Neutron Energy Range Research Articles

Cross section empirical formulation for (n, 2n) nuclear reactions on natural isotopes from Z= 21 to Z= 79 for neutron energy range, 8–20 MeV

Neutron induced reactions with different target materials and isotopes are very important for completion of nuclear data base as well as for many applications. Since the experimental determinations of the reaction cross sections for such reactions are not always available, attempts were made to introduce empirical or semi empirical formulae. In this work, new empirical formulation for (n, 2n) nuclear reactions on natural isotopes for neutron energies range 8–20 MeV, based on some published experimental works. The deuced formulae are then used to calculate the excitation functions for the (n,2n) on target isotopes having atomic numbers in the range 21–79. The cross section values for many reactions, having previously measured values, are determined and inter-comparison with the experimental and the TENDL-2019 database values are given using some statistical error parameters. Special attention was made to some Lanthanide isotopes. The results showed satisfied predictions in most cases.

Development of back-illuminated thin silicon diode applied to fast neutron sensor in active personal dosimeter

A back-illuminated thin Si diode was developed for fast neutron sensor in active personal neutron dosimeter for detecting fast neutrons below 1 MeV. Neutron response functions for the fast neutron sensor with the back-illuminated thin Si diode were evaluated based on measurements of monoenergetic neutron beams at the National Metrology Institute of Japan and Monte Carlo simulations using the Particle Heavy Ion Transport Code System in the neutron energy over 500 keV. The lower limit of neutron detection energy was decreased to 500 keV and neutron detection efficiencies were increased several times over other fast neutron sensors. It is possible to measure neutron doses in the neutron energy range of 500 keV–1 MeV using this novel fast neutron sensor, but the neutron doses were undetectable using other fast neutron sensors. The neutron detection efficiencies were closer to the personal dose equivalent conversion curve than those for other fast neutron sensors in this neutron energy range. Neutron doses of individuals employed at a nuclear facility were evaluated using the active personal neutron dosimeter, which consists of both the novel fast and slow neutron sensors.

Decomposition of a Tissue Into Its Constituent Elements for Specifying the Values of Doses

In this research, a real tissue has been decomposed to its constituent materials and elements applying its volumetric density and initial mass. This study has been carried out to obtain the exact amounts of constituent elements existing in liver tissue as well as specify the absorbed dose because of the tissue getting exposed to radiation in radiotherapy and radiography. In order, a medical model has been defined in a way that its compositions are just like the materials existing in the tissue of a human. Then, the accurate mass, density, and volume of every element in it are specified. In the next stage, the related tissue is exposed to a neutron beam, and then the values of doses absorbed in the main constituent material namely water, and also the total absorbed dose is obtained by the MCNP nuclear code. This study can be used to study the radiotherapy and radiography scopes for obtaining the absorbed dose as a result of interaction between radiation beam and biological cells. In this research, the values of absorbed doses have been obtained for a wide range of neutron energies in a way that the absorbed doses can accurately be obtained.

The 3He BF3 Giant Barrel (HeBGB) neutron detector

(α,n) reactions play an important role in nuclear astrophysics and applications and are an important background source in neutrino and dark matter detectors. Measurements of total (α,n) cross sections employing direct neutron detection often have a considerable systematic uncertainty associated with the energy-dependent neutron detection efficiency and the unknown initial neutron energy distribution. The 3He BF3 Giant Barrel (HeBGB) neutron detector was built at the Edwards Accelerator Laboratory at Ohio University to overcome this challenge. HeBGB offers a near-constant neutron detection efficiency of (7.5 ± 1.2)% over the neutron energy range 0.01 MeV–9.00 MeV, removing a significant source of systematic uncertainty present in earlier (α,n) cross section measurements.

Neutron activation cross section data library

To satisfy the requirements of nuclear reaction cross sections in nuclear engineering applications and nuclear physics studies, the Neutron Activation Cross Section Data Library has been established. 818 target nuclei including unstable target or isomeric target nuclei are considered in this library. The induced neutron energy range region is between 10−5 eV and 20 MeV. The standard ENDF-6 format is adopted, including general information, reaction cross sections, multiplicities, and so on. The recommended reaction cross sections were obtained using UNF code system and FDRR nuclear model codes or systematic analysis based on available experimental data. The full evaluated dataset containing the evaluated activation cross sections is openly available at http://doi.org/10.57760/sciencedb.j00113.00024.

Energy Dependence of Atmospheric Neutron-Induced Failures in Silicon Carbide Power Devices

A time-of-flight measurement technique is used to determine energy-dependent neutron cross sections of silicon carbide power devices exposed to a beam that approximates atmospheric neutrons up to 600 MeV. The Irradiation of Chip Electronics (ICE)-II neutron beam at the Los Alamos Neutron Science Center (LANSCE) is used during these experiments. The measurements show that a wide range of neutron energies are likely to cause device failures close to the rated voltage, from neutron interactions with lattice atoms within the active device volume. The data indicate that the energy-dependent cross sections are relatively large, revealing a high probability for a few percent of the neutron–lattice atom interactions to give rise to a failure when the device is biased close to its rated voltage. Additionally, knowledge of the neutron energies and associated cross sections giving rise to failures within these devices provide insights regarding the critical charges and the sensitive volumes of these devices as a function of neutron energy for single-event burnouts. Lastly, the energy-dependent cross sections provide a preliminary set of data to calculate a correlation factor between failure rates obtained using a broad energy neutron source, and the commonly used mono-energetic high-energy proton beams, utilized due to their wider availability for determining bias dependent failure-in-time curves.

Study on the measurement of high energy neutrons using Zirconium capped CR-39 (Zr-CR-39) combined detector - Monte Carlo simulation and experimental validation

Neutron measurement, be it dose or fluence is a challenging field in high energy accelerators where the neutron energy varies from thermal to high-energy neutrons usually extending beyond 20 MeV. CR-39 SSNTD (Solid State Nuclear Track Detector) used in neutron measurements has near flat energy response in the range of 500 keV to 10 MeV. By capping the CR-39 with a suitable converter such as Zirconium (Zr), it's response can be enhanced in the higher neutron energy ranges (beyond 10 MeV). Zr has significant (n,2n) cross-section of ∼1 b in the neutron energy range of 10–20 MeV. Hence, when high energy neutrons impinge on Zr, neutrons of lower energy are produced by (n,2n) reactions, which create additional recoil proton tracks in CR-39. At higher neutron energies (>20 MeV) additional tracks can also be created due to other charged secondaries such as deuteron, triton, alphas and recoil carbon and oxygen. The enhancement in track density of CR-39 in presence of Zr as compared to that of bare CR-39 is termed as the Total Enhancement Factor (TEF). Monte Carlo simulations with FLUKA and GEANT4 codes are used to optimize the Zr converter thickness with 14 MeV and 20 MeV, and are also verified with measurements. At an optimum converter thickness of 1 mm, FLUKA and GEANT4 Monte Carlo codes were used to observe the enhancement in the response of the combined detector (Zr-CR-39) against the bare detector for different mono-energetic neutron beams in the range 14 MeV - 1 GeV. The enhancement obtained for Zr converter are also compared with Lead (Pb) converter. Experimental studies were carried out to validate the simulations by irradiating Zr-CR-39 to 14 MeV, 20 MeV and high energy neutrons spectra available at CERF facility. Measured TEF values are 1.27, 2.5 and 7.85 for 14 MeV, 20 MeV and CERF neutrons spectra respectively. Results of simulations and measurements are found to be in agreement within 20%.

New cross section measurements on tungsten isotopes around 14 MeV neutrons and their excitation functions * *Supported by the National Natural ScienceFoundation of China (11575090)

New cross sections of the 183W(n,α)180mHf, 186W(n,d*)185Ta, 182W(n,p)182Ta, 184W(n,p)184Ta, 182W(n,2n)181W, 184W(n,α)181Hf, and 186W(n,α)183Hf reactions were measured in the neutron energy range of 13.5-14.8 MeV via the activation technique to improve the database and resolve discrepancies. Monoenergetic neutrons in this energy range were produced via the T(d,n)4He reaction on a solid Ti-T target. The activities of the irradiated monitor foils and samples were measured using a well-calibrated high-resolution HPGe detector. Theoretical calculations of the excitation functions of the seven nuclear reactions mentioned above in the neutron energies from the threshold to 20 MeV were performed using the nuclear theoretical model program TALYS-1.9 to aid new evaluations of cross sections on tungsten isotopes. The experimental data obtained were analyzed and compared with that of previous experiments conducted by other researchers, and with the evaluated data available in the five major evaluated nuclear data libraries of IAEA (namely ENDF/B-VIII.0 or ENDF/B-VII.0, JEFF-3.3, JENDL-4.0u+, CENDL-3.2, and BROND-3.1 or ROSFOND-2010), and the theoretical values acquired using TALYS-1.9 nuclear-reaction modeling tools. The new cross section measurements agree with those of some recent experiments and theoretical excitation curves at the corresponding energies. The consistency of the theoretical excitation curves based on TALYS-1.9 with these experimental data is better than that of the evaluated curves available in the five major nuclear data libraries of IAEA.

Cross-section measurements of (n, 2n) and (n, p) reactions on 124,126,128,130,131,132Xe in the 14 MeV region and theoretical calculations of their excitation functions * *Supported by the National Natural Science Foundation of China (11875016, 12165006)

The reaction cross-sections of 124Xe(n, 2n)123Xe, 126Xe(n, 2n)125Xe, 128Xe(n, 2n)127Xe, 130Xe(n, 2n)129mXe, 132Xe(n, 2n)131mXe, 130Xe(n, p)130I, 131Xe(n, p)131I, and 132Xe(n, p)132I were measured at the 13.5, 13.8, 14.1, 14.4, and 14.8 MeV neutron energies. The monoenergetic neutrons were generated via the 3H(d,n)4He reaction at the China Academy of Engineering Physics using the K-400 Neutron Generator with a solid 3H-Ti target. A high-purity germanium detector was employed to measure the activities of the product. The reactions 93Nb(n, 2n)92mNb and 27Al(n, α)24Na were adopted for neutron flux calibration. The cross sections of the (n, 2n) and (n, p) reactions of the xenon isotopes were obtained within the 13–15 MeV neutron energy range. These cross-sections were then compared with the IAEA-exchange format (EXFOR) database-derived experimental data, together with the evaluation results of the CENDL-3, ENDF/B-VIII.0, JENDL-4.0, RUSFOND, and JEFF-3.3 data libraries, as well as the theoretical excitation function obtained using the TALYS-1.95 code. The cross-sections of the reactions (except for the 124Xe(n, 2n)123Xe and 132Xe(n, p)132I) at 13.5, 13.8, and 14.1 MeV are reported for the first time in this study. The obtained results are beneficial in providing better cross-section constraints for the reactions in the 13–15 MeV region, thus improving the quality of the corresponding database. Meanwhile, these data can also be used for the verification of relevant nuclear reaction model parameters.

Cross-section of (n,2n) reaction for niobium and strontium isotopes between 13.97 to 20.02 MeV neutron energies

The activation cross-sections of 93Nb(n,2n)92Nbm and 88Sr(n,2n)87Srm reactions were measured using the activation and off-line γ-ray spectrometry technique in the neutron energy range 13.97–20.02 MeV. The present measurements have been done at the neutron energies where there are deficiencies and scarcity in the reaction cross-section data. The neutron flux was determined using the 27Al(n,α)24Na monitor reaction. The γ-ray self-attenuation effect and the low energy background neutron corrections were done in this experiment. The results of the present work were compared with the previously published data and evaluated nuclear data libraries. Theoretically, the cross-section for 93Nb(n,2n)92Nbm and 88Sr(n,2n)87Srm reactions was predicted by TALYS-1.9 nuclear code and compared with the present results.

Measurements of the Bi210g to Bi210m activation ratio for the Bi209 ( n ,γ) reaction with thermal and epithermal neutrons at the Soreq IRR1 reactor

Irradiations of bismuth samples with thermal and epithermal neutrons have been performed for a comparison of activation to the $^{210g}\mathrm{Bi}$ ground state and to the $^{210m}\mathrm{Bi}$ metastable state. The bismuth irradiations were conducted at the Soreq IRR1 research reactor at a near-core location with integrated neutron flux of approximately $2\ifmmode\times\else\texttimes\fi{}{10}^{18}\mathrm{neutrons}/\mathrm{c}{\mathrm{m}}^{2}$ and a gold cadmium ratio of 3.1. Two nearly identical bismuth samples were irradiated, one without and one with a cadmium shield, to enable a comparison between thermal and epithermal neutron capture cross sections. Subsequent gamma spectrometry indicates nearly equal levels of activation to the ground and to the metastable states of $^{210}\mathrm{Bi}$, both with thermal and epithermal neutrons. Absolute thermal cross sections and resonant integrals were deduced by including previously reported bismuth activation measurements performed conjointly with gold foil activation for normalization. A comparison of the present results with existing data is presented. It is argued that the bismuth resonances lie in the neutron energy range relevant for stellar temperatures, and therefore the neutron capture cross section ratio $({\ensuremath{\sigma}}_{g}/{\ensuremath{\sigma}}_{m})$ measured for the epithermal neutrons is relevant also for stellar $^{210}\mathrm{Bi}$ production. This is of importance since bismuth lies at the end of the chain for the s\penalty1000-\hskip0ptprocess reaction for nucleosynthesis in stars. Precise knowledge of bismuth activation to the $^{210m}\mathrm{Bi}$ metastable state is also vital for the planning of GEN-IV reactors given the 3-million-yr half-life of $^{210m}\mathrm{Bi}$.

Modeling ionization quenching in organic scintillators

Ionization quenching models were assessed by evaluating light yield data from multiple organic scintillators and recoil ions over a fission spectrum neutron energy range, important for basic science and applications.

Compton Imaging and Machine-Learning techniques for an enhanced sensitivity in key stellar (n,γ) measurements

Neutron capture cross-section measurements are fundamental in the study of astrophysical phenomena, such as the slow neutron capture (s-) process of nucleosynthesis operating in red-giant stars. To enhance the sensitivity of such measurements we have developed the i-TED detector. i-TED is an innovative detection system which exploits the Compton imaging technique with the aim of obtaining information about the incoming direction of the detected γ-rays. The imaging capability allows one to reject a large fraction of the dominant γ-ray background, hence enhancing the (n,γ) detection sensitivity. This work summarizes the main results of the first experimental proof-of-concept of the background rejection with i-TED carried out at CERN n_TOF using an early i-TED demonstrator. Two state-of-the-art C6D6 detectors were also used to benchmark the performance of i-TED. The i-TED prototype built for this study shows a factor of ~3 higher detection sensitivity than C6D6 detectors in the ~10 keV neutron-energy range of astrophysical interest. This works also introduces the perspectives of further enhancement in performance attainable with the final i-TED array and new analysis methodologies based on Machine-Learning techniques. The latter provide higher (n,γ) detection efficiency and similar enhancement in the sensitivity than the analytical method based on the Compton scattering law. Finally, we present our proposal to use this detection system for the first time on key astrophysical (n,γ) measurements, in particular on the s-process branching-point 79Se, which is especially well suited to constrain the thermal conditions of Red Giant and Massive Stars.

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.

Exploring nuclear photonics with a laser driven neutron source

Neutron beams have been providing indispensable tools for a wide range of fields in modern science and engineering. Recently, a new type of pulsed neutron source has been developed, known as laser-driven neutron source (LDNS). The LDNSs utilize laser-accelerated ions, including protons and deuterons as a primary beam and generate neutrons from a secondary target (lithium, beryllium, etc) via a nuclear reaction. Applying an additional moderator part, LDNSs can provide a broad energy range of neutrons (meV ∼ MeV). This paper aims to introduce the current status of LDNS and the results of application-oriented experiments implemented at the Institute of Laser Engineering (ILE) of Japan.

Fast neutron spectra measurement in a copper using a 252Cf standard neutron source

The presented paper focuses on the measurement of fast neutron leakage spectra in the copper assembly using an intense standard 252Cf(sf) source. The neutron spectra were measured by a stilbene organic scintillation detector and compared with MCNP6.2 calculations using copper nuclear data from the ENDF/B-VIII.0, ENDF/B-VII.1, JEFF-3.3, JEFF-3.2, JENDL-4.0 and CENDL-3.1 libraries in the energy range of 1–10 MeV. Discrepancies appear in the intermediate neutron energy range between experiment and calculation. The best agreement with an experiment was achieved using ENDF/B-VIII.0 library in the region of 1–4.5 MeV. In the region of 4.5–10 MeV, the best agreement was achieved using JENDL-4.0 or CENDL-3.1 library. The largest discrepancies in the whole energy range were found for calculations involving JEFF-3.3 copper evaluations.

Evaluation of neutron shielding performance for some alloys

Alloys, in general, and stainless steel, in particular, are among the basic materials for shielding against nuclear radiation. Some selected alloys among a lot of shielding materials in numerous literature were investigated as shielding materials against neutrons in a wide neutron energy range of 10−5 eV to 20 MeV. The neutron mass attenuation coefficient (NMAC) for the materials under investigation was calculated using the Monte Carlo N-Particle transport code, version 5 (MCNP5). The results indicate that the alloys under investigation have a good shielding performance against neutrons along the studied neutron energy range especially 5% B4C (S8) alloy. The 5% B4C sample shows a high shielding performance for neutrons of energy < 1.6 x 10-3 MeV as compared to the other investigated alloys. While the stainless steel 316 (S3) shows the best shielding performance for neutrons of energies > 0.2 MeV. The study also shows that the alloy must contain a percentage of a light element that has a high ability to absorb slow neutrons such as boron. This study is useful in selecting the appropriate shielding material for specific neutron energies. MCNP5 calculations were confirmed by comparing its results for iron with the corresponding evaluated data ENDF/B-V extracted from International Atomic Energy Agency (IAEA) database.

Effect of using heavy aggregates on the high performance concrete used in nuclear facilities

For nuclear power plants construction, each country has to make use of its own available raw materials. Therefore, through the current research we had to search for the relevant local materials suitable for usage as concrete components and satisfy the needed requirements. The target of this research is to investigate the effect of some type’s heavy aggregates on the physico-mechanical properties and attenuation coefficient of the high performance heavy concrete used as radiation shielding of nuclear facilities. For this purpose radiation shielding properties of high performance heavy concrete mixture prepared using different types of heavy aggregates such as magnetite in addition to industrial waste of lead slag as a coarse aggregate and fine aggregate of local sand or replacing sand by fine magnetite or lead slag, were investigated comparatively with hardened dolomite concrete. The physico-mechanical properties of type aggregates and hardened concretes were examined. Gamma transmission parameters for the studied concrete for Cs137 and Co60 as gamma source were measured, also the effective removal cross-section (ΣR) of fast neutrons spectra for a wide range of neutron energies of [2–12 MeV] with thicknesses 4, 6, 8, 10 and 12 cm of heavy concrete mixes for Plutonium-Beryllium as neutrons source. While the thermal neutrons tests were performed using californium radioisotope. An innovative material that the current research has introduced to be used as a heavy coarse aggregate in HPC is “lead slag”, lead slag has been converted from a subsidiary industrial and environment polluting waste to a very valuable material as one of the ingredients of heavy concrete. It was found that, the magnetite and lead slag aggregates improved the physico-mechanical properties of heavy concrete compared to dolomite concrete. Replacing fine sand by the fine aggregates of magnetite or lead slag enhances the efficiency of attenuation by the concrete mixes against these radiations and the compressive strength values increase by 12 and 27% respectively, of heavy concrete after 28 days. The concrete mix containing lead slag aggregate has higher mechanical properties and enhances the attenuation properties for fast neutrons, gamma rays and thermal neutrons than the other concrete mixes.

Measurement of U238(n,n′γ ) cross section data and their impact on reaction models

A better knowledge of $(n,xn)$ reaction cross sections is important for both reaction modeling and energy applications. This article focuses on inelastic scattering of neutrons off $^{238}\mathrm{U}$ for which improvements are needed to better constrain evaluations and solve inconsistencies in nuclear power reactor calculations. A new precise measurement of $(n,xn\ensuremath{\gamma})$ reaction cross sections on $^{238}\mathrm{U}$ has been performed at the GELINA (Geel Electron LINear Accelerator) neutron facility operated by EC-JRC-Geel (Belgium) with the GRAPhEME (GeRmanium array for Actinides PrEcise MEasurements) setup. The prompt $\ensuremath{\gamma}$-ray spectroscopy method coupled to time-of-flight measurements is used to extract $(n,xn\ensuremath{\gamma})$ cross section values which can be further combined to infer the total neutron inelastic scattering cross section. Cross section data for 18 $\ensuremath{\gamma}$ transitions (five never measured before) are presented and compared to the data in the literature. Emphasis is especially given to the uncertainty determination to produce partial cross section data as accurate as possible. Due to intrinsic limitations of the experimental method, the use of additional nuclear structure information coupled with theoretical modeling is required to determine the total $(n,{n}^{\ensuremath{'}})$ cross section over the whole neutron energy range. We have investigated modeling aspects of the $^{238}\mathrm{U}(n,{n}^{\ensuremath{'}}\ensuremath{\gamma})$ cross sections related to the description of compound nucleus and preequilibirum mechanisms as well as the discrete part of nuclear structure. Through comparison between experimental and calculated $(n,{n}^{\ensuremath{'}}\ensuremath{\gamma})$ cross sections, we pinpoint inaccuracies in the description of specific reaction mechanisms and challenge recently implemented models. This helps improving the whole modeling of the $(n,{n}^{\ensuremath{'}})$ reaction.

Spectrometer design of low energy neutrons for boron neutron capture therapy

We designed a neutron spectrometer to measure low-energy neutron beam for boron neutron capture therapy (BNCT) without exchanging thicknesses of neutron moderator. The spectrometer consists of a boron carbide block as a neutron moderator and nine thin neutron silicon sensors with variable detection efficiency. Using the Monte Carlo technique, we simulated the spectrometer’s neutron energy response functions. It represented sharp rise in the response functions with various lower limits of neutron energies distributed over a wide range of neutron energy from a thermal energy to a few 100 keV. The lower limit of neutron energy was dependent on a depth in the boron carbide block. By applying thicker enriched 6LiF neutron converter and covering the deep positioned neutron sensors with polyethylene plates, neutron detection efficiencies were increased to be able to measure the BNCT neutron beam in a short time.