Neutron Angular Distributions Research Articles

The Energy and Angular Distributions of Neutrons from (γ, N)-Reactions on γ-Beams Produced by Backward Compton Scattering at Eγ ≲ 40 MeV

The Energy and Angular Distributions of Neutrons from (γ, N)-Reactions on γ-Beams Produced by Backward Compton Scattering at Eγ ≲ 40 MeV

Development and Verification of Sampling Code NECP-SOUL for Evaluated Nuclear Data Files in ENDF-6 Format

A code called NECP-SOUL is developed to generate samples of evaluated nuclear data files in the ENDF-6 format. The motivation for the development is to obtain more accurate uncertainty quantification results based on the covariance data given in evaluated nuclear data files than the sampling method based on multigroup covariances. NECP-SOUL can handle all covariance data in ENDF-6 files, including average fission neutron multiplicities, resonance parameters, cross sections, angular distributions, energy distributions, and cross sections for the production of radioactive nuclides. In this work, samples of ENDF-6 files are input to the nuclear data processing code to generate A Compact ENDF (ACE)-formatted libraries, and these ACE-formatted libraries are used by the Monte Carlo code. Through statistical analysis, the uncertainty of the neutronics results can be obtained. NECP-SOUL is verified against another similar code, SANDY. In addition, the advantage of sampling ENDF-6 files is analyzed by comparing the uncertainty quantification results with those obtained by sampling multigroup covariance matrices. Moreover, the covariances of cross sections and angular distributions newly provided in ENDF/B-VIII.0 are used to analyze their effect on the uncertainty qualification results.

Investigation of breakup process in 9Be(α,n)12C reaction

Neutron energies and angular distributions were measured in 9Be(α,n)12C reaction for α energies of 5.5 and 6.5 MeV. Three major neutron groups were observed in the spectrum, which correspond to the ground and first two excited states of 12C. Measured data could only be explained by the TALYS calculation if reaction at more than one location within the target is considered for a given beam energy. The preferred locations are driven by the resonance energy levels existing in 13C. Neutron yield due to the 9Be breakup process was determined which is found to be 12.6 ± 0.2% and 18.4 ± 0.5% of the total reaction cross-section for 5.5 and 6.5 MeV respectively.

Preparation and properties of chromium protective coatings on lithium targets for accelerator-based neutron sources

Chromium (Cr) with its high antioxidative properties acts as a protective layer for lithium targets in accelerator-based boron neutron capture therapy (AB-BNCT). Employing a controlled variable method, this study explores the impact of magnetron sputtering parameters (sputtering power, argon flow, and sputtering pressure) on the surface morphology and crystalline structure of Cr films. Under the optimal preparation parameters (300W, 20ssm, 0.5Pa), a dense and uniform film with a (110) phase was formed, and a compact Cr2O3 layer developed during air exposure. A 1 μm Cr coating provided effective protection for approximately 2 h, and increased thickness further improved the protection. Monte Carlo simulations revealed that a 1 μm Cr coating caused around 1.83% energy loss and 5.54% neutron yield reduction for 2.8 MeV protons, meeting AB-BNCT requirements with minimal impact on neutron angular distribution. This study demonstrates that by investigating the effects of magnetron sputtering preparation parameters on the surface morphology and crystalline structure of Cr coatings, optimal preparation parameters can be selected. The fabrication of a uniform and dense 1 μm Cr protective layer on a lithium target effectively prevents the degradation of the lithium target during short-term exposure to air, thereby enhancing its transportability and ease of handling.

Design of proton-beam degrader for high-purity 89Zr production

This work investigated the most suitable type of degrader (Cu, Al or Nb) and its thickness, taking into consideration the salient aspects of concrete activation for high-purity 89Zr production by 89Y(p,n)89Zr nuclear reaction. The MCNP and FISPACT codes were used to determine the optimal degrader thickness and the radioactivity of shielding concrete by neutron activation, respectively. The results showed that the optimal thickness of the beam degraders was 1.16, 3.19, and 1.33 mm for Cu, Al, and Nb, respectively. The neutron production rate per proton and the energy and angular distributions of neutrons varied depending on the type of degrader. Considering the radioactivity of the target-room concrete and the amount of radioactive waste expected to be generated, the use of a 1.33-mm-thick Nb degrader for 89Zr production was determined to be the best choice.

Measurement and analysis of leakage neutron spectra from Lead slab samples with D–T neutrons

The leakage neutron spectra from three different sizes of Lead samples were measured by a TOF technique at 60° and 120°. The essential characteristic properties of the experimental measurement spectra can be reproduced well by MCNP code simulations with the ENDF/B-VIII.0, CENDL-3.2, JENDL-5.0, JEFF-3.3 and TENDL-2021 evaluated nuclear data libraries. The calculated results of JENDL-5.0 and JEFF-3.3 libraries agree better with the experimental data in the whole energy range. The results from ENDF/B-VIII.0 and CENDL-3.2 are overestimated in the 4–9 MeV range at 60° and in the 4–12.5 MeV range at 120°. The differences of the leakage neutron spectra by MCNP simulations using five evaluated nuclear data libraries mainly originate from the differences of the spectrum distributions of neutron reaction channels in these libraries. And the secondary neutron energy distribution and angular distribution from the five libraries have been present to explain it.

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.

Estimation of the yield and angular distribution of neutrons for calculating the biological shielding of reconstructed heavy ion accelerators with energies from 1 to 6 MeV/u

A simple analytical formulas for estimation of the yield and angular distribution of neutrons from thick targets would be useful for calculating the biological protection of heavy ion accelerators with energy (1–6) MeV/nucleon. The results of the calculation with this formulas were compared with those obtained using the LISE++ and FLUKA programs as well as with the available experimental data. As a result, it can be stated that the proposed calculation method can be used for quick assessment. The deviation of the results of such calculations from more accurate ones does not exceed a factor of two, which is comparable with the deviations in the calculation of biological shielding due to a variety in the protection properties of the materials used or in their thickness.

The Neutron Scattering Cross Section and Angular Distribution Measurement Program at LANL

Neutron scattering cross sections and angular distributions are leading terms in descriptions of neutron transport through any system. Despite the fundamental importance of nuclear data on these quantities, significant gaps in understanding and lack of experimental data persist in heavier elements and down to lighter structural materials, such as iron and aluminum. Recent measurements on carbon have also shown definitive proof that the neutron angular distribution can change with respect to the emission angle of γ-rays from inelastic scattering, thereby complicating γ-tagged measurements of inelastic neutron scattering. In this work, we describe the emerging program at Los Alamos National Laboratory for measurements of neutron scattering cross sections and neutron, γ-ray, and correlated n-γ angular distributions utilizing liquid scintillator detectors and the Correlated Gamma-Neutron Array for Scattering (CoGNAC) of CLYC scintillators. Currently, these detectors are operated simultaneously with the in-progress array of CLYC detectors in an inverted position, and data analysis techniques are being developed to span scattering measurements from light nuclei up through actinides. Preliminary results for measurements on 12C, 27Al, and 56Fe are presented here with a description of the analysis methods applied and anticipated capabilities of the full-scale detection system.

Neutron physics investigations of fundamental processes of statistical mechanics

Studies in spin dynamics of disordered media and multiple ultra-small angle neutron scattering are considered. The experiments were carried out on unique installations designed in ITEP laboratory of neutron physics: beta-NMR spectrometer and universal neutron diffractometer. The main attention is paid to random walks in disordered systems and ultra-small angle neutron scattering (USANS) on objects with space correlations in positions of scatterers. Synthesis of concentration expansion, semi-phenomenological theory and numerical simulations produced satisfactory description of the nuclear polarization transfer within disordered 8Li–6Li spin subsystem in LiF single crystal. The theory of USANS starts from eikonal approximation for the scattering amplitude, which (a) reproduces the phenomenological Moliere–Bethe theory for the observable neutron angular distributions for uncorrelated random positions of scatterers, and (b) gives a possibility to take into account their spatial correlations.

Probing yields and angular distributions of Be(d,n) neutrons with novel target of compressed beryllium-powder

A comprehensive study was conducted in the Cyclotron laboratory at National Research Tomsk Polytechnic University on the use of a new type of target utilized in the production of fast-neutrons. The flux and spectra of the fast-neutrons emitted by a novel target of compressed beryllium powder were studied. The neutron angular distribution was measured using Fe and Al foils as detectors in different angular situations of the target relative to the incident deuteron beam. It was found that there are some discrepancies in the neutron flux at 0° degree for the detectors, up to about 20% for different target-beam angle. Likewise, the flux of neutrons in the energy range between 1 and 6 MeV make up about 83% of the total flux. Compering with previous works using solid target, we found an increase of nearly twice the yield of fast neutrons. And there is a probability that increase would rise steadily as the deuteron energies exceed 12 MeV as was the case in our work. The results of this study are new and promising and we can obtain a relatively greater neutron yield when using a compressed powder target. This is reflected positively on many applications that use fast-neutron beams, such as radiotherapy, radioactive isotopes, and even nuclear fusion experiments, among others.

Erratum: Precision analysis of electron energy spectrum and angular distribution of neutron β− decay with polarized neutron and electron [Phys. Rev. C 95 , 055502 (2017)

Erratum: Precision analysis of electron energy spectrum and angular distribution of neutron <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML"><mml:msup><mml:mi>β</mml:mi><mml:mo>−</mml:mo></mml:msup></mml:math> decay with polarized neutron and electron [Phys. Rev. C <b>95</b> , 055502 (2017)

Development and validation of a model for assessing neutron fluence with lithium fluoride thermoluminescent detectors

Lithium fluoride thermoluminescent detectors are commonly used for neutrons measurements. However, the calibration is usually performed with readily available gamma sources in terms of gamma equivalent air kerma instead of the more meaningful neutron fluence. Neutron calibration coefficients in terms of neutron fluence depend strongly on the experimental conditions and can thus usually not be found in literature. Therefore, the goal of this work was to develop and validate a model to calculate such neutron calibration coefficients and to use this model to quantify the main sources of uncertainty. Comparison of the neutron calibration coefficients predicted by the model with experimental data from literature and this work showed reasonable agreement. However, a slight systematic deviation was found, probably caused by the significant uncertainty on the relative luminescence efficiency values predicted by the recently developed Microdosimetric d(z) Model for the low energy charged particles released during the 6Li(n,α)3H reaction. Therefore, it was decided to use relative luminescence efficiency values based on the experimental results instead. The model also identified the neutron energy and angular distribution and the luminescence reader light collection geometry as the main sources of uncertainty. Imprecise knowledge of these experimental input parameters can easily lead to uncertainties of the order of tens of percent on the calculated neutron calibration coefficients. The developed model is applicable for neutron energies below 0.01 MeV, but could be extended to higher energies in the future.

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.

Real-time nuclear activation detectors for measuring neutron angular distributions at the National Ignition Facility (invited).

The Real Time Nuclear Activation Detector (RTNAD) array at NIF measures the distribution of 14MeV neutrons emitted by deuterium-tritium (DT) fueled inertial confinement fusion implosions. The uniformity of the neutron distribution is an important indication of implosion symmetry and DT shell integrity. The array consists of 48 LaBr3(Ce) crystal gamma-ray spectrometers mounted outside the NIF target chamber, which continuously monitor the slow decay of the 909 keV gamma-ray line from activated 89Zr located in Zr cups surrounding each crystal. The measured decay rate dramatically increases during a DT implosion in proportion to the number of 14MeV neutrons striking each Zr cup. The neutrons produce activated 89Zr through an (n, 2n) reaction on 90Zr, which is insensitive to low energy neutrons. The neutron flux along the detector line-of-sight at shot time is determined by extrapolating the fitted 909 keV decay curve back to shot time. Automatic analysis algorithms were developed to handle the non-stop data stream. The large number of detectors and the high statistical accuracy of the array enable the spherical harmonic modes of the neutron angular distribution to be measured up to L ≤ 4 to provide a better understanding of implosion dynamics. In addition, these data combined with measurements of the down-scattered neutrons can be used to derive fuel areal density distributions. This paper will describe the RTNAD hardware and analysis procedures.

Benchmark experiment for bismuth by slab samples with D-T neutron source

A benchmark experiment is carried out with slab sample to validate nuclear data of bismuth by measurement of leakage neutron spectra from 0.8–16 MeV. The measured experiments are performed at China Institute of Atomic Energy with the integral benchmark facility established on the D–T neutron generator. The neutrons leaking from the sample are recorded at the measuring angle of 60° and 120° by time-of-flight technique. To make understanding of the impact of accuracy of data files in evaluated library on the neutron transport simulation, the calculation is carried out by MCNP-4C code with a series of latest available evaluated libraries. The calculated spectra are analyzed by comparing with experimental ones in neutron spectrum shape and C/E values. The result indicates that the theoretical simulation with different data libraries overestimated or underestimated the measured ones in some energy ranges, and the secondary neutron energy distribution and angular distribution in data files have been present to explain it.

Anisotropy in fission fragment and prompt neutron angular distributions

Several physics mechanisms can lead to the deviation from an isotropic angular distribution for both fission fragments and the neutrons that are emitted during the fission event. Two of these effects have recently been implemented into CGMF, the Monte Carlo fission event generator developed at Los Alamos National Laboratory: angular distribution sampling for fission fragments and pre-equilibrium neutrons (those emitted before the compound nucleus forms). Using these new developments, we show that the anisotropy of the neutrons reflects the anisotropy of the fission fragments, in particular as the outgoing energy of neutrons increases. Correlations between the fission fragment and neutron anisotropies could be used to extract the fission fragment anisotropy from the neutron angular distributions.

ON-THE-FLY INTERPOLATION OF CONTINUOUS TEMPERATURE-DEPENDENT THERMAL NEUTRON SCATTERING DATA IN RMC CODE

Thermal neutron scattering data have an important influence on the high-fidelity neutronics calculation of thermal reactors. Due to the limited storage capabilities of computers, a discrete ACE representation of the secondary neutron energy and angular distribution has been used for Monte Carlo calculation since the early 1980s. The use of this discrete representation does not produce noticeable effects in the integral calculations such as keff eigenvalues, but can produce noticeable deficiencies for differential calculations. A new continuous representation of the thermal neutron scattering data was created in 2006, but was not widely known. Recently, the continuous representation of the thermal neutron scattering ACE data based on ENDF/B-Ⅷ.0 library was officially released and was available for all users. The new representation shows great difference compared with the discrete one. In order to utilize the more physical and rigorous representation data for high fidelity neutronic-thermohydraulic coupling calculation, the on-the-fly treatment capability was proposed and implemented in RMC code. The two-dimensional linear-linear interpolation method was used to calculate the inelastic scattering cross sections and the secondary neutron energies and angles. The on-the-fly treatment capability was tested by a pressurized water reactor assembly. Results show that the on-the-fly treatment capability has high accuracy, and can be used to consider the temperature feedback in the neutronic-thermohydraulic coupling calculations. However, the efficiency of the on-the-fly treatment still need to be improved in the near future.

Improved MCNP Simulation Considering Neutron Angular Distribution and its Experimental Verification

Adeuterium-deuterium (DD) neutron generator has been simulated with MCNP 6.2. Two methods of simulating the emission of the neutrons created by the generator have been utilized: One with an isotropic emission of resultant neutrons and a second (anisotropic) accounting for the angular and energy distribution of resultant neutrons. The isotripic source was simulated using data generated by the DROSG-2000 program. A series of gold foils were irradiated at three different geometries around the neutron generator and the results compared to those produced from the two simulation methods.

Monte Carlo simulations of correlation between anisotropic emission characteristics of prompt fission neutrons and sample multiplication

In the present study, the correlation between prompt neutron angular distributions and sample multiplication for three spontaneously fissionable nuclides (252Cf, 244Cm, and 240Pu) was quantitatively determined by Monte Carlo simulations with MCNP6.2 and Geant4.10. To this end, various parameters (i.e., point, area, gradient, and net-area-based parameters) were investigated to analyze the asymmetry of the angular distribution. It was confirmed that the asymmetry decreases as leakage multiplication increases, and that the extent of this variation depends on the given nuclear isotopes. After comparing the sensitivity of the asymmetry values to leakage multiplication, the net-area-based parameter was determined to be the optimal asymmetry parameter. Finally, fitting equations were determined in order to determine, by regression analysis, leakage multiplication as a function of the asymmetry parameter for each isotope of interest. The results presented here could contribute to the development of a fast-neutron-based coincidence counting technique for nuclear material accountancy in safeguards applications.