Neutron Cross Sections Research Articles

Total neutron cross-section measurement on CH with a novel 3D-projection scintillator detector

In order to extract neutrino oscillation parameters, long-baseline neutrino oscillation experiments rely on detailed models of neutrino interactions with nuclei. These models constitute an important source of systematic uncertainty, partially because detectors to date have been blind to final state neutrons. Three-dimensional projection scintillator trackers comprise components of the near detectors of the next generation long-baseline neutrino experiments. Due to the good timing resolution and fine granularity, this technology is capable of measuring neutron kinetic energy in neutrino interactions on an event-by-event basis and will provide valuable data for refining neutrino interaction models and ways to reconstruct neutrino energy. Two prototypes have been exposed to the neutron beamline at Los Alamos National Laboratory (LANL) in both 2019 and 2020, with neutron energies between 0 and 800 MeV. In order to demonstrate the capability of neutron detection, the total neutron-scintillator cross section as a function of neutron energy is measured and compared to external measurements. The measured total neutron cross section in scintillator between 98 and 688 MeV is 0.36 ± 0.05 barn.

Characterization of thermal neutron distribution of an Am–Be neutron source setup by CdZnTe detector

Owing to the developments in prompt gamma neutron activation analysis (PGNAA) and prompt gamma ray activation imaging (PGAI), it is necessary to develop an online thermal neutron distribution measurement method. The CdZnTe detector is considered as an alternative thermal neutron detector because of its high thermal neutron capture cross section. In this study, the thermal neutron field of an 241Am–Be neutron source was determined by CdZnTe detector. The intrinsic neutron detection efficiency of CdZnTe detector was calculated by using indium foil activation, and the value was 3.65%. Then the characteristics of neutron source were conducted with the calibrated CdZnTe detector. The thermal neutron fluxes in front of beam port were measured at several distances ranging from 0 cm up to 28 cm. Thermal neutron field at distances of 1 cm and 5 cm were also measured. The experimental data were then compared with Monte Carlo simulation. The results demonstrated that the simulated data agree well with experimental measurements.

General displacement function for displacement damage cross-section calculation

The estimation of displacement damage rate requires the prior knowledge of displacement damage cross-section and displacement function is one part in the displacement damage cross-section calculation. The NRT model, which is used to calculate the displacement function for many years, always overestimates the number of displaced atoms in material. Although the arc-dpa model proposed recently can give correct results for the materials irradiated by neutron, there is still no suitable displacement function that can be used to accurately obtain both neutron and charged particles displacement cross-section comparing with experimental data. In the present work, a general displacement function is proposed, which is composed by power function model, arc-dpa model and BCA model. The model parameters are obtained for aluminum, iron, copper and tungsten. The displacement damage cross-sections for the materials irradiated by neutron and charged particles calculated using the general displacement function agree well with available experimental data.

A systematic analysis of deuteron breakup

Inclusive breakup cross sections for deuterons incident on a wide range of nuclei at different incident energies are studied. In our analysis, we use the post-form distorted-wave-Born approximation (DWBA) to calculate the elastic and nonelastic breakup cross sections. We also calculate exciton and statistical compound-nucleus model cross sections to account for pre-equilibrium and equilibrium emissions. The calculations are performed using the code EMPIRE and are compared to the available experimental data. The comparisons of the integrated cross sections reveal a generally good agreement between the calculations and the data. However, systematic discrepancies are observed between the calculated double differential neutron and proton cross sections and the data, especially for heavier targets. Possible reasons for the disagreements between the theoretical predictions and the experimental data are discussed.

Statistical (n,gamma ) cross section model comparison for short-lived nuclei

Neutron-capture cross sections of neutron-rich nuclei are calculated using a Hauser–Feshbach model when direct experimental cross sections cannot be obtained. A number of codes to perform these calculations exist, and each makes different assumptions about the underlying nuclear physics. We investigated the systematic uncertainty associated with the choice of Hauser-Feshbach code used to calculate the neutron-capture cross section of a short-lived nucleus. The neutron-capture cross section for ^{73}hbox {Zn}(n,gamma )^{74}hbox {Zn} was calculated using three Hauser-Feshbach statistical model codes: TALYS, CoH, and EMPIRE. The calculation was first performed without any changes to the default settings in each code. Then an experimentally obtained nuclear level density (NLD) and gamma -ray strength function (gamma hbox {SF}) were included. Finally, the nuclear structure information was made consistent across the codes. The neutron-capture cross sections obtained from the three codes are in good agreement after including the experimentally obtained NLD and gamma hbox {SF}, accounting for differences in the underlying nuclear reaction models, and enforcing consistent approximations for unknown nuclear data. It is possible to use consistent inputs and nuclear physics to reduce the differences in the calculated neutron-capture cross section from different Hauser-Feshbach codes. However, ensuring the treatment of the input of experimental data and other nuclear physics are similar across multiple codes requires a careful investigation. For this reason, more complete documentation of the inputs and physics chosen is important.

Analysis of radiation shielding effectiveness of hydride and borohydride metals for nuclear industry

Hydrogenous materials are of great interest in nuclear industry because of the ability of hydrogen to slow-down neutrons. In comparison with common hydrogenous materials, metal hydrides highly enriched with hydrogen at high temperatures and good mechanical properties. This study concerns on evaluation the requirements for photons, charged particles and fast neutrons shielding parameters for several types of hydride and borohydride metals. Firstly, photons shielding ability of these materials were examined using Phy-X/PSD program. Assigning the attenuation factors of photons were evaluated for an energy range (1 keV–100 GeV). Secondly, the charged particle (electron, proton and alpha particle) interactions were examined by using ESTAR (Stopping Powers and Ranges for Electrons) and SRIM (The Stopping and Range of Ions in Matter) programs. Finally, fast neutron attenuation was tested by calculating the removal cross-sections (ΣR) for neutron with energy 4.5 MeV. The dependence of photons, charged particles and fast neutrons parameters on the constituent of tested samples was given and discussed. The obtained results through this work show that, the shielding proficiency parameters depend on the energy of the incident radiation and chemical constituents of the examined materials. As well as, these results showed that BaH2, ZrH2, VH2 and TiH2 samples own a good shielding performance compared to the other investigated samples.

Hot hydrogen testing of direct current sintered NbC and (ZrNbW)C for nuclear thermal propulsion

Refractory carbides are promising materials for use in nuclear thermal propulsion, either as the primary material in nuclear fuel elements or as protective coatings. In particular, NbC and ZrC have high melting points, relatively low vapor pressures at elevated temperatures, and low thermal neutron cross sections. In this study, NbC and (Zr0.4Nb0.4W0.2)C were fully consolidated by direct current sintering, and relative densities of 98% were achieved. The samples were characterized by scanning electron microscopy and X-ray diffraction before and after hydrogen testing. The samples were exposed to hot flowing hydrogen gas at a pressure of 1 atm and a flow rate of 8 SLPM, at temperatures up to 2500 K, and for durations up to 3 h. The mass loss rate (MLR) of NbC decreased with increasing exposure, before reaching a steady state value of about 1 mg/m2s. For (ZrNbW)C, a comparison was made by exposing samples to either hot flowing hydrogen or argon gas. The mass loss rates decreased with increasing exposure to either gas, but the final steady-state MLR was one to two orders of magnitude greater for those samples exposed to hydrogen. As evidenced by X-ray data, small amounts of residual oxygen in the (ZrNbW)C samples were preferentially removed upon hydrogen exposure.

Influence of B2O3/Bi2O3 on the radiation absorption and nuclear safety of bismuth borosilicate glass system

In the quest to probe the potential of (40-x)B2O3–39SiO2–(20+x)Bi2O3–1Dy2O3 (BSBD1 –BSBD4) glasses, their interaction parameter with multispectral radiation is calculated, analyzed and presented in this report. The FLUKA code was used to simulate photons, charged particles (electrons, protons, α-particle, and carbon ions), and slow neutron cross sections of the glasses. In addition, the fissile and thermal neutron attenuation cross sections were evaluated through analytic expressions. The mass attenuation coefficient of the BSBD glasses varied from 0.0364 cm2/g to 68.9445 cm2/g for BSBD1, 0.0382 cm2/g to 80.1942 cm2/g for BSBD2, 0.0392 cm2/g to 87.5237 cm2/g for BSBD3, and 0.04001 cm2/g to 92.6780 cm2/g for BSBD4. Other photon attenuation parameters showed that the BSBD4 glass is the best photon attenuator. The range and stopping power of the charged particles also revealed that charged particle attenuation depends strongly on the density and Bi2O3 quantity. The neutron absorbing/shielding ability of the glasses showed a contrasting trend compared to photons and charged particles. While BSBD2 was the best fast neutron absorber, BSBD1 on the other hand, showed greater resilience to thermal and slow neutrons. Based on the analyzed data analyzed in this study, the BSBD glasses showed great potential for radiation protection ability.

Cross section for inelastic neutron scattering from 193Ir at 6 MeV

Two samples, one of natural iridium and the other of enriched 193Ir, were irradiated with a monoenergetic neutron beam of energy 6.0 MeV at the Triangle Universities Nuclear Laboratory. The product of the 193Ir(n,n′)193mIr reaction was determined by means of measuring X-rays following electron conversion of the isomeric state at 80.2 keV in 193Ir. The cross section for inelastic neutron scattering is reported disagreeing with the literature data.

Neutron capture cross section of ^{83}Kr

The neutron capture cross section of ^{83}Kr has been measured via the time-of-flight technique between 25 meV and 500 keV. The experiment used the DANCE array at the Los Alamos National Laboratory. Maxwellian Averaged Cross Sections have been derived for a range of stellar temperatures and are found to be in good agreement with previous data. The impact of the new cross sections on stellar nucleosynthesis has been investigated.

A study on characteristics of background gamma spectrum from LaCl3 detector in pulsed neutron logging

The scintillator detectors such as LaCl3(Ce) play an important role in some fields of scientific research, environment, safeguards, medicine, security and industry due to its superior energy resolution and exceptional luminescence properties, etc. However, Cl element in a LaCl3 crystal produces uncertainty of determining oil saturation in pulsed neutron logging because of the background spectrum caused by secondary gamma ray from the reaction of Cl nuclei with the neutron. In this paper, we employed Monte Carlo method to simulate secondary gamma ray generated LaCl3 crystal induced by thermal neutron with different borehole and formation conditions and establish a reference spectrum of Cl element. The relations between elemental window or peak areas counts and borehole and formation conditions were also investigated. The background was obtained by combining the reaction rate derived from thermal neutron capture cross section for Cl element and neutron flux with the reference spectrum. The results indicate that the contribution of secondary gamma ray to measuring spectrum decreases with formation porosity, limestone content, borehole diameter, and water salinity increasing. Nevertheless, the relative peak areas of Cl at different energies remain constant, indicating that the logging conditions have less of an effect on the background spectrum shape. As evidenced by the measured spectra in the sandstone and limestone calibration wells processed, the peaks of Si and Ca elements are enhanced while the peaks of Cl element are weakened. After subtracting detector background, the computations of oil saturation based on calibration wells are 38% more accurate than the original method.

Effect of Ce-dopant on photon interaction parameters of zirconolite-based borosilicate glass ceramics

In the present investigation, 70 wt% of zirconolite (CaZr1-xCexTi2O7, x = 0, 0.2, 0.4, 0.6, 0.8 and 1) and 30 wt% of sodium borosilicate glass (NaB0.5Al0.5Si2O6) (sample C0 to C10) have been synthesized. The photon interaction parameters viz. mass attenuation coefficient, effective atomic number, effective electron density, mean free path, and half value layer have been investigated for glass-ceramic samples at incident photon energies of 59.58 keV, 81.2 keV, 356 keV, 662 keV, 1173 keV and 1333 keV. Furthermore, effective neutron removal cross-section values, mass stopping power (MSP) and range of alpha particles and proton, were also calculated using SRIM codes. Moreover, gamma-ray energy absorption buildup factor (EABF) and exposure buildup factor (EBF) were also estimated. The experimental results are in agreement with the theoretically calculated values using WinXCOM program. Among the given prepared samples, C10 shows the best shielding properties, and C0 shows the best capability for neutron absorption. The MSP and projected range of α-particles and protons in all the prepared samples show very small variation with the change of Ce content in zirconolite in the energy range upto 15 MeV. The measured values are extensively desirable in fission and fusion reactors, medical applications, shielding researchers, and radiation physics.

Neutron capture cross-section measurement by mass spectrometry for Pb-204 irradiated in JRR-3

ABSTRACT Reliable neutron capture cross sections of lead isotopes are necessary for the development of lead or lead-bismuth cooled fast reactors. Although 204Pb has the smallest natural abundance in stable Pb isotopes, its neutron capture cross-section data are important because a long-lived radionuclide 205Pb (T1/2 = 17.3 million years) is produced by a neutron capture reaction on 204Pb. This study applied a mass spectrometry method to measure the neutron capture cross section of 204Pb. An enriched 204Pb sample was irradiated for 24 days with a neutron flux on the order of 1013 n/cm2/sec at the Japan Research Reactor-3. The irradiated 204Pb sample was analyzed by thermal ionization mass spectrometry to obtain the isotopic ratio between 205Pb and 204Pb. Consequently, the thermal-neutron capture cross-section of 204Pb was found to be 0.536 ± 0.030 barns.

Development of transient Monte Carlo in a fissile system with β-delayed emission from individual precursors using modified open source code OpenMC(TD)

In deterministic and Monte Carlo transport codes, β-delayed emission is included using a group structure where all of the precursors are grouped together in 6 groups or families, but given the increase in computational power, nowadays there is no reason to keep this structure. Furthermore, there have been recent efforts to compile and evaluate all the available β-delayed neutron emission data and to measure new and improved data on individual precursors. In order to be able to perform a transient Monte Carlo simulation, data from individual precursors needs to be implemented in a transport code. This work is the first step towards the development of a tool to explore the effect of individual precursors in a fissile system. In concrete, individual precursor data is included by expanding the capabilities of the open source Monte Carlo code OpenMC. In the modified code – named Time Dependent OpenMC or OpenMC(TD)– time dependency related to β-delayed neutron emission was handled by using forced decay of precursors and combing of the particle population. The data for continuous energy neutron cross-sections was taken from JEFF-3.1.1 library. Regarding the data needed to include the individual precursors, cumulative yields were taken from JEFF-3.1.1 and delayed neutron emission probabilities and delayed neutron spectra were taken from ENDF-B/VIII.0. OpenMC(TD) was tested in a monoenergetic system, an energy dependent unmoderated system where the precursors were taken individually or in a group structure, and in a light-water moderated energy dependent system, using 6-groups, 50 and 40 individual precursors. Neutron flux as a function of time was obtained for each of the systems studied. These results show the potential of OpenMC(TD) as a tool to study the impact of individual precursor data on fissile systems, thus motivating further research to simulate more complex fissile systems.

Study of mechanical, thermal, and nuclear radiation attenuation properties of tourmaline loaded heavy weight heat resistant concrete as reactor shielding materials

Three heavy weight concrete mixtures incorporating barite and goethite as coarse and fine aggregates, respectively, were prepared in order to examine the nuclear radiation attenuation properties of such concrete. The three mixtures incorporated 0 wt%, 25 wt%, and 50 wt% tourmaline ore by mass as partial replacement of fine aggregates (goethite). Workability in terms of slump test and mechanical properties in terms of compressive and tensile strength tests were carried out. Moreover, physical properties in terms of bulk density and water absorption and porosity tests were done. Finally, thermal analysis and nuclear radiation attenuation measurements were performed. Ten different gamma-ray energy lines up to 1407 keV were used to determine the mass attenuation coefficients (σ) of gamma rays for the concrete specimens. Good agreement was obtained between the experimental results and theoretical calculations carried out using “WinXCom” computer program. Four neutron energy bands (slow neutrons, total slow neutrons, neutrons of energies greater than 0.4 eV as well as neutrons of energies greater than 10 keV) were used to determine the macroscopic neutron cross-sections (Σ) for the concrete mixtures under investigation. Additionally, once the optimum results of mixing proportions were achieved, the study was expanded to examine the effect of heat on the mechanical and nuclear properties of such concretes. Results showed that the addition of tourmaline ore to the concrete mixtures has significantly improved its mechanical properties, in addition to the obvious improvement in its ability to attenuate neutrons of different energy bands. Heating of concrete specimens up to 400 °C show no significant effect on the attenuation of gamma rays, while a slight decrease in the ability of samples to attenuate neutrons. Finally, concrete mix containing 50 wt% tourmaline is recommended as a heat resistant concrete at 100 °C and 400 °C, for power nuclear reactors.

Measurement of the spin-wave stiffness and energy gap in the magnon spectrum of amorphous ferromagnets by small-angle scattering of polarized neutrons

This work studies the magnetic excitations of amorphous ferromagnetic alloys (Fe40Ni40P14 11B6 and Fe48Ni34P18) by small-angle polarized neutron scattering in the geometry when the magnetic field is inclined towards the neutron beam direction. Polarized neutrons are used in order to extract the scattering arising from the spin waves only. In this case, the energy-integrated neutron cross section contains a component which depends on neutron polarization and has a left–right asymmetry in the plane determined by the directions of the field and the neutron beam. Small-angle polarized neutron scattering measurements on spin waves in amorphous iron–nickel alloys were performed at different values of the external magnetic field H and neutron wavelength λ. The spin-wave spectrum is quadratic in terms of the momentum transfer and contains both a field gap and an inherent gap Δ of a non-field nature: E q = Aq 2 + gμB H + Δ. Only by measuring simultaneously two dependencies of the cut-off angle θc – on the applied magnetic field and on the neutron wavelength – is it possible to obtain reliable information about the spin-wave stiffness and the spin-wave gap in these amorphous ferromagnetic systems.

Rheological, Mechanical, Microstructural and Radiation Shielding Properties of Cement Pastes Containing Magnetite (Fe3O4) Nanoparticles

This work examines the influence of iron oxide nanoparticles (Fe3O4 NPs) on neutron and gamma-ray radiation shielding characteristics of Portland cement paste. Experimental evaluations were supplemented with theoretical studies using NXCom program. Portland cement pastes with 5, 10, 15, 20, and 30 wt% of nanomagnetite cement replacement were produced. Moreover, rheological, early strength development, compressive strength, and mercury intrusion porosimetry (MIP) tests were performed. The results showed that increasing the amount of Fe3O4 NPs in a mix leads to a gradual increment in measured viscosity and yield stress. High nano-Fe3O4 content substantially impeded the early strength development process and led to a decrement in the 7- and 28-day compressive strength of cement paste. The MIP studies exhibited a gradual increment in total porosity, and average pore volume, as nano-Fe3O4 content was increased. All the macroscopic cross-sections of slow, fast and thermal neutrons constantly increased as a result of the addition of magnetite nanoparticles, with their variations being markedly linear. Similarly, gamma attenuation test results indicated that the addition of Fe3O4 powder enhances the shielding capability of paste in the energy range of interest (0.08–2.614 MeV). In conclusion, Fe3O4 nanoparticles can be successfully used in producing lead-free cementitious composites with improved gamma-ray and neutron shielding properties. However, certain drawbacks related to an increment in matrix porosity and thus a decrement in mechanical performance should be taken into account.

Deuterated Malonamide Synthesis for Fundamental Research on Solvent Extraction Systems

ABSTRACT Malonamide derivatives, which are the most extensively investigated extractants for solvent extraction of lanthanides, actinides, and platinum group metal ions, were deuterated by using Pd/C and Rh/C catalysts in a D2O/2-propanol mixture. This method replaces 1H atoms in the malonamides with 2H atoms at a controllable deuteration rate. The maximum rate reached about 75%, determined by nuclear magnetic resonance and electrospray ionization – mass spectrometry. The extraction behavior of metal ions by the malonamides was unchanged by the deuteration. Deuterated malonamides are a powerful tool for fundamental research on solvent extraction systems and for structural analysis of organic phases. The large difference in the cross section of coherent neutron scattering between 1H and 2H leads to a large difference in neutron scattering length density of malonamide derivatives before and after the deuteration. Therefore, using deuterated malonamides in small-angle neutron scattering and neutron reflectivity studies may help to reveal the microscopic structure of the specific solute species in the bulk organic phase and at the liquid–liquid interface, respectively. In this paper, we report the advantage of the deuterated malonamides for a contrast-matching method in the SANS experiment. This deuteration method could be generalized and extended to a wide variety of extractant molecules.

Evaluation of neutronics parameters during RSG-GAS commissioning by using Monte Carlo code

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

Direct reactions for nuclear astrophysics

The neutron activation technique is a well established method to measure neutron capture cross sections relevant for the s-process. The7Li(p,n) reaction at Ep= 1912 keV is often used as a neutron source since the energy distribution of the emitted neutrons closely resembles a Maxwell-Boltzmann spectrum ofkBT= 25 keV, mimicking the22Ne(α,n) phase in TP-AGB stars. The weak s-process, which takes place in massive stars, can reach energies up tokBT = 90 keV. Neutron spectra corresponding to a Maxwell-Boltzmann distribution withkBT> 25 keV cannot be produced by the7Li(p,n) reaction directly. We developed a method to obtain quasi-Maxwellian neutron capture cross sections over a wide energy range by combining a set of spectrum average cross sections measured at six different proton energies and distances between the lithium target and the sample. The measured spectrum averaged cross section can be used to calculate the Maxwellian-Averaged cross-section (MACS) fromkBT= 25 keV tokBT= 90 keV. Over the last two years neutron capture cross sections on over 20 isotopes have been measured at Goethe University Frankfurt using this methodology. An overview of the current experimental method, challenges during data analysis and the first results are presented.