Fast Neutron Region Research Articles

Reexamination of a 40Ca(n,α)37Ar cross section measurement from existing literature

Background 37Ar, produced from the neutron activation of 40Ca in sub-surface rock and soil, is important for nuclear explosion monitoring. Most of the cross section measurements done for the 40Ca(n,α) 37 Ar reaction are in the fast neutron region. A thermal cross section measurement was attempted in 1989, with the assumption that the 4-8 MeV neutrons would be completely thermalized with 8-cm of water. Methods The experimental setup was modeled using Monte Carlo N-Particle code to see if the assumptions for the neutron distribution were accurate. A model of a 3.0 MeV deuteron source was used to recreate the neutron source from the experiment. Additional modeling was done using SCALE to compare the predictions of 37Ar yields when using the measured cross section from 1989 compared to current nuclear data models. Results Modeling showed the neutron distribution was closer to 10% thermal and 90% fast as opposed to entirely thermalized. The additional modeling showed how the overestimation of the thermal cross section would affect the predicted yield of 37Ar. Conclusions Based on the modeling results, the assumption that all the 4-8 MeV neutrons are thermalized with 8-cm of water is not accurate. The high fraction of fast neutrons interacting with the Ca will result in an overestimation of the thermal cross section.

Reexamination of a 40Ca(n,α)37Ar cross section measurement from existing literature

Background 37Ar, produced from the neutron activation of 40Ca in sub-surface rock and soil, is important for nuclear explosion monitoring. Most of the cross section measurements done for the 40Ca(n,α) 37 Ar reaction are in the fast neutron region. A thermal cross section measurement was attempted in 1989, with the assumption that the 4-8 MeV neutrons would be completely thermalized with 8-cm of water. Methods The experimental setup was modeled using Monte Carlo N-Particle code to see if the assumptions for the neutron distribution were accurate. A model of a 3.0 MeV deuteron source was used to recreate the neutron source from the experiment. Additional modeling was done using SCALE to compare the predictions of 37Ar yields when using the measured cross section from 1989 compared to current nuclear data models. Results Modeling showed the neutron distribution was closer to 10% thermal and 90% fast as opposed to entirely thermalized. The additional modeling showed how the overestimation of the thermal cross section would affect the predicted yield of 37Ar. Conclusions Based on the modeling results, the assumption that all the 4-8 MeV neutrons are thermalized with 8-cm of water is not accurate. The high fraction of fast neutrons interacting with the Ca will result in an overestimation of the thermal cross section.

Prediction of fast neutron spectra with a spherical TEPC using a machine-learning algorithm

For many years, new unfolding methods based on machine learning have been studied and developed to improve the prediction capabilities of fluence spectra in neutron fields. These methods are mainly applied to traditional devices used for specific measurements like Bonner spheres, activation detectors or liquid scintillators. In this paper, we attempt to develop a new method based on the unfolding of the fluence spectrum from the micro-dosimetric spectrum measured by a tissue-equivalent proportional counter (TEPC). This type of counter is commonly used for neutron kerma measurements and quality factor assessments, but has never been employed as a neutron spectrometer. This work focuses on the fast neutron region, which is an extremely relevant subject in various fields of nuclear energy. We have tested different machine-learning models to define an unfolding algorithm that can be used to reconstruct the energy distribution of the fluence for spectra of various origins, ranging between 50 keV and 20 MeV.

Detection of fast neutrons with the pixel detector Timepix3

We examined the response of the pixel detector Timepix3 with silicon sensor to well-defined fast neutron fields. Part of the pixel detector silicon sensor was additionally equipped with a neutron mask of distinct converter regions. The mask consists of separate thermal and fast neutron regions using 6LiF and hydrogen (plastic) converters, respectively. Measurements were performed with mono-energetic fast neutrons produced at D-D and D-T sources from a Van de Graaff accelerator and a neutron generator, respectively. Data were collected with low background including measurements with moderator material to provide a thermalized neutron component. All the signals produced in the detector were analyzed and decomposed in terms of the spectral-tracking response of the pixel detector. The effect of the fast and thermal components of the neutron converter were determined and compared with direct interactions in the silicon sensor which are significant and can be dominant for fast neutrons. We identify and classify the neutron-induced tracks in terms of the broad-type particle-event track classes. A partial overlap is unavoidable with tracks from direct detection of other radiations in particular protons and low-energy light ions as well as X rays. This will limit the neutron-event discrimination in mixed-radiation fields. The detection response according sensor-mask region was examined and calibrated for the investigated neutron fields. The neutron detection efficiency is selectively derived for the detector particle-event classes. This approach enables to enhance the neutron-discrimination and suppress background and unwanted events. This work enables to extend the response matrix of the detector for broad-type radiations to include neutrons both fast and thermal. The results serve to enhance the sensitivity and determine the neutron component in unknown and mixed-radiation fields such as outer space and particle radiotherapy environments.

Neutron elastic and inelastic scattering differential cross sections on carbon

Elastic and inelastic neutron scattering angular distributions were measured on natural carbon samples to confirm existing experimental data and evaluations in the fast neutron region and to guide improvements in resonance parameters, where needed. Sixty-four (n,n′) differential cross section measurements were performed at 45 incident neutron energies between 0.5 and 8.0 MeV.Experimental angle-integrated elastic scattering cross sections are consistent with ENDF/B-VIII.0 values with the exception of the region from 3.2 to 4.0 MeV where our results are ∼3% higher. In the 3.4 to 3.6 MeV region our differential cross sections are slightly lower at forward angles and somewhat higher at backward angles than the ENDF/B-VIII.0 calculations; however, our results are consistent with measured data from other research groups. The first- through fourth-order elastic scattering Legendre coefficients from fits to these experimental data are consistent with ENDF/B-VIII.0 values across this range of incident neutron energies.Inelastic scattering cross sections were measured at 12 incident neutron energies between 5.6 and 7.8 MeV. Angle-integrated cross sections agree well with the 1978 Perey and ENDF/B-VIII.0 values; however, the angular-distribution shapes deviate somewhat from the ENDF/B-VIII.0 calculations. These data will allow refinement of the resonance parameter description in this difficult energy region.

Progress on the reevaluation and validation of the n+233U neutron cross sections

The set of 233U resonance parameters of the ENDF/B-VIII.0 nuclear data library was adopted from the previous ENDF/B-VII.1 evaluation using the external levels to update the thermal values. Adoption of IAEA 2017 thermal standards (σf=533.0±2.2 b, σc=44.9±0.9 b, and ν‾tot=2.487±0.011) and of the IAEA-recommended thermal-neutron induced prompt fission neutron spectrum (PFNS) with average PFNS energy of 2.030±0.013 MeV requires a re-evaluation of 233U neutron cross sections in the resolved resonance region. A newly produced evaluation is being tested on benchmarks carefully selected from the Handbook of International Criticality Safety Benchmark Experiments (ICSBEP) which are highly sensitive to 233U data. An important goal of this work was to eliminate the strong negative gradient of the calculated effective multiplication factors with respect to the epithermal fission fraction observed in the validation of the ENDF/B-VIII.0 library for those assemblies. A significant improvement in integral performance of critical 233U solutions is observed for the newly proposed evaluation. Further work addressing the fast neutron region is needed.

Neutron beam filter system for fast neutron cross-section measurement at the ANNRI beamline of MLF/J-PARC

A neutron filtering system has been introduced in the Accurate Neutron–Nucleus Reaction Measurement Instrument (ANNRI) beamline in the Materials and Life Science (MLF) facility of the Japan Proton Accelerator Research Complex (J-PARC) in order to produce quasi-monoenergetic neutron beams. The filtered neutron spectrum by the filter assemblies was analyzed by means of capture and transmission measurements and also by Monte Carlo simulations using PHITS. The characteristics of the filtered neutron beam are discussed alongside its viability in future applications for neutron cross-section measurements in the fast neutron region.

The neutron cross section of barite-enriched concrete for radioprotection shielding in the range 1 meV–1 keV

We present a characterization of a series of barite-enriched concrete samples at the atomic scale using neutron-based techniques. In particular, neutron transmission measurements provide the macroscopic cross section (also known as neutron removal cross section) as a function of the incident neutron energy in the range 1 meV–1 keV. In this range, where fewer experimental investigations are available in the literature, the cross section is dominated by the scattering events from hydrogen, as opposed to the fast-neutron region (MeV-energy neutrons) where capture events by Ba are more important. Moreover, below 1 eV, the cross section depends on the molecular or crystal structure of the components. For each sample, the amounts of barium and hydrogen are provided by neutron resonance capture analysis and deep inelastic neutron scattering, respectively. We find that the amounts of barium and hydrogen are correlated, with a lower amount of hydrogen in the samples with more barium, likely because of the absence of some calcium-silicon-hydrate structures, whose formation is inhibited by the presence of barite. Moreover, we quantify the non-negligible contribution to the neutron macroscopic cross section arising from water molecules loosely bound in concrete. This contribution makes the shielding performance of concrete dependent upon the humidity and temperature conditions affecting the installation. Our results provide additional guidelines for radioprotection workers to determine the optimal concentration of barite in mixed gamma/neutron radiation environments.

The research on the characteristics of target and core in a gas-cooled ADS with passive safety

The Accelerator Driven Subcritical Reactor System (ADS) has a strong ability to transmute minor actinides (MAs) and it has inherent safety because of its subcritical core. However, the removal of decay heat is a major problem for the ADS and the design of spallation target is still a difficulty. Therefore, this paper presents a design of gas-cooled ADS with passive safety to attain the good performance of removing decay heat and transmutation.For the design of the target, the helium-cooled tungsten target is decided to adopt, and it is modelled into the cone structure to strengthen the convective heat transfer. A series of numerical experiments have been made to determine the optimal parameters of the target using the MCNP and CFX code. In the design of core, the subcritical core is divided into fast neutron region which the MAO2 is placed in and thermal neutron region which the UO2 is placed in to improve the performance of the transmutation. Two methods are proposed in this paper for the thermal-hydraulic calculation in the steady-state analysis, and the results show its good thermal performance.

Evaluation of the neutron induced reactions on235U from 2.25 keV up to 30 MeV

An evaluation of fast neutron induced reactions on 235 U is performed in the 2.25 keV–30 MeV incident energy range with the code EMPIRE–3.2 Malta, combined with selected experimental data. The reaction model includes a dispersive optical model potential (RIPL 2408) that couples seven levels of the ground-state rotational band and a triple-humped fission barrier with absorption in the wells described within the optical model for fission. EGSM nuclear level densities are used in Hauser-Feshbach calculations of the compound-nuclear decay. The starting values for the model parameters are retrieved from the RIPL-3 data-base. Excellent agreement is achieved with available experimental data for neutron emission, neutron capture and fission, which gives confidence that the quantities for which there is no experimental information are also predicted accurately. In the fast neutron region of the evaluated file, the fission cross section is taken from Neutron Standards, and neutron capture includes fluctuations observed in recent experiments. Other channels are taken directly from model calculations. New evaluation is validated against ICSBEP criticality benchmarks with fast neutron spectra with excellent results.

An analytic approach to probability tables for the unresolved resonance region

The Unresolved Resonance Region (URR) connects the fast neutron region with the Resolved Resonance Region (RRR). The URR is problematic since resonances are not resolvable experimentally yet the fluctuations in the neutron cross sections play a discernible and technologically important role: the URR in a typical nucleus is in the 100 keV – 2 MeV window where the typical fission spectrum peaks. The URR also represents the transition between R-matrix theory used to described isolated resonances and Hauser-Feshbach theory which accurately describes the average cross sections. In practice, only average or systematic features of the resonances in the URR are known and are tabulated in evaluations in a nuclear data library such as ENDF/B-VII.1. Codes such as AMPX and NJOY can compute the probability distribution of the cross section in the URR under some assumptions using Monte Carlo realizations of sets of resonances. These probability distributions are stored in the so-called PURR tables. In our work, we begin to develop a scheme for computing the covariance of the cross section probability distribution analytically. Our approach offers the possibility of defining the limits of applicability of Hauser-Feshbach theory and suggests a way to calculate PURR tables directly from systematics for nuclei whose RRR is unknown, provided one makes appropriate assumptions about the shape of the cross section probability distribution.

Fission Fragment Angular Distribution measurements of235U and238U at CERN n_TOF facility

Neutron-induced fission cross sections of 238U and 235U are used as standards in the fast neutron region up to 200 MeV. A high accuracy of the standards is relevant to experimentally determine other neutron reaction cross sections. Therefore, the detection effciency should be corrected by using the angular distribution of the fission fragments (FFAD), which are barely known above 20 MeV. In addition, the angular distribution of the fragments produced in the fission of highly excited and deformed nuclei is an important observable to investigate the nuclear fission process.In order to measure the FFAD of neutron-induced reactions, a fission detection setup based on parallel-plate avalanche counters (PPACs) has been developed and successfully used at the CERN-n_TOF facility. In this work, we present the preliminary results on the analysis of new 235 U(n,f) and 238 U(n,f) data in the extended energy range up to 200 MeV compared to the existing experimental data.

Studies of54,56Fe Neutron Scattering Cross Sections

Elastic and inelastic neutron scattering differential cross sections and γ-ray production cross sections have been measured on 54,56 Fe at several incident energies in the fast neutron region between 1.5 and 4.7 MeV. All measurements were completed at the University of Kentucky Accelerator Laboratory (UKAL) using a 7-MV Model CN Van de Graaff accelerator, along with the neutron production and neutron and γ-ray detection systems located there. The facilities at UKAL allow the investigation of both elastic and inelastic scattering with nearly mono-energetic incident neutrons. Time-of-flight techniques were used to detect the scattered neutrons for the differential cross section measurements. The measured cross sections are important for fission reactor applications and also for testing global model calculations such as those found at ENDF, since describing both the elastic and inelastic scattering is important for determining the direct and compound components of the scattering mechanism. The γ-ray production cross sections are used to determine cross sections to unresolved levels in the neutron scattering experiments. Results from our measurements and comparisons to model calculations are presented.

The Neutron Time-of-Flight Cross Section Program at the University of Kentucky - Adventures in Analysis II

Elastic and inelastic neutron differential cross sections are measured at the University of Kentucky Accelerator Laboratory (www.pa.uky.edu/accelerator/) at incident energies in the fast neutron region. The labo- ratorys facilities and instrumentation will be described and our measurement and analysis procedures outlined. Many corrections are required for neutron scattering experiments and the analysis utilizes information from many other cross section data sets and model calculations. Exploring and understanding the limitations of the foundational information and procedures are important for controlling the accuracy of the cross section results. We are examining the limitations in neutron detection efficiency, the normalization of (n,n � γ) cross sections, background reduction, spectrum stripping techniques, and attenuation and multiple scattering corrections. The resulting differential cross sections provide information on the compound elastic and coupled channels reaction mechanisms important for advanced reactor designs.

Uncertainty Study of Nuclear Model Parameters for the n+56Fe Reactions in the Fast Neutron Region below 20 MeV

In this work we study the uncertainty of nuclear model parameters for neutron induced 56Fe reactions in the fast neutron region by using the Total Monte Carlo method. We perform a large number of TALYS runs and compare the calculated results with the experimental data of the cross sections to obtain the uncertainties of the model parameters. Based on the derived uncertainties another 1000 TALYS runs have been performed to create random cross section files. For comparison with the experimental data we calculate a weighted χ2 value for each random file as well as the ENDF/B-VII.1, JEFF-3.1, JENDL-4.0 and CENDL-3.1 data libraries. Furthermore, we investigate the optical model parameters correlation obtained by way of this procedure.

Experimental Uncertainty and Covariance Information in EXFOR Library

Compilation of experimental uncertainty and covariance information in the EXFOR Library is discussed. Following the presentation of a brief history of information provided in the EXFOR Library, the current EXFOR Formats and their limitations are reviewed. Proposed extensions for neutron-induced reaction cross sections in the fast neutron region and resonance region are also presented.

Neutron Cross Section Covariances for Structural Materials and Fission Products

We describe neutron cross section covariances for 78 structural materials and fission products produced for the new US evaluated nuclear reaction library ENDF/B-VII.1. Neutron incident energies cover full range from 10 − 5 eV to 20 MeV and covariances are primarily provided for capture, elastic and inelastic scattering as well as (n,2n). The list of materials follows priorities defined by the Advanced Fuel Cycle Initiative, the major application being data adjustment for advanced fast reactor systems. Thus, in addition to 28 structural materials and 49 fission products, the list includes also 23Na which is important fast reactor coolant. Due to extensive amount of materials, we adopted a variety of methodologies depending on the priority of a specific material. In the resolved resonance region we primarily used resonance parameter uncertainties given in Atlas of Neutron Resonances and either applied the kernel approximation to propagate these uncertainties into cross section uncertainties or resorted to simplified estimates based on integral quantities. For several priority materials we adopted MF32 covariances produced by SAMMY at ORNL, modified by us by adding MF33 covariances to account for systematic uncertainties. In the fast neutron region we resorted to three methods. The most sophisticated was EMPIRE-KALMAN method which combines experimental data from EXFOR library with nuclear reaction modeling and least-squares fitting. The two other methods used simplified estimates, either based on the propagation of nuclear reaction model parameter uncertainties or on a dispersion analysis of central cross section values in recent evaluated data files. All covariances were subject to quality assurance procedures adopted recently by CSEWG. In addition, tools were developed to allow inspection of processed covariances and computed integral quantities, and for comparing these values to data from the Atlas and the astrophysics database KADoNiS.

Issues in Neutron Cross Section Covariances

We review neutron cross section covariances in both the resonance and fast neutron regions with the goal to identify existing issues in evaluation methods and their impact on covariances. We also outline ideas for suitable covariance quality assurance procedures.We show that the topic of covariance data remains controversial, the evaluation methodologies are not fully established and covariances produced by different approaches have unacceptable spread. The main controversy is in very low uncertainties generated by rigorous evaluation methods and much larger uncertainties based on simple estimates from experimental data. Since the evaluators tend to trust the former, while the users tend to trust the latter, this controversy has considerable practical implications. Dedicated effort is needed to arrive at covariance evaluation methods that would resolve this issue and produce results accepted internationally both by evaluators and users.

On the evaluation of 23Na neutron-induced reactions and validations

New recommendations for neutron-induced reactions on 23Na are presented for incident energies from thermal to 200 MeV including resonance parameters up to 1 MeV, and covariance files in the resonance and fast neutron regions. Calculations are based on the nuclear reaction code TALYS and covariances in the fast neutron range are obtained with a Monte Carlo approach. Discrepancies between experimental data and evaluations are presented, indicating needs for new measurements. Validations with integral benchmarks and a simplified model of the Kalimer-600 Sodium Fast Reactor are presented with a special emphasis on the void coefficient, including its uncertainty assessment. Finally, conclusions on the state of knowledge of neutron induced reactions on sodium are presented with recommendations for new measurements.

Extensive Set of Cross-Section Covariance Estimates in the Fast Neutron Region

We generated, for the first time, a very comprehensive set of estimates of cross-section covariance data in the neutron energy range of 5 keV to 20 MeV. The covariance matrices were obtained for 307 materials, from 19F to 209Bi, covering structural materials, fission products, and heavy nonfissile nuclei. These results offer model-based, consistent assessments of covariance data for nuclear criticality safety applications. The evaluation methodology combines the nuclear reaction model code EMPIRE, which calculates the sensitivity of the cross sections to nuclear reaction model parameters, and the Bayesian code KALMAN, which propagates uncertainties of the model parameters to these cross sections. Taking into account the large number of materials studied, we refer only marginally to experimental data. The covariances were derived from the perturbation of several key model parameters selected by the sensitivity analysis. These parameters refer to the optical model potential, the level densities, and the strength of the preequilibrium emission. Our work represents the first attempt to generate neutron cross-section covariances on such a large scale.