Neutron Cross Section Data Research Articles

Verification of shielding calculation capability of cosRMC with SINBAD fusion benchmarks

In the design of fusion reactors, extensive shielding calculations are necessary to verify the feasibility and safety of the design. The issue of deep penetration is a common and significant problem in fusion reactor shielding calculations. When the thickness of the shielding layer exceeds a certain value, the statistical error in Monte Carlo method calculations significantly increases. Therefore, it is imperative to adopt variance reduction techniques to enhance the computational accuracy for deep penetration problems. cosRMC is a reactor Monte Carlo code, and its accuracy in calculating neutron-deep-penetration shielding problems should be verified. For this purpose, three deep penetration benchmarks from Shielding Integral Benchmark Archive Database (SINBAD) were simulated by cosRMC and neutron cross section data library ENDF/B-VII.0. These benchmarks describe the penetration behaviors of neutrons in models with different geometric shapes and materials. This paper shows detailed cosRMC models based on benchmark experimental models, used to calculate neutron flux spectra at different penetration depths, which are recorded by detectors in the related benchmark experiments. In this way, a direct comparison between cosRMC results and experimental measurements is provided. In order to improve the accuracy of cosRMC in calculating deep penetration problems, the variance reduction technique of cell importance was used. The neutron flux spectrum calculated by cosRMC was compared with the neutron flux spectrum results calculated by MCNP5 and the experimental values of the benchmarks. The comparison results show that the calculations of cosRMC are in compliance with the MCNP5 reference values and experimental values, and the relative deviations between calculations of cosRMC and MCNP5 is within three time of standard errors. Therefore, this paper has verified that cosRMC can applies to neutron deep penetration problems of fusion reactors.

The influence of Monte Carlo radiation transport codes and hadron physics models on stray neutron dose estimations in proton therapy: An extended intercomparison

PurposeSelecting a Monte Carlo (MC) radiation transport code and nuclear physics model affects the simulation of secondary neutron fluence energy spectra and the estimation of the neutron dose equivalent in proton therapy. This study compares the MC codes FLUKA, MCNPX and GEANT4 from previous work with two additional MC codes, TOPAS (with either the Binary Intra-nuclear Cascade (BIC) or the Bertini intra-nuclear cascade model (Bert)) and PHITS (with either the Intra-Nuclear Cascade of Liege (INCL) or the Bert model), to facilitate a more comprehensive comparison. Material and methodsSecondary neutron fluence energy spectra were scored at different locations inside a water phantom, exposed to a 10 × 10 cm2 parallel mono-energetic proton beam with energies of 110 MeV, 150 MeV, 180 MeV and 210 MeV. The neutron dose equivalent was estimated by applying fluence-to-dose equivalent conversion coefficients. ResultsDiscrepancies between MC codes and physics models reach 90 % for the total neutron fluence. Differences are most pronounced in the forward direction of the beam due to a notable presence of high-energy neutrons and the higher uncertainty on the neutron cross-section data for these neutrons. Nuclear model selection within PHITS (INCL or Bert) and TOPAS (BIC or Bert) shows a worse agreement for the Bert model than for the other MC codes and nuclear models. The discrepancies in the neutron spectrum simulations propagate into differences in the estimated neutron dose equivalents. They are more prominent for distal positions (up to a factor of 3) than for lateral positions (up to a factor of 2) and decrease at higher proton energies. ConclusionsThis study shows considerable differences in the out-of-field neutron fluence energy spectra calculated by different MC codes and nuclear physics models. Differences in the neutron dose equivalent reach a factor of 3. The discrepancies are most apparent in distal positions and at lower proton energies.

Evaluation of Neutron Cross-Section Data Using Gaussian Mixture Model and Digital Filter

For the accurate estimation of neutron cross-section data, evaluation of nuclear data is mandatory to fulfill the need of nuclear science and technology. In this work, the evaluation of 232Th(n,2n)231Th, 241Am(n,2n)240Am, 100Mo(n,2n)99Mo, and 96Mo(n,p)96Nb reaction cross-section data is carried out using a novel method. This novel method of evaluation employs a cluster-based piecewise evaluation followed by a digital filter for merging the evaluated curves. The clusters in the experimental data and model data are identified using the probabilistic method of the Gaussian Mixture Model. The clustered experimental data are then regressed using the polynomial regression technique. The model data are generated using the Talys 1.9 code, and the model deficiency due to the complex random nature of nuclear reactions is also accounted here using chi-squared analysis. Evaluation in each cluster is then carried out independently using the popular Kalman filter technique. For obtaining a single smooth evaluated curve for the whole energy range, the popular smoothing digital filter, the Savitzky-Golay Filter, is employed for the first time in nuclear data evaluation. The proposed evaluated curves and existing evaluated curves of 232Th(n,2n)231Th, 241Am(n,2n)240Am, 100Mo(n,2n)99Mo, and 96Mo(n,p)96Nb reactions from nuclear data libraries such as ENDF/BVIII.0, JEFF-3.3, JENDL-4.0, CENDL-3.1, and TENDL 2021 are compared and found to be in good agreement. It is also found that generally, evaluation methods are data dependent, and so, a single evaluation method may not be applicable for all reactions of all nuclides. Since piecewise evaluation is cluster dependent, selecting the appropriate cluster makes this method robust for almost all reactions of all nuclides. Also, it is proven that this novel method of evaluation is a promising method demonstrating the potential of this approach for evaluation based on the chi-squared goodness-of-fit test with respect to standard evaluated library ENDF/BVIII.0 and experimental data.

Development of a new Monte-Carlo neutron transport code for radiation shielding

A continuous energy Monte-Carlo method-based neutron transport code for fixed source in 3D solid geometry has been developed. The main objective is to integrate this code with recently developed Monte-Carlo gamma shielding code MCBLD, so as to build an in-house code for coupled neutron-gamma shielding study. Also it is intended to use of the new ENDF data library (ENDF/B-VIII) in this code. This paper gives an overview of the code in terms of processing neutron interaction data and numerical scheme to track scattered neutrons. Present capability of the code is demonstrated by presenting fluence due to mono-energetic neutron source as well as widely used spontaneous neutron source from 252Cf isotope, placed at the centre of different size iron spheres. Results are also generated from general purpose open-source Monte-Carlo code FLUKA. Overall good agreement between the two validates the numerical implementation of angle and energy dependent neutron cross-section data in this code.

Methodology for physics-informed generation of synthetic neutron time-of-flight measurement data

Accurate neutron cross section data are a vital input to the simulation of nuclear systems for a wide range of applications from energy production to national security. The evaluation of experimental data is a key step in producing accurate cross sections. There is a widely recognized lack of reproducibility in the evaluation process due to its artisanal nature and therefore there is a call for improvement within the nuclear data community. This can be realized by automating/standardizing viable parts of the process, namely, parameter estimation by fitting theoretical models to experimental data. There are numerous candidate methods to approach this type of problem, but many rely on large, labeled datasets that are not accessible to the nuclear data evaluator. For a reaction cross-section, there are usually just a handful of datasets, none of which can be considered labeled because evaluators never have access to the exact solution (cross section). This work leverages problem-specific physics, Monte Carlo sampling, and a general methodology for data synthesis to generate unlimited, labeled experimental cross-section data of high-utility. The synthesized data is said to be of high-utility because it is statistically similar to the observed data. Heuristic and, where applicable, rigorous statistical comparisons to observed data support this claim. The methodology is split into two generative models. The first generates a realization of an energy-differential cross section for a given isotope. The second takes the output from the first as a determined input and generates noisy experimental observables (radiation detector signals) from the determined cross section realization. The latter is the primary development of this article and is based/limited to transmission measurements at Rensselaer Polytechnic Institute (RPI). The former leverages an existing method for model parameter sampling in the resolved resonance region (RRR), thus limiting the current demonstration to the RRR of incident neutron energies. An open-source software is published alongside this article that executes the complete methodology to produce high-utility synthetic datasets. The goal of this work is to provide an approach and corresponding tool that will allow the evaluation community to begin exploring more data-driven, ML-based solutions to long-standing challenges in the field.

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.

Automated resonance evaluation; Non-convex decomposition method for resonance regression and uncertainty quantification

This work serves as a proof of concept for an automated tool to assist in the evaluation of experimental neutron cross section data in the resolved resonance range. The resonance characterization problem is posed as a mixed integer nonlinear program (MINLP). Since the number of resonances present is unknown, the model must be able to be determine the number of parameters to properly characterize the cross section curve as well as calculate the appropriate values for those parameters. Due to the size of the problem and the nonconvex nature of the parameterization, the optimization formulation is too difficult to solve as whole. A novel method is developed to decompose the problem into smaller, solvable windows and then stitch them back together via parameter-cardinality and parameter-value agreement routines in order to achieve a global solution. A version of quantile regression is used to provide an uncertainty estimate on the suggested cross section that is appropriate with respect to the experimental data. The results demonstrate the model's ability to find the proper number of resonances, appropriate average values for the parameters, and an uncertainty estimation that is directly reflective of the experimental conditions. The use of synthetic data allows access to the solution, this is leveraged to build-up performance statistics and map the uncertainty driven by the experimental data to an uncertainty on the true cross section.

Application of Kalman filtering technique for evaluation of neutron cross section data of [formula omitted]Mo (n, 2n) [formula omitted]Mo reaction

In this paper, for the first time, evaluation of neutron cross section data of 100Mo (n, 2n) 99Mo reaction is performed using experimental data available in IAEA-EXFOR database library and nuclear model-based data generated using Talys 1.9 code by applying a novel method of combining Kalman filtering technique with Machine Learning (ML) regression algorithms. The neutron cross section data evaluation has been performed after a detailed study of all the EXFOR papers corresponding to 100Mo (n, 2n) 99Mo reaction and nuclear model based cross section generation by executing T6 random input files using Talys 1.9 code. The evaluated curve generated is then compared with the existing evaluated curve of 100Mo (n, 2n) 99Mo reaction from nuclear data libraries such as ENDF/B-VIII.0, JEFF-3.3, JENDL-4.0, CENDL-3.1 and TENDL 2017 and found to be in good agreement with them. Chi-square and generalized Chi-square tests were employed to assess the proposed evaluation techniques and found them to be good in estimating evaluated mean values and evaluated uncertainties of cross section.

Neutron spectrum determination of accelerator-driven d(10)+Be neutron source using the multi-foil activation technique

New neutron field of the compact accelerator-driven fast neutron source based on the cyclotron and beryllium target station was studied and developed at the NPI CAS. Neutrons were generated using the deuteron bombardment on 8 mm thick beryllium layer by the 10.3 MeV deuteron beam. Broad neutron energy spectra of the d(10)+Be source reaction at two source-to-sample distances (14 mm and 154 mm) were determined for the first time at the NPI. Sets of eight activation materials comprising Au, Co, Ti, In, Al, Fe, Ni, and Nb were irradiated in the d(10)+Be neutron field, and activated dosimetry foils were analysed using the nuclear gamma-spectroscopy technique (HPGe detector) and reaction rates were measured. Neutron energy spectra of the d + Be source reaction for 10.3 MeV deuterons were reconstructed employing the modified version of the SAND-II unfolding code. For 9.7 μA deuteron beam, the fast neutron flux of developed d(10)+Be neutron field reached the value of 1.4 × 1010 cm−2s−1 at a distance of 14 mm from the source target and 4.1 × 108 cm−2s−1 at a distance of 154 mm. The novel d(10)+Be neutron field with energy range up to 15 MeV is convenient for neutron cross-section data validation, experimental simulation of the fast neutron spectrum of experimental nuclear reactors, applications of neutron activation analysis, and radiation hardness tests of materials important for nuclear energetics.

Estimations of (n, p) reaction cross-sections at 14.5 MeV incident neutron energy by artificial neural networks

The cross-sections for (n, p) reactions around 14 MeV are important for nuclear research field and applications of nuclear energy. Also these neutron-induced reactions have been extensively studied in the literature. In this study, we have theoretically estimated the (n, p) reaction cross-section of different elements for 14.5 MeV incident neutron energy by using artificial neural network (ANN) methods whose experimental data are not available in the literature. For the construction of the ANN, available experimental data have been used. The results from the constructed ANN have been compared with the experimental data and those obtained from the theoretical formula. Furthermore, by using the constructed ANN through the limited experimental data, the new cross-section data for some (n, p) reactions have been generated and the results have been compared with a common database. The results give a good agreement with the literature and indicate that the method can be a powerful tool for the estimation of neutron cross-section data for the neutron-induced reaction.

Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Component-Critical Experiments

A series of critical experiments were conducted at the National Criticality Experiments Research Center (NCERC) in Nevada to evaluate the operational performance of a compact reactor that eventually will resemble the flight unit the National Aeronautics and Space Administration will use for deep space exploration. The results from the experiments are compared to preliminary results from computational models using MCNP and ENDF/B-7.1 neutron cross-section data.

Shut-down dose rate analysis for the activated target assembly in IFMIF-DONES during maintenance

• The activation and shutdown dose analyses of the IFMIF-DONES. • Transport and activation calculations using McDeLicious code. • The dose rate maps in the access cell as well as the shielding assessments is presented. • The analysis of the shielding for activated target of DONES. This paper presents the activation and shutdown dose analyses of the IFMIF-DONES (International Fusion Materials Irradiation Facility – DEMO-Oriented NEutron Source) Target assembly. To this end, a series of coupled transport and activation calculations are performed using McDeLicious code, which is based on the MCNP Monte Carlo code for simulating the neutron production from d-Li reactions, and the FISPACT inventory code. The latest DONES test cell model has been employed and FENDL neutron cross-section data library has been used. Shut-down dose rate calculations were performed using the R2Smesh code system developed at KIT. The obtained dose rate maps in the access cell as well as the shielding assessments will be presented in the paper.

Effects of different nuclear evaluation data on the RMC keff calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

Simulation of neutron transmission performance of metallic spherical shell under temperature dependent neutron cross section

To study the effect of temperature dependent neutron cross section on the neutron transmission performance of typical metallic spherical shells, this paper generated the temperature dependent neutron cross section data by using NJOY program, and established the computational model of interaction of neutron with iron and aluminium spherical shell. The temperature evolution of energy spectrum of transmitted neutron and secondary gamma ray, and the number ratio of transmitted neutrons to secondary gamma rays has been studied at various neutron energies. The results show that only when the energies of the incident neutron are in a certain range, the temperature dependent cross sections can have a significant effect on the energy spectrum of the rays penetrating the spherical shell. For different metal materials, the transmitted neutron spectrum and the secondary gamma ray spectrum will be significantly affected when the temperature increases to different high value. Generally, the transmitted neutron number decreases with the increase of temperature, and the number ratio of transmitted neutron to secondary gamma ray also decreases with the increase of temperature, while the changing trend of number of secondary gamma ray with temperature is related to the type of shell material.

New evaluation of general purpose neutron data for stable W-isotopes up to 200 MeV

In the frame of the Power Plant Physics and Technology of EUROfusion, new evaluations of general purpose neutron cross-section data were performed for the 180,182,183,184,186W isotopes covering the neutron energy up to 200 MeV. A special version of the TALYS nuclear model code implementing the geometry dependent hybrid model supplied with models for the non-equilibrium cluster emission was applied for calculations of the nuclide production and the energy distribution of the emitted particles. The parameters of the GDH model were properly estimated using measured data for individual tungsten isotopes. The neutron cross-sections were evaluated making use of available experimental data, systematics including estimated A-dependence of components of gas production cross-sections, and covariance information produced as part of the evaluation process. The BEKED code package, developed at KIT, was applied for calculations of co-variances using a dedicated Monte Carlo method. The evaluated data were processed into standard ENDF data format using the TEFAL code and the FOX module of the BEKED system. The evaluated data files were checked for errors and inconsistencies, processed with the NJOY code into ACE data format, and benchmarked against available integral experiments with MCNP neutron transport calculations.

Distinction between super-cooled water and ice with high duty cycle time-of-flight neutron imaging

We report on measured neutron cross section data for super-cooled water and ice by time-of-flight (TOF) neutron transmission imaging. In particular, we demonstrate the use of high duty cycle (HDC)-TOF measurements to determine the local aggregate state of water with spatial resolution, by exploiting the neutron cross section dependence on the mobility of hydrogen atoms for long neutron wavelengths (>4 Å). While one can envision many different applications for this method, one example is to provide insights into the freezing mechanism during the start-up of polymer electrolyte fuel cells from below zero degrees. Unlike for other wavelength selective measurements (e.g., Bragg edge imaging), only a limited wavelength resolution is required for this method. With a chopper setup with HDC (30%), we reached a high contrast-to-noise ratio (CNR) with a contrast between ice and super-cooled water of 8%. To maximize the CNR, we optimized the duty cycle, pulse period, and image processing parameters. Moreover, we present a theoretical framework for performing such optimization calculations, which can be used to maximize CNR for any beam line and any substances. For the optimization procedure presented in this publication, we used cross section values for ice and super-cooled water measured with high wavelength resolution using wavelength frame multiplication choppers. Our results show that the aggregate state of water of a sufficiently thick layer of water (>0.5 mm) can be reliably determined for a small area (1 mm2) and with a reasonable short acquisition time of 5 min.

An overview of the development of a multigroup Monte Carlo code for TRIGA reactors

This paper discusses the overview of the development of TRIMON, a next-generation core management code for TRIGA reactors. The newly developed code employs Monte Carlo method which uses diffusion theory type homogenized neutron cross section data, thus, speeding up the stochastic simulation via simplification of a complex core geometry. It has the capability of running three-dimensional core simulation up to 16 neutron energy groups. In general, the group neutron cross sections are pre-processed and homogenized via the Effective Diffusion Homogenization (EDH) method. The important features of TRIMON include core eigenvalue calculation, fuel burnup management, three-dimensional power and neutron flux distribution prediction. In this paper, the numerical results obtained using TRIMON were compared with the actual measured operational data. The performance comparison was also made between TRIMON and MCNP.

Measurement of neutron scattering cross section of nano-diamond with particle diameter of approximately 5 nm in energy range of 0.2 meV to 100 meV

Abstract A nano-diamond is an attractive neutron reflection material below cold neutron energy. The total neutron cross section of a nano-diamond produced using the detonation explosive technique was derived from a neutron transmission measurement over the neutron energy range of 0.2 meV to 100 meV because total neutron cross section data were not available. The total cross section of a nano-diamond with particle size of approximately 5 nm increased as the neutron energy decreased to 0.2 meV. It was approximately two orders of magnitude larger than that of carbon at 0.2 meV. The contribution of inelastic scattering to the total cross section was shown to be negligibly small at neutron energies of 1.2, 1.5, 1.9, 2.6, and 5.9 meV in inelastic neutron scattering measurements. Moreover, small-angle neutron scattering measurements of the nano-diamond showed a large scattering cross section in the forward direction at low neutron energies.

Doppler broadening of neutron-induced resonances using ab initio phonon spectrum

Neutron resonances observed in neutron cross section data can only be compared with their theoretical analogues after a correct broadening of the resonance widths. This broadening is usually carried out by two different theoretical models, namely the Free Gas Model and the Crystal Lattice Model, which, however, are only applicable under certain assumptions. Here, we use neutron transmission experiments on UO2 samples at $T=23.7$ K and $T=293.7$ K, to investigate the limitations of these models when an ab initio phonon spectrum is introduced in the calculations. Comparisons of the experimental and theoretical transmissions highlight the underestimation of the energy transferred at low temperature and its impact on the accurate determination of the radiation widths $\Gamma_{\gamma_{\lambda}}$ of the 238U resonances $\lambda$ . The observed deficiency of the model represents an experimental evidence that the Debye-Waller factor is not correctly calculated at low temperature near the Neel temperature ( $ T_{N}=30.8$ K).

Investigation of novel moderator geometries at the Necsa accelerator based neutron sources

Investigations into novel neutron moderator geometries yielding significant yield enhancements is a challenging exercise when trying to model possible structures, primarily due to a lack in suitable cross-section data, as a function of neutron energy, for various materials at cryogenic temperatures. Current studies at the small accelerator based neutron sources based at Necsa, South Africa, centre around investigations into the effect of unusual geometries for the primary moderator. Due to the inherently low sub-thermal neutron yield of small accelerator based neutron sources, further studies of neutron moderator and reflector configurations that yield a directional bias have been pursued, specifically for a multi-layered polyethylene and silicon wafer concept, maintained at cryogenic temperatures. The status of these investigations, including comparison between experiment and MCNP modelling are presented. Identification of particular deficiencies in neutron cross-section data will also be presented.