Incident Neutron Research Articles

Neutron calorimeter designfor pulsed neutron experiment diagnostics.

Neutron calorimeter for diagnostic support of both fission- and fusion-based pulsed neutron experiments is described. The calorimeter generates a voltage signal output via a pair of thermocouples contacting a sample material, which experiences a temperature increase due to interactions with an incident neutron fluence. Advantages of the calorimeter include the versatility derived from its small size, modularity, and the variety of employable neutron absorber materials and sizes. These characteristics permit use with neutron energies between the thermal and fast (up to 14MeV) spectra. Designoverview and requirements, Monte Carlo particle transport simulation results, test data, and possibilities for calorimeter development are provided and discussed.

Monte Carlo modeling of production of three alpha-particles in low energy n +12C interactions

To describe [Formula: see text] interactions with production of three [Formula: see text]-particles at incident neutron kinetic energy of 14 MeV in a nuclear (photo) emulsion, a Monte Carlo model is proposed for four channels of decay of an excited carbon-12 nucleus into three [Formula: see text]-particles. The Monte Carlo calculation results describe well the experimental data on the distribution of the angle between the three-dimensional momenta of all pairs of [Formula: see text]-particles in a collision event, on the distribution of the angle between the projections of the momentum vectors of all pairs of [Formula: see text]-particles in collision event on each of the coordinate planes, on the distribution of the sum of the kinetic energies of all pairs of [Formula: see text]-particles in a collision event, and the distribution of projections of the momenta of [Formula: see text]-particles on the coordinate planes. The best agreement of the Monte Carlo model results with the experimental data is achieved if the direct decay [Formula: see text] and decay through the formation of an intermediate beryllium nucleus [Formula: see text] are generated with equal probabilities, while the excitation energies of 3.04 MeV, 1.04 MeV, and 0.1 MeV for the beryllium nucleus are generated with relative weights of 75%, 15%, and 10%, respectively.

Neutron-Induced Nuclear Cross-Sections Study for Plasma Facing Materials via Machine Learning: Molybdenum Isotopes

In this work, we apply a machine learning algorithm to the regression analysis of the nuclear cross-section of neutron-induced nuclear reactions of molybdenum isotopes, 92Mo at incident neutron energy around 14 MeV. The machine learning algorithms used in this work are the Random Forest (RF), Gaussian Process Regression (GPR), and Support Vector Machine (SVM). The performance of each algorithm is determined and compared by evaluating the root mean square error (RMSE) and the correlation coefficient (R2). We demonstrate that machine learning can produce a better regression curve of the nuclear cross-section for the neutron-induced nuclear reaction of 92Mo isotopes compared to the simulation results using EMPIRE 3.2 and TALYS 1.9 from the previous literature. From our study, GPR is found to be better compared to RF and SVM algorithms, with R2=1 and RMSE =0.33557. We also employed the crude estimation of property (CEP) as inputs, which consist of simulation nuclear cross-section from TALYS 1.9 and EMPIRE 3.2 nuclear code alongside the experimental data obtained from EXFOR (1 April 2021). Although the Experimental only (EXP) dataset generates a more accurate cross-section, the use of CEP-only data is found to generate an accurate enough regression curve which indicates a potential use in training machine learning models for the nuclear reaction that is unavailable in EXFOR.

Measurements of the neutron capture cross section of 243Am around 23.5 keV

ABSTRACT The neutron capture cross section of 243Am was measured by the time-of-flight method with a pulsed neutron beam from a spallation neutron source of the Japan Proton Accelerator Research Complex. An Fe neutron beam filter was used to make the incident neutron beam quasi-monoenergetic around 23.5 keV. The neutron capture γ-rays were detected with a NaI(Tl) detector. The pulse height weighting technique was employed to derive the neutron capture cross section from the pulse height spectrum. The cross section was determined relative to the capture cross section of 197Au of JENDL-4.0. The obtained neutron capture cross section of 243Am was 2.52 ± 0.14 b, determined with a much smaller uncertainty than those of past measurements.

Temperature-dependent dynamic structure factors for liquid water inferred from inelastic neutron scattering measurements.

Temperature-dependent dynamic structure factors S(Q, ω) for liquid water have been calculated using a composite model, which is based on the decoupling approximation of the mean square displacement of the water molecules into diffusion and solid-like vibrational parts. The solid-like vibrational part Svib(Q, ω) is calculated with the phonon expansion method established in the framework of the incoherent Gaussian approximation. The diffusion part Sdiff(Q, ω) relies on the Egelstaff-Schofield translational diffusion model corrected for jump diffusions and rotational diffusions with the Singwi-Sjölander random model and Sears expansion, respectively. Systematics of the model parameters as a function of temperature were deduced from quasi-elastic neutron scattering data analysis reported in the literature and from molecular dynamics (MD) simulations relying on the TIP4P/2005f model. The resulting S(Q, ω) values are confronted by means of Monte Carlo simulations to inelastic neutron scattering data measured with IN4, IN5, and IN6 time-of-flight spectrometers of the Institut Laue-Langevin (ILL) (Grenoble, France). A modest range of temperatures (283-494K) has been investigated with neutron wavelengths corresponding to incident neutron energies ranging from 0.57 to 67.6 meV. The neutron-weighted multiphonon spectra deduced from the ILL data indicate a slight overestimation by the MD simulations of the frequency shift and broadening of the librational band. The descriptive power of the composite model was suited for improving the comparison to experiments via Bayesian updating of prior model parameters inferred from MD simulations. The reported posterior temperature-dependent densities of state of hydrogen in H2O would represent valuable insights for studying the collective coupling interactions in the water molecule between the inter- and intramolecular degrees of freedom.

Estimation of fission yield and cross-section for thorium-based molten salt reactor and accelerator driven system

Thorium is the primary candidate fuel in Thorium-based Molten Salt Reactor (TMSR) and Accelerator Driven System (ADS). However, to utilize Thorium in the reactor and accelerator systems, nuclear data in the literature are not sufficient for neutron-induced fission reaction process. Therefore, we estimated neutron fission cross-section [Formula: see text] (mb), independent [Formula: see text] and primary fission fragment yield [Formula: see text], mass-dependent average neutron multiplicity [Formula: see text], energy-dependent average neutron multiplicity [Formula: see text] and neutron emission multiplicity distribution [Formula: see text] of [Formula: see text] for different neutron incident energy values, 1 MeV, 1.5 MeV, 3 MeV, 5.5 MeV, 7.7 MeV, 10 MeV, 14.8 MeV and 20 MeV, where the energy region is important for the reactor and accelerator systems used in the industry. The obtained results are compared with experimental data obtained from the literature and are discussed for each energy value.

Cross Sections for Neutron-Production from 6- and 10-MeV Neutrons Incident on 10B and 11B

Using time-of-flight techniques, we have measured neutron emission spectra from 6- and 10-MeV incident neutron energies on 10B and 11B for laboratory angles between 20 and 145 deg using the unique neutron source reaction 1H(t,n). Double-differential cross sections and their integrated values have been extracted for elastic and inelastic processes, and their integrated values are summarized. In cases where discrete scattering peaks could be analyzed, differential cross-section angular distributions have been extracted and are compared to other results and to evaluated nuclear data.

Unfolding neutron spectrum using Doppler broadened γ peak shapes

The Doppler broadening of gamma rays occurs in the fast neutron analysis with photons emitted in inelastic scattering reactions on certain nuclei, such as 12C and 14N. In this paper, we developed an algorithm based on the Monte Carlo method to calculate the Doppler broadened gamma peak shape of the 4.438 MeV gamma ray emitted by 12C in graphite when exposed to fast neutrons of different energies. The SAND-II unfolding iterative method is applied to obtain the spectrum of the incident neutron beam by using the Doppler broadened gamma spectra. In the calculation example, the unfolded neutron spectrum shows reasonable agreements with the original incident neutron spectrum.

Semi-empirical systematics formulas for the (n, p), (n, α), and (n, 2n) reaction cross-sections at 14.5 MeV

New Semi-empirical mass formulas were developed for the (n, p), (n, α), and (n, 2n) reaction cross-sections at 14.5 MeV neutron energy by using the literature data available on EXFOR library. The new formulas were derived on the basis of the statistical model considering Q − value dependence. The results obtained from the present exercise were compared with other systematic formulas developed in past years as well as with the experimental literature data reported on EXFOR library. The comparison has been done on the basis of statistical model as well as on the asymmetry parameter dependence. The present work offers an alternative to reproduce nuclear data at 14.5 MeV incident neutron energies.

C13(n,2nγ)C12 γ -ray production in the 14–16 MeV incident neutron energy range

$\ensuremath{\gamma}$-ray emission from $^{12}\mathrm{C}$ and $^{13}\mathrm{C}$ samples irradiated with deuterium-tritium fusion neutrons was experimentally measured at the Omega Laser Facility and at the Ohio University Edwards Accelerator Laboratory. The intent of these measurements was to determine the feasibility of using $^{13}\mathrm{C}$-based plastic ablators with embedded $^{12}\mathrm{C}$ layers for ``dark mix'' diagnosis of inertial confinement fusion implosions. Spectrally resolved measurements at Ohio University identified significant 4.44-MeV $\ensuremath{\gamma}$-ray emission from the $^{13}\mathrm{C}(n,2n\ensuremath{\gamma})^{12}\mathrm{C}$-L1 reaction channel. The recorded 4.44-MeV $^{13}\mathrm{C}$ signal was compared against emission from an identically irradiated $^{12}\mathrm{C}$ target with known $^{12}\mathrm{C}$($n,{n}^{\ensuremath{'}}\ensuremath{\gamma})^{12}\mathrm{C}$-L1 cross section, which resulted in an average $^{13}\mathrm{C}(n,2n\ensuremath{\gamma})^{12}\mathrm{C}$-L1 cross section of 117 $\ifmmode\pm\else\textpm\fi{}$ 17 mb over the incident neutron energy distribution range from 14.4 to 15.8 MeV. Integrated $^{13}\mathrm{C}$ $\ensuremath{\gamma}$-ray signals above 2.9 MeV recorded with the Gas Cherenkov Detector at Omega exceeded MCNP6.1 predictions by a factor of 3. The additional signal was attributed to 4.44-MeV $\ensuremath{\gamma}\mathrm{s}$ resulting in an inferred $^{13}\mathrm{C}(n,2n\ensuremath{\gamma})^{12}\mathrm{C}$-L1 cross section of $95\ifmmode\pm\else\textpm\fi{}11$ mb at 14.1-MeV average incident neutron energy. As a result, the $^{13}\mathrm{C}$-based ``dark mix'' diagnostic concept was deemed infeasible.

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.

New design of a magnetic device for wide-angle XYZ polarization analysis PASTIS-3, from the concept to first tests with thermal neutrons

We present the new concept and neutron test results of the third generation PASTIS device for XYZ wide-angle polarization analysis for thermal neutrons. PASTIS-3 uses two polarized 3He Neutron Spin filter Cells (NSF) for polarizing and analyzing neutrons spins, flipping the incident neutron beam polarization by reversing the 3He spin state of the polarizing entrance cell. We detail a new coil design for high spatial homogeneity of the magnetic field in any field direction, and the calculation method employed to optimize the geometry and to characterize the field gradients. Various components of the apparatus are described with particular attention to their performance and impact on the relaxation time of the NSF cells. Finally, we present test results of the new device tested on the thermal triple-axis spectrometer IN20 at the ILL.

Neutron interferometry and tests of short-range modifications of gravity

We consider tests of short-distance modifications of gravity based on neutron interferometry in the scenario of large extra dimensions. Avoiding the non-computability problem in the calculation of the internal gravitational potential of extended sources, typical of models with zero-width brane, we determine the neutron optical potential associated with the higher-dimension gravitational interaction between the incident neutron and a material medium in the context of thick brane theories. Proceeding this way, we identify the physical quantity of the extra dimension model that the neutron interferometry is capable of constraining. We also consider interferometric experiments in which the phase shifter is an electric field, as in the test of the Aharanov-Casher effect. We argue that this experiment, with this non-baryonic source, can be viewed as a test of the short-range behavior of Post-Newtonian parameters that measure the capacity of the pressure and the internal energy for producing gravity.

Suppression of spurious neutrons for the neutron detector used in small angle neutron scattering

Small Angle Neutron Scattering (SANS) measures the angular distribution of neutrons scattered by the sample to calculate the momentum transfer in the scattering process. However, the incident neutrons may firstly undergo unwanted collisions with the detector’s structural material before they are registered. The detection of such spurious neutrons might lead to a biased interpretation to the microstructure of the inspected sample. A newly-designed boron-lined honeycomb converter based gaseous neutron detector is introduced in this study to provide a capability for suppressing the spurious neutrons. The simulation and experimental results indicate that a 3 to 10 times improvement for the spurious neutron suppression capability could be achieved.

Pu239(n,2n)Pu238cross section measurement using a recoil method

A new measurement of the $^{239}\mathrm{Pu}(n,2n)$ $^{238}\mathrm{Pu}$ cross section was performed at 7.1, 7.7, and 9.3-MeV incident neutron energy using an activation technique to help resolve inconsistencies in the existing database above the threshold of the reaction. The method is based on the collection of $^{238}\mathrm{Pu}$ nuclei produced in the $(n,2n)$ reaction. Although never used before to obtain a neutron cross section, this recoil technics overcomes the difficulty of having deposits of $^{239}\mathrm{Pu}$ of extremely high purity. The collection efficiency must be determined using previously published data at higher energy, and our measurements are relative to these points. The neutrons were produced via the $^{2}\mathrm{H}(d,n)$ $^{3}\mathrm{He}$ reaction. The incident neutron flux was determined via the $^{27}\mathrm{Al}$($n$,\ensuremath{\alpha}) $^{24}\mathrm{Na}$ standard reaction. The present measurements are consistent with the previous data and disagree with the jeff3.3 evaluation.

Response of TLD-600/TLD-700 and CR-39 to neutrons for medical dosimetry.

In this study, thermoluminescence detectors (TLD)-600 and TLD-700 were used under different conditions to study neutron dosimetry for its application in medical dosimetry as albedo dosimeters. 6 LiF has a high cross-section for neutrons and is more sensitive to neutrons than to gamma radiation. Conversely, TLD-700 showed a better response to gamma radiation. Therefore, to obtain the response for neutrons, the responses for TLD-600 and TLD-700 were subtracted. A cadmium sheet was used to absorb incident thermal neutrons so that detector measured only backscattered neutrons from the albedo dosimeter. A Perspex sheet was used as a moderator to thermalize the fast neutron source Am-Be used in the experiment. CR-39 was used to detect fast neutrons and act as an albedo dosimeter when covered with a 10 B disc. Both TLD and CR-39 dosimeters were successful in detecting scattered neutrons in radiotherapy rooms.

Development of lightweight and highly efficient fast neutrons composites shields based on epoxy, UHMWPE fibres and boron carbide particles

In a vision to design wearable and lightweight protective thermal neutrons shields, a new hybrid composite was developed from Ultra High Molecular Weight Polyethylene (UHMWPE) fibres, epoxy and boron carbide (B4C) particles. The hybrid constituents were selected in a way to maximize both the shielding performances and the overall mechanical behaviour. Indeed, two first constituents of the hybrid attenuate the incident neutrons through scattering, while the B4C particles mainly interact with neutrons by absorption. More importantly, the UHMWPE fibres, as one of the strongest polymeric fibres, were also chosen to sustain the mechanical loads and provide the requested flexibility. The measurements were performed using an optimized experimental setup at NUR research reactor, Algiers. The effect of the amount of the B4C particles was also studied to maximize the shielding ability. The obtained results confirmed the high shielding efficiency of all the developed materials. Meanwhile, the best performance was recorded at B4C amount of 10 wt% with a macroscopic cross-section (Σ) of 0.188 cm−1 equivalent to a mean free path (λ) of 3.7 cm. In general, the developed materials performed better than the traditional neutrons shields and can be seen as promising materials for the intended use.

Multilayered Solid-State Neutron Sensor

Recent advances in semiconductor processing technology have enabled development of a solid-state neutron sensor with exceptional efficiency for its unique reduced size. The boron-10 ( n, alpha) capture reaction is used to detect incident thermal neutrons. This reaction offers significant advantages for a small detector because the reaction products have high energies which can generate a large number of charge carriers in semiconductor materials by impact ionization. The enabling and differentiating technology of the sensor design is a stack consisting of multiple bilayers alternating between converter material (boron) and collector semiconductor material (silicon). As the number of these bilayers increases, the probability of a neutron interacting with boron-10 in at least one of the bilayers-so it is captured in the detector before it has a chance to pass through-approaches unity. The sensor can be made very thin while also remaining highly efficient, offering a unique possibility for low-profile in situ dosimetry. Calculations have shown that a multilayer sensor with an efficiency greater than 70% is feasible in a sensor stack 2 mm thick.

Application of VERDI detectors for neutron fluence measurements during the JET 2019 Deuterium-Deuterium campaign

The Novel Neutron Detector for Fusion (VERDI) project aims to develop a passive detector which will provide a robust approach for neutron detection in current and future fusion devices. The detector comprises a low activation capsule, capable to withstand the harsh fusion conditions, containing defined added metallic elements, the activation of which provides neutron fluence and energy spectrum information. In the present work, VERDI detectors were employed for the first time to determine neutron fluence and energy spectra at the Joint European Torus (JET) Long Term Irradiation Station (LTIS) during the 2019 Deuterium-Deuterium (DD) campaign. Two materials, graphite and carbon fibre reinforced silicon carbide composite, were used for the capsule fabrication containing disc-shaped activation foils (Ag, Mn, Nb, Ni, Rh, Ti, Y). Post-irradiation gamma-spectrometry measurements were performed in order to identify the product isotopes and derive the incident neutron fluence. The results were fed into the MAXED code to reconstruct the neutron spectrum. The derived spectrum was in very good agreement with the calculated one using the MCNP code. A difference of 8% was found between the derived and the experimentally determined total integrated neutron flux. This work demonstrates the feasibility of employing VERDI detectors for neutron measurements in real fusion environments.

Application of an energy-dependent instrument response function to analysis of nTOF data from cryogenic DT experiments

Neutron time-of-flight (nTOF) detectors are used to diagnose the conditions present in inertial confinement fusion (ICF) experiments and basic laboratory physics experiments performed on an ICF platform. The instrument response function (IRF) of these detectors is constructed by convolution of two components: an x-ray IRF and a neutron interaction response. The shape of the neutron interaction response varies with incident neutron energy, changing the shape of the total IRF. Analyses of nTOF data that span a broad range of energies must account for this energy-dependence in order to accurately infer plasma parameters and nuclear properties in ICF experiments. This work briefly reviews a matrix multiplication approach to convolution, which allows for an energy-dependent change in the shape of the IRF. This method is applied to synthetic data resembling symmetric cryogenic DT implosions to examine the effect of the energy-dependent IRF on the inferred areal density. The results of forward fits that infer ion temperatures and areal densities from nTOF data collected during cryogenic DT experiments on OMEGA are also discussed.