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An implementation of both ALICE code and TENDL evaluated nuclear data library in order to describe nuclear reactions induced by low-energy projectiles in the Monte Carlo code MARS15 is presented. Comparisons between results of modeling and experimental data on reaction cross sections and secondary particle distributions are shown.
In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock and Wilcox (B and W) Spectral Shift Control Reactor (SSCR), critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO{sub 2}-ThO{sub 2} fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated. Benchmark criticality results using different cross-section libraries are reported. ENDF/B-VI.6 calculation was run using the S(α, β) model based on ENDF/B-VI.3 data while ENDF/B-VII.0 and ENDF/BVII.1 used an S(α, β) model based on ENDF/B-VII.0 data. JEFF calculations were run using the S(α, β) model based on JEFF 3.1.1 data except for the aluminum isotope, for which the S(α, β) model was missing. The maximum absolute deviation from the benchmark model was obtained using the JEFF-3.1.1 cross sections while closer results were obtained using ENDF/ B-VII.0 and ENDF/B-VII.1. Only ENDF/B-VII.1 and ENDF/B-VII.0 results are inside the 3σ uncertainty band (+2.8σ and 2.4σ, respectively), having deviations on the order of 300 pcm. The maximum deviation is -827 pcm (-6.1σ), using JEFF.3.1.1. Sample calculations of the TDF benchmark were also performed, studying the effects of the main nuclear data libraries. (authors)
With the need for improving existing nuclear data evaluations, (e.g., ENDF/B-VIII.0 and JEFF-3.3 releases) the first step was to evaluate the standards for use in such a library. This new standards evaluation made use of improved experimental data and some developments in the methodology of analysis and evaluation. In addition to the work on the traditional standards, this work produced the extension of some energy ranges and includes new reactions that are called reference cross sections. Since the effort extends beyond the traditional standards, it is called the neutron data standards evaluation. This international effort has produced new evaluations of the following cross section standards: the H(n,n), 6 Li(n,t), 10 B(n,α), 10 B(n,α 1 γ), nat C(n,n), Au(n,γ), 235 U(n,f) and 238 U(n,f). Also in the evaluation process the 238 U(n,γ) and 239 Pu(n,f) cross sections that are not standards were evaluated. Evaluations were also obtained for data that are not traditional standards: the Maxwellian spectrum averaged cross section for the Au(n,γ) cross section at 30 keV; reference cross sections for prompt γ-ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the 252 Cf spontaneous fission neutron spectrum and the 235 U prompt fission neutron spectrum induced by thermal incident neutrons; and the thermal neutron constants. The data and covariance matrices of the uncertainties were obtained directly from the evaluation procedure.
Overview of the Nuclear Data Management and Analysis System detailed as: Overview of NDMAS System, NDMAS NQA-1 process, Data in NDMAS, and Future work and new additions.
An overview of the NJOY Nuclear Data Processing Code System, including recent history, current status and near-term upgrade plans is presented.
Talk for the National Nuclear Data Center at Brookhaven National Laboratory, November 19, 2020
This report documents an evaluation of 235 U prompt fission neutron spectra (PFNS) that is a release candidate for the upcoming U.S. nuclear data library, ENDF/B-VIII.1. This evaluation differs from its predecessor, ENDF/B-VIII.0, mainly by the inclusion of 235 U PFNS measured by the Chi-Nu team of LANL and LLNL. This data set is the first one that covers the 235 U PFNS for continuous incident-neutron energies of 1⁻20 MeV and outgoing-neutron energies from 10 keV⁻10 MeV with high precision. Previous data sets were either measured in a limited energy range or with less precision. Hence, these new Chi-Nu data provide decisive information for the evaluation. The resulting evaluated data correspond well to the new experimental PFNS. The evaluated PFNS also produce average mean energies and 239 Pu/ 235 U PFNS in agreement with associated Chi-Nu data. If one uses the new evaluated data to predict the neutron multiplication factor, k eff , of selected ICSBEP critical assemblies, the differences of simulated values compared to those using ENDF/B-VIII.0 is modest (less than 55 pcm). This difference in k eff can be easily accommodated by changes in the 235 U average prompt fission neutron multiplicity that is currently being re-evaluated. In addition to that, the new PFNS predict on average 235 U LLNL pulsed-sphere neutron-leakage spectra better than ENDF/B-VIIII.0 PFNS.
The Los Alamos Monte Carlo transport code MCNP6 has recently benefitted from significant work to improve the photon-electron transport and atomic-relaxation capabilities of the code. The improvements include extension of existing photon data to lower energies than previously supported; addition of new data such as subshell-specific photoelectric cross sections; completion of the incoherent and coherent form factor data to cover all scattering angles at all supported energies; new data for electron interactions including Bremsstrahlung, elastic scattering, subshell-specific electroionization, and atomic excitation; a much more detailed representation for atomic relaxation; and an entirely new (to MCNP) approach to electron transport: the single-event method. These improvements were made available with the initial production release of MCNP6 along with an extended electron, photon, and relaxation data file eprdata12 to support the new capabilities. More recently new data have become available and have now been incorporated into MCNP6, correcting a few issues with the previous work, and adding new sets of cross sections that were not included in eprdata12. The purpose of this paper is to introduce these newer changes and enhancements, which will be included with the next public release of the MCNP6 code system. The database that supports these developments relies on an evolving set of libraries originally released by D. E. Cullen and collaborators as separate data for photo-atomic interactions (EPDL), atomic relaxation (EADL), and electro-atomic interactions (EEDL). An updated version of this release was officially adopted in the ENDF/B VI.8 release of the Evaluated Nuclear Data File and it was this source that was used to produce the ACE (A Compact ENDF) format library eprdata12 for use in the initial production release of MCNP6.1 (and the subsequent interim release MCNP6.1.1). The latest improved version of this database was released by Cullen in 2014. This is the EPICS2014 collection, which includes both an ENDF-formatted file and separate EPDL, EADL, and EEDL files. In the area of photon-electron transport and atomic-relaxation, the newest enhancements to the current version of MCNP6 (scheduled for release as MCNP6.2) are primarily based on improvements and additions drawn from the EPICS2014 data.
This document constitutes the final report for the “Improving Nuclear Data for Short Lived Actinides” research project. During this project 29 papers were published (or are about to be) in peer reviewed journals and conference proceedings which document the research and results extensively. The publications are listed below. Numerous invited and contributed talks were given at national and international nuclear physics meetings. The data collected during this project has in large part made it into the relevant databases for use by the customers. Numerous undergraduate, graduate and post-doctoral researchers were trained in the area of nuclear physics from over 12 different institutions and multiple countries during this research effort. That is reflected in the names and affiliations of the co-authors in the journal articles listed below.
The pursuit for a soluble-boron-free (SBF) PWR core has been on-going for more than three decades. It is known that feasibility of such a design is generally higher as the core power (hence core size) is reduced. It is upon this observation that a new approach for an SBF small modular reactor (SMR) core was recently proposed. This paper aims to present criticality control management of the said investigation. All calculations in this study were completed using the 2-step Serpent /Coredax reactor analysis with ENDF/B-7.1 nuclear data library. (authors)
Monte-Carlo codes simulating particle transport usually make use of well-known variance reduction techniques, leading to more efficient simulations. Even in configurations which do not deal with deep-penetration, some of these techniques, such as the implicit capture, are actually often used by default. The statistical weight of the simulated particles is then modified in order to take into account the modification of the sampling without introducing any bias in the estimation process. For instance, the implicit capture decreases the statistical weight of the particles at each collision, whereas the sampling of a single outgoing particle increases this statistical weight in case of scattering reactions producing multiple neutrons. In addition, Russian roulette and splitting operations, which are usually also played at each collision, lead to other modifications of the statistical weights. As a result, Monte-Carlo codes provide particle histories that are valid on average, but collect statistical weights which differ from 1, when tallying the estimation of the quantities of interest. Yet, some nuclear instrumentation calculations inherently need to work with statistical weights equal to 1, since either individual events have to be collected (such as the arrival of a particle at a detector, the measurement of a time of flight, etc), or the coincidence of events or also correlations between individual events have to be taken into account. To address these specific applications with Monte-Carlo simulation codes, users need a full analog simulation mode. The continuous-energy Monte-Carlo code TRIPOLI4{sup R} is mainly dedicated to shielding, reactor physics and criticality safety calculations. Its capabilities have recently been extended to widen its use to nuclear instrumentation with the possibility of γ-spectrometry calculations and the simulation of photonuclear reactions for instance. Other features needed by nuclear instrumentation applications have been available in TRIPOLI-4 for a long time, such as the sampling of the time variable and the possibility of tallying in a time grid for most of the available tallies (fluxes, reaction rates, dose equivalent rates, energy depositions, etc). In TRIPOLI-4, the transport between two consecutive collisions is analog if no variance reduction technique is requested by the user to handle deep-penetration configurations. However, up to the version 9 of the code, the collision itself is sampled in a non-analog way. This paper presents a first implementation of a full analog neutron transport mode in TRIPOLI-4, preliminary verification results and an application of this functionality to the simulation of the Nucifer detector located in Saclay, France. It should be noted that only neutron transport is dealt with in this work and that all nuclear data come from JEFF-3.1.1 or ENDF/B-VII.0 evaluations, including fission data when needed. (authors)
Neutron cross section measurements and data files have advanced significantly since 1989 when Los Alamos published 'A Nuclear Cross Section Data Handbook', a detailed compendium of point-wise cross section plots and pie charts of relative contributions of individual reactions to the total cross section for 129 isotopes. Other compendiums with similar data and analysis tools have also been published. The intent of this work is to develop a modern, open-source application for interactive exploration and analysis of Evaluated Nuclear Data File (ENDF) libraries that also allows users to create their own 'Nuclear Cross Section Data Handbooks'. EXSAN is an easy-to-use graphical user interface application for exploring ENDF photon and neutron cross sections developed in Python. In addition to interactive plotting, EXSAN includes methods for computing rank-order lists of cross section for any energy and calculating composite cross sections for isotopic and elemental compositions and additional batch automation features. (authors)
Highlights: • Compilation of experimental transition probabilities or B(E2) values for the known first 2{sup +} states in even–even nuclei. • Evaluation of B(E2) values, lifetimes and deformation parameters for 447 even–even nuclei. • Extensive discussion on data evaluation policies and procedures. • Nuclear shell model calculation of the first 2{sup +} states energies and B(E2) values in even–even nuclei. Experimental results of E2 transition probabilities or B(E2) values for the known first 2{sup +} states in 447 even–even nuclei have been compiled and evaluated. The evaluation policies for the analysis of experimental data have been described and new results are discussed. The recommended B(E2) values have been compared with comprehensive shell model calculations for a selected set of nuclei, where such theoretical procedures are amenable. The present work was motivated by a rapid increase in the number of new B(E2) measurements for the first 2{sup +} states since the previous evaluation of such data by S. Raman et al. published in 2001. Future plans to investigate the systematics of B(E2)↑ values, and intercomparison of different experimental techniques to obtain these data are outlined.
The Nuclear and Energy Research Institute - IPEN, offers post-graduate programs, namely: Nuclear Technology - Applications (TNA), Nuclear Technology - Materials (TNM), Nuclear Technology - Reactors (TNR). The Institute programs mission is to form expert technicians, physicists and engineers with a strong knowledge in their discipline to work in the nuclear area. The course: 'Theoretical Fundamentals and Practices of the Instrumentation used in Nuclear Data Acquisition' covers the use of laboratory nuclear instrumentation and the accomplishment of experiments to obtain nuclear parameters. One of these experimental exercises is object of this work: 'Study of influence of plastic scintillators to detect Beta particles and Gamma radiation by means of spectral analysis of {sup 90}Sr, {sup 90}Y and {sup 137}Cs sources'. The use of scintillators plastic for the detection has the advantage of low cost, high mechanical strength, is not hygroscopic and can be manufactured in large volumes. This work aims to present the analysis of relative efficiency of detection of plastic scintillators of various thicknesses for beta particles and gamma radiation by the spectrum of {sup 137}Cs and {sup 90}Sr. Due to lack of resolution of the detectors plastic scintillators we worked with relative efficiency. The evaluation was done by reading deposited energy, using the software MAESTRO, for each detector thickness. For beta particles was observed an ideal thickness around 3 mm and the better photon efficiency was observed with increasing the thickness of the detector. The present experiment does not intend to establish a new technique for this subject: it solely aims student's practical exercises in nuclear properties of elements and detectors being part of the nuclear experimental course. (authors)
The Nuclear Data Team has released updated continuous-energy neutron data files for: 1) hydrogen, and 2) S (α; β) (thermal scattering) on SiO 2 . A list of new ZAIDs and the data that is updated (Old ZAID) is given in Table 1. The old data are still accessible, but are not the default.
Current experimental fusion systems and conceptual designs of fusion pilot plants (FPPs) are growing in complexity and size. Several radiation metrics are crucial to the safe operation of fusion machines, including neutron flux streaming through openings and the shutdown dose rate (SDDR). Most current designs of advanced experimental fusion systems—and the most probable candidates for FPPs—are based on the tokamak concept, which is prone to neutron streaming through the myriad openings needed for diagnostic and support systems. SDDR is caused by decay gamma rays from radionuclides that become activated by neutrons during the operation of a fusion system that use deuterium-deuterium (DD), tritium-tritium, or deuterium-tritium plasma. Because computational tools have become essential for determining these radiation metrics, they must be validated against reliable and applicable experimental data. Experiments at the Joint European Torus (JET) provide a unique source of experimental data for validating computational tools and nuclear data used to determine SDDR and neutron fluxes in streaming-dominated geometries. Here, this paper presents the comprehensive analysis of the high-performance DD JET SDDR, and streaming experiments performed using Oak Ridge National Laboratory (ORNL) fusion workflows. The computational results were compared with experimental results that consist of online SDDR measurements with ionization chambers and neutron fluence streaming measurements using thermoluminescent detectors. The ratio of calculated-to-experimental SDDR values ranges from 0.6 to 2.5, and the streaming results range from 0.5 to 8.0. Future work will include analyzing the JET 2021 DTE2 campaign alongside the integration of the Shift Monte Carlo transport code into all ORNL fusion neutronics workflows.
The Organization for Economic Cooperation and Development (OECD) Working Party on Nucelar Criticality Safety (WPNCS) has proposed a benchmark exercise representative of neutronic behavior in criticality experiments. Here, the goal is to develop confidence in data assimilation techniques used to adjust nuclear data. Participants are given synthetic experimental models with associated measured data and asked to estimate the model parameters given the model and measurements as well as provide predictions for separate application models. In this work, we performed data assimilation using Bayesian inverse Uncertainty Quantification (UQ) with machine learning surrogate models to produce posterior parameter distributions for the requested parameters and posterior predictive distributions for the requested responses. Several experimental models are shown to insufficiently inform the posterior parameter distributions for the applications involved. However, given sufficient experimental data, posterior parameter estimates yielded reduced uncertainty in the response predictions of interest while covering the experimental data.
A summary of nuclear data evaluation for the 39,41 K + γ reaction is given, which focuses on the production of radio-active isotopes. The statistical model calculation is performed with the CoH 3 code, and the calculated 39 K( γ,n) 38m K cross section is compared with the available experimental data.