrum yield a result that does not properly include the important effects of neutron scattering in the experimental environment. These effects act to distort the measured spectrum shape and create notable errors in spectrum-integrated quantities. While this effect is one of a series of known potential sources of systematic error including, but not limited to target impurities, angular distributions, incident beam flux, and others, the error introduced from differences in environmental scattering effects between the reference and desired spectra are frequently overlooked. In this work, we demonstrate the origin of this error using simulated data corresponding to measurements of the neutron-induced prompt fission neutron spectra of 233U, 235U, 238U, and 239Pu as well as the spontaneous fission neutron spectrum of 252Cf. Errors in the spectrum shape can be as high as 10%–15% depending on the combination of incident and outgoing neutron energy. The spectrum integral and average spectrum energy also contain errors that change with incident neutron energy and can be as high as ∼2%. While this study is based on simulated data, we also show a method for proving that this error exists in an experimental data set and we show a correction that can be applied to results of this type to obtain a more accurate result, thereby largely removing systematic errors from environmental scattering effects. The potential for these kinds of errors should be considered in measurements producing continuous neutron spectra.

A Continuous Energy Neutron Transport Monte Carlo Simulator Project: Decomposition of the Neutron Energy Spectrum by Target Nuclei Tagging

In this work the latest developments a Monte Carlo simulator with continuous energy is reported. This simulator makes use of a sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum and the Maxwell-Boltzmann distribution. The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution, such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies. It is then shown how this procedure can emulate the up-scattering effect by the increase in the kinetic energy of the neutron population. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process.

Burnup computations of multi-pass fuel loading scenarios in HTR-10 using a pre-generated fuel composition library

This study aims to precisely investigate the core characteristics of HTR-10 using an online-refueling fuel loading scenario through an innovative approach that incorporates a pre-generated fuel composition library. HTR-10 is an experimental high-temperaturegas- cooled pebble-bed reactor constructed in China. Applying multi-pass fuel loading scenario, the pebbles in HTR-10 pass through the core several times via onlinerefueling. In this study, we proposed a model that simulates the dynamic operation of HTR-10 by using continuously static calculations. The model was constructed by dividing the core into 3 flow channels with different amounts of horizontal zones according to the residence time of fuel pebbles. However, random distribution of the pebbles with various burnup values complicates the refueling process. In order to simulate online-refueling more conveniently, a comprehensive method based on multipass fuel loading scenarios was thus developed. Firstly, a pre-generated fuel composition library containing the content of both actinides and fission products of fuel pebbles as a function of burnup, was built by implementing full core depletion calculations. During the refueling processes, a specific ratio of discharged fuel pebbles to fresh fuel pebbles was reloaded into the top zones. Thus, the average burnup values in the top zones can be resulted by considering the supplement of fresh fuel pebbles. Eventually, the fuel composition in the top zones were obtained by interpolating the average burnup value from the pre-generated library. In this study, all the calculations were performed using the MCNP6 computer code together with the ENDF/B-VII continuous energy neutron data library. The results revealed that the core reached a state of equilibrium after seven refueling processes with an increased rate of volume fraction. Our findings correspond to those of other studies. Another model constructed by locating the actual distribution of fresh pebbles also confirms the verification of the refueling method. Eventually, the burnup characteristics as well as the power and flux distributions in the HTR-10 core were analyzed using the simulation method.

Unfolding simulation of single-energy and continuous fast neutrons spectrum based on micro-pattern gas detector

This paper focuses on the feasibility of fast neutron energy spectrum measurement. The MCNPX and Geant4 are used to simulate two conversion models of stacking neutrons to protons in the triple GEM cathode coupled with multilayer polyethylene, with five kinds of single-energy neutron sources and Am-Be continuous neutron sources taken as research objects. The response function to 160 single energy neutrons and the recoil proton spectrum distribution of the above sources of the detection system are obtained by simulation. Using GRAVEL algorithm and MLEM algorithm and through simulation, the recoil proton spectra of six kinds of fast neutron sources are obtained, and they are further analyzed. The spectrum outcome is compared with the standard input spectrum, showing that they are in good agreement with each other. The relative uncertainty of the unfolding spectrum is around 10%–15%. In this part the relation of gas detector with the precision of unfolding spectrum is also discussed. The result shows that when the energy resolution of micro-pattern gas detection is better than 30%, the accuracy of fast neutron spectrum can meet the needs of practical applications. Furthermore, a new transformation model is proposed based on previous experiments and proves the feasibility of applying micro-pattern gas detector to fast neutron detection of simulation. Moreover, spectrum reconstruction can be achieved by using the obtained recoil proton spectrum combined with a suitable inversion algorithm. The modeling and spectrum analysis of this study can provide a different method of applying the fast neutron detection system composed of micro-pattern gas detectors to the detection of unknown fast neutron sources and also to the source recognition through spectrum reconstruction.

Deuteron inelastic scattering on 6Li and 7Li nuclei within the three-body cluster model * *The reported study was funded by RFBR, (20-02-00004)

The problem of the deuteron interaction with lithium nuclei, treated as a system of two coupled pointlike clusters, is formulated to calculate the cross sections of the d+Li reaction. The d+Li reaction mechanism is described using the Faddeev theory for the three-body problem of deuteron-nucleus interaction. This theory is slightly extended for calculation of the stripping processes 6Li(d,p)7Li, 7Li(d,p)8Li, 6Li(d,n)7Be, and 7Li(d,n)8Be, as well as fragmentation reactions yielding tritium, -particles, and continuous neutrons and protons in the initial deuteron kinetic-energy region MeV. The phase shifts found for Li and Li elastic scattering, as part of the simple optic model with a complex central potential, were used to find the cross sections for the 6Li and 7Li radiation captures. The three-body dynamics role is also summarized to demonstrate its significant influence within the Li system.

Fast neutrons at LNL Legnaro

In this contribution we describe NEPIR, the fast-neutron irradiation facility under construction at the 70 MeV cyclotron SPES facility of the INFN laboratory of Legnaro (LNL). NEPIR will be constructed in stages, according to the available funds. The initial configuration, based on a thick Be neutron production target, will be operational in 2022; it will be used for shielding studies against fast neutrons for space applications and to investigate neutron-induced Single Event Effects (SEE) in microelectronic devices and systems. In its final configuration NEPIR will have two target systems: one will deliver a Quasi Mono-energetic Neutron (QMN) beam, of general interest, with an adjustable energy peak in the 20–70 MeV range; the second target will deliver a specialized continuous energy neutron beam for studying the effects of fast neutrons produced in cosmic ray air-showers in electronic devices and systems. We review the use of NEPIR to characterize the sensitivity of electronics, describe the neutron production targets and the facility layout. In closing we describe ways, presently under investigation, to use the 15 MV XTU Tandem of LNL to produce nearly monochromatic fast neutrons that would complement the QMN system by allowing one to probe for SEE below 20 MeV.

Frame overlap Bragg edge imaging

Neutron Bragg edge imaging enables spatially resolved studies of crystalline features through the exploitation and analysis of Bragg edges in the transmission spectra recorded in each pixel of an imaging detector. Studies with high spectral resolutions, as is required e.g. for high-resolution strain mapping, and with large wavelength ranges have been largely reserved to pulsed neutron sources. This is due to the fact, that the efficiency for high wavelength resolution measurements is significantly higher at short pulse sources. At continuous sources a large fraction of the available neutrons must be sacrificed in order to achieve high wavelength resolution for a relevant bandwidth e.g. through a chopper system. Here we introduce a pulse overlap transmission imaging technique, which is suited to increase the available flux of high wavelength resolution time-of-flight neutron Bragg edge imaging at continuous neutron sources about an order of magnitude. Proof-of-principle measurements utilizing a chopper with a fourfold repeated random slit distribution of eight slits were performed at a thermal neutron beamline. It is demonstrated, that disentanglement of the overlapping pulses is achieved with the correlation theorem for signal processing. Thus, the Bragg edge pattern can be reconstructed from the strongly overlapping Bragg edge spectra recorded and the results demonstrate the feasibility of the technique.

Low counting rate measurement on thermal neutron induced fission using cross-correlation technique

The measurement procedure based on the continuous thermal neutron beam modulation with a mechanical chopper was developed for delayed neutron yield measurement of the thermal neutron induced fission of [Formula: see text]Np. The idea of the procedure is similar to that widely used in modern computer communications to prevent unauthorized data access. The data is modulated with a predefined pattern before transmission to the public network, and only the recipient that has the modulation pattern is able to demodulate it upon receipt. For thermal neutron induced reaction applications, the thermal neutron beam modulation pattern was used to demodulate the measured delayed neutron intensity signals on the detector output, resulting in nonzero output only for the detector signals correlated with the beam modulation. The comparison of the method with the conventional measurement procedure was provided, and it was demonstrated that the cross-correlation procedure has special features making it superior to the conventional one, especially when the measured value is extremely small in comparison with the background. Due to the strong sensitivity of the measurement procedure on the modulation pattern of the neutron beam, one can implement the modulation pattern of a specific shape to separate the effect of the thermal part of the beam from the higher energy part in the most confident way in the particular experiment. The remarkable property of our method is related to the unique possibility of separation of the effects caused exclusively by thermal neutrons using the neutron time-of-flight measurement available on the IBR-2 pulsed reactor.

Validation of Unresolved Neutron Resonance Parameters Using a Thick-Sample Transmission Measurement

Often discrepancies can be found in the corresponding cross sections of different evaluated nuclear data libraries. Traditional integral benchmarks that are used to validate such libraries are sensitive to cross-section values across many different energies. This means an erroneously low cross section at one energy may compensate for an erroneously high cross section at another energy, and the integral benchmark value may still be met. While the evaluated cross section may agree with that single benchmark, it could affect other systems differently. To reduce the potential for this error, an energy differential validation method is proposed herein for continuous energy Monte Carlo neutron transport models in the resolved resonance region and the unresolved resonance region (URR). The proposed method exposes the underlying physics of the URR and validates both the average cross section and resonance self-shielding effect driven by the fluctuations in that cross section. This is done by measuring the neutron transmission of a thick sample that, by its nature, exaggerates the resonance self-shielding effect. This validation method is shown to be very sensitive to the cross-section model used (resolved versus unresolved) and the fluctuation correction employed, allowing it to probe the validity of the previously mentioned cross-section evaluations. Tantalum-181 is used as an example to demonstrate the impact of different resonance evaluations. It was found that the JEFF-3.3 and JENDL-4.0u evaluations made reasonable choices for cross-section models of 181Ta; none of the current evaluations, however, can be used to properly model the validation transmission over all energies. It was also found that updating resonance parameters in the URR provided better agreement with the validation transmission.

Optimization of a magnetostatic cavity for a 3He spin analyzer on the CANDOR polychromatic reflectometer

A new instrument, the CANDOR (Chromatic Analysis Neutron Diffractometer or Reflectometer) polychromatic beam reflectometer is being built at the National Institute of Standards and Technology Center for Neutron Research. CANDOR will have a capability of polarization analysis for both specular and off-specular scattering, and allow for measuring the magnetic structures of magnetic thin films with a significantly shortened measurement time as compared to a conventional monochromatic beam reflectometer at a continuous neutron source. A critical component of the polarization analysis capability is to develop a 3He neutron spin filter (NSF) for polarization analysis for both specular and off-specular scattering for a continuous wavelength range between 4 Å and 6 Å and a large solid angle coverage for up to 20 energy dispersive detector channels. For ex-situ spin-exchange optical pumping (SEOP) operation, a magnetically shielded solenoid (MSS) will be employed to provide a volume averaged field gradient, better than 5×10−4 cm−1 to maintain 3He polarization relaxation times of at least a few hundred hours on the beam line even when there are significant stray fields from a magnet operating at a field up to 3 T at the sample position. We describe a procedure to design, construct, and optimize a MSS with non-identical compensation coil dimensions on the ends to match the scattered neutron beam rays. The volume averaged field gradient is quantified and optimized over a cell for a MSS using the finite-element software package RADIA and compared with that from the experiment.

A Hierarchical Bayesian Approach to Neutron Spectrum Unfolding With Organic Scintillators

We propose a hierarchical Bayesian model and a state-of-the-art Monte Carlo sampling method to solve the unfolding problem, i.e., to estimate the spectrum of an unknown neutron source from the data detected by an organic scintillator. Inferring neutron spectra is important for several applications, including nonproliferation and nuclear security, as it allows the discrimination of fission sources in special nuclear material (SNM) from other types of neutron sources based on the differences of the emitted neutron spectra. Organic scintillators interact with neutrons mostly via elastic scattering on hydrogen nuclei and therefore partially retain neutron energy information. Consequently, the neutron spectrum can be derived through deconvolution of the measured light-output spectrum and the response functions of the scintillator to monoenergetic neutrons. The proposed approach is compared to three existing methods using the simulated data to enable controlled benchmarks. We consider three sets of detector responses. One set corresponds to a 2.5-MeV monoenergetic neutron source and two sets are associated with (energywise) continuous neutron sources ( <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">252</sup> Cf and <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">241</sup> AmBe). Our results show that the proposed method has similar or better unfolding performance compared with other iterative or Tikhonov regularization-based approaches in terms of accuracy and robustness against limited detection events while requiring less user supervision. The proposed method also provides a posteriori confidence measures, which offers additional information regarding the uncertainty of the measurements and the extracted information.

Fluorine determination in biological and environmental samples with INAA using fast neutrons from a p(19 MeV) + Be neutron generator

We present a new activation method based on the 19F(n,2n)18F reaction using fast neutrons produced by a p(19 MeV) + Be accelerator-driven fast neutron source providing continuous neutron spectrum up to 17 MeV with a neutron fluence rate of approximately 9 × 1010 cm−2 s−1 at a 12-µA proton beam current. We describe elimination of interferences in measurement of 18F, a pure positron emitter with T1/2 = 1.83 h. We present results of fluorine determination for several biological and environmental reference materials with the new procedure and compare them with those achieved by other methods in terms of limits of detection, accuracy and precision, where available.

Conceptual study on a novel method for detecting nuclear material using a neutron source

This paper presents a conceptual study of a novel active method for detecting nuclear material by using a continuous neutron source. When a container housing nuclear material is irradiated with neutrons, the measurement data recorded by neutron detectors includes both neutrons from the neutron source and those from induced fission reactions. For detecting nuclear material, it is necessary to distinguish the fission neutron component from the source neutron component. The uniqueness of the new method is that it varies the effective intensity of a continuous neutron source by moving it rapidly near the container. As a result of the variation on the effective source intensity, the delay between the moment the source is geometrically closest to the sample and the peaking time of the neutron count rate is detected. This delay is caused mainly by the fission neutron component, that is, the diffusion time of thermalized neutrons, which induce fission reactions in the container, which enables us to detect nuclear material from the measurement data. Because this novel method does not require a D-T tube, which is considerably more expensive than a continuous neutron source, the new system is expected to be affordable and easy to assemble. Moreover, this system can be made so compact that it is suitable for use in mobile applications.

Monte Carlo characterization and benchmarking of extended range REM meters for its application in shielding and radiation area monitoring in Compact Proton Therapy Centers (CPTC)

Compact Proton Therapy Centers, CPTC, have a single treatment room, and are technologically more affordable, smaller, advanced and easier to use. From a radiological protection point of view, the leading concern in CPTC are interactions of protons with components of the facility and patients that yield a broad emission of secondary particles, mainly high-energy neutrons, up to 230 MeV, and photons. Optimal design of shielding involves theoretical assumptions in the design phase and, consequently, experimental measurements with extended range neutron detectors must be carried out in the facility during the commissioning period to verify the design, assumptions and building of the enclosures. There are almost 50 CPTC under construction and planning around the world, hence the improvement of methodologies to verify the shielding and to evaluate the dose to workers and general public in CPTC is a trending issue.The aim of this work was to evaluate and compare the response of two commercial extended range REM meters, WENDI-II and LUPIN-II, for their application in shielding verification and radiation area monitoring in CPTC facilities, by estimating the ambient dose equivalent, H*(10), through the Monte Carlo code MCNP6. The results have been compared with previous works. Likewise, the performance evaluation of these devices in continuous energy neutron field have been carried out, using the AmBe/241 neutron source of the Neutronics Hall (NH) of the Neutron Measurements Laboratory of the Energy Engineering Department of Universidad Politecnica de Madrid (LMN-UPM), through Monte Carlo simulation with the MCNP6 code and experimental measurements. The work is framed into the project Contributions to Shielding and Dosimetry of Neutrons in CPTC.

A study of burnup credit in criticality safety analysis for PBR spent fuel pebbles

This study attempts to investigate the influence due to the use of burnup credit in the criticality safety analysis for pebble-bed reactor (PBR) spent fuel pebbles. Recently, the development of PBR is very quick in China, thus the storage of spent fuel pebbles will become a significant issue in the foreseeable future. For the safe storage, the criticality safety analysis is of vital important. Furthermore, the utilization of burnup credit in the criticality safety analysis is essential due to the fact that it can result in a more compact design of storage system. In this study, all the calculations were performed using MCNP6 associated with the continuous energy neutron data library ENDF/B-VII. In addition, two geometrical models which represent the HTR-10 core and a proposed storage cask were adopted. The proposed storage cask model was established based on the real storage cask for HTR-10 spent fuel pebbles, but the capacity was increased in order to evaluate the impact of burnup credit. Moreover, the most conservative condition for this proposed cask occurs when the cask is full of water with a water density around 0.35 g/cm3. For burnup credit calculations, three operating parameters related to HTR-10 were investigated, including the usage of control rod, the fuel temperature, and the volume fraction of fuel pebbles in total pebbles. Additionally, these operating parameters were combined and classified as three single effects and four compound effects. The single and compound effects were defined as the influence on the effective multiplication factor (keff) due to simultaneous variations of one and multiple operating parameters, respectively. It is worth noting that in most of the compound effects, the reactivity deviation (or the change of keff, Δk) resulting from the compound effect was not a summation of Δk’s resulting from the associated single effects. This phenomena may affect the precise assessment to some extent. Finally, the mechanisms of both the single and compound effects were explored in depth from the analysis of the spectral distribution of fission. From this analysis, a harder spectral distribution of fission resulting from the single and compound effects corresponds a larger value of Δk.

On the slowing-down of 14 MeV fusion neutrons: A spectrometry benchmark and perspectives on future neutron science facilities

The spectral fluence rate of the thermalized neutron field obtained by slowing down the 14 MeV fusion neutrons produced at the accelerator-driven Frascati Neutron Generator is measured by means of the Bonner Sphere Spectrometer. Neutron thermalization is achieved by means of a moderator assembly made of a copper pre-moderator and a polyethylene moderator. A Monte Carlo simulation reproducing the experimental set-up is also performed by means of the MCNP code and the results are compared to the experimental data. The benchmarked Monte Carlo is then used to predict the brilliance of a thermal moderator operating at a potential high-intensity 14 MeV neutron continuous source featuring a neutron emission rate of . The results obtained show that if a D-T neutron source featuring a continuous neutron emission rate of could be made operative, it could be effectively exploited for neutron science.