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ENDFtk is a recently developed C++ and Python interface to interact with ENDF-6 formatted nuclear data files. It provides a robust and complete interface, allowing the reading and writing of all formats currently part of the ENDF-6 formats manual, as well as some non-ENDF formats used by the NJOY processing code. It provides an interface that mimics the names in the ENDF-6 formats manual as well as an equivalent interface using human-readable attribute names. It is robust and powerful enogh for nuclear data experts to develop complex applications, while also simple enough to be used non-experts to retrieve and manipulate evaluated nuclear data. ENDFtk offers the ability to easily interrogate and manipulate data either in large-scale code projects or in simple Python scripts. In this paper, a brief overview of the interface is given, as well as more substantial examples demonstrating plotting simple data, interacting with more complex data, and writing new data to files. ENDFtk is open source and available for download via GitHub (https://github.com/njoy/ENDFtk).

Values on cumulative fission yields for the Xenon isotopes ¹³¹Xe, ^131mXe, ¹³³Xe, ^133mXe, ¹³⁵Xe and ^135mXe were compiled from the nuclear data libraries ENDF/B-VIII.0, JEFF-3.3 and JENDL-4.0, available via the Nuclear Data Services of the the International Atomic Energy Agency. This dataset lists the collected values and their uncertainties as reported.

Values on cumulative fission yields for the Xenon isotopes ¹³¹Xe, ^131mXe, ¹³³Xe, ^133mXe, ¹³⁵Xe and ^135mXe were compiled from the nuclear data libraries ENDF/B-VIII.0, GEFY-8.1, JEFF-3.3 and JENDL-4.0, partly available via the Nuclear Data Services of the the International Atomic Energy Agency. This dataset lists the collected values and their uncertainties as reported.

Abstract This paper presents a suggestion of a Nuclear Physics teaching activity in which the International Atomic Energy Agency (IAEA) nuclear data are used. The motivating theme is the understanding of the Physics in Boron Neutron Capture Therapy (BNCT). Neutron-Induced fission of B-10, in 5 10 B ( n , α ) 3 7 Li reaction, and gamma decay has been analyzed and discussed in this context. To understand the conceptual aspects of reaction energy, fission products – divided into two different energetic sets of particles –, and gamma radiation from the radioactive decay of the Li-7 isomeric state were used both qualitative and quantitative methodologies.

To satisfy the increasing needs of nuclear data from diversified nuclear applications, the Neutron Activation Cross Section Data Library has been developed at China Nuclear Data Center (CNDC) in cooperation with China Nuclear Data Coordination Network. It contains calculated and evaluated cross sections of nuclear reactions for 818 stable and unstable target nuclei including isomeric targets nuclei from 1H to 257Fm are included in this library in neutron energy region of 10-5 eV-20 MeV. The ENDF/B-6 data format was adopted. The general information, comments (MF=1), reactions cross sections (MF=3), nucleus dictionary(MF=8) and split threshold reaction channels (MF=10) were included in the library. The reaction cross sections were obtained using UNF and FDRR model calculations or systematic analysis based on available experimental data.

This work investigates the quality of the ENDF 56Fe cross-section libraries for describing the transport of fast neutrons in iron. We have used the D(d,n)3He reaction with a pulsed 7-MeV deuteron beam energy as a neutron source and analyzed the neutrons transmitted through two natural iron spheres of thicknesses 3 and 8 cm. The experimental neutron time-of-flight transmitted spectra for various angles are compared with MCNP simulations. Our result indicates the possibility of an underestimation of the nonelastic cross section and an overestimation of the elastic cross section for 56Fe in the ENDF/B-VII.1 library for the neutron energy range of 7.2 to 10.2 MeV. Our result agrees qualitatively with the Ramsauer model and optical model calculations. This discrepancy in the library cross section might lead to an underestimation/overestimation of material damage in nuclear reactor calculations. A newer evaluation, ENDF/B-VIII.0, was released subsequent to the completion of the majority of this project. The new evaluation has a decreased elastic cross section and an increased inelastic cross section for 56Fe in our energy range of interest, which agrees qualitatively with our result.

The experimental thermal neutron cross sections are important in nuclear data validation. The Evaluated Nuclear Data File (ENDF) library is a comprehensive repository of evaluated nuclear reaction data, including thermal neutron cross sections (thermal scattering laws, TSLs), for a wide range of materials. ENDF libraries importance lies in providing critical data for nuclear science and engineering applications, such as reactor design, nuclear medicine, and safety analysis, where accurate thermal neutron cross sections are essential for modeling neutron interactions and behaviors in various materials and environments. There are some known issues with graphite and polytethylene TSLs in the ENDF/B-VIII.0 and ENDF/B-VIII.1 releases, hence we would like to perform some transmission (total cross section) measurements to rectify these issues. We would also like to use these measurements on graphite to characterize the impact of small angle neutron scattering on the total cross section, as well as to validate the models used to calculate this contribution.

This paper details and implements a framework for evaluating thermal neutron scattering cross sections that provide S(α,β) data and covariance data for hydrogen in light water. This methodology involves perturbing model parameters of molecular dynamics potentials and fitting the simulation results to experimental data. The framework is general and can be applied to any material or simulation method. The fit is made using the Unified Monte Carlo method to experimentally measure double-differential scattering cross sections of light water at the Spallation Neutron Source at Oak Ridge National Laboratory. Mean values and covariance data were generated for model parameters, phonon density of states, double-differential cross sections, and total scattering cross sections. These posterior parameter values were very similar to their prior values with a maximum relative error of 0.54%. This falls within in the Unified Monte Carlo–calculated uncertainties on the order of 2.7%. Additionally, posterior double-differential cross sections agree favorably with ENDF/B-VIII.0 cross sections. The new thermal scattering law was tested by comparing it against benchmarks from the International Criticality Safety Benchmark Evaluation Project Handbook, which showed a slight improvement over the ENDF/B-VIII.0 library. Additionally, the covariance matrix of the phonon density of states was validated to confirm that the spread of keff from the density of states used to generate the covariance matrix was similar to the spread of keff from the density of states of the sampled covariance matrix.

The dataset contains results from a series of integral benchmark experiments performed at the China Institute of Atomic Energy (CIAE) to study neutron interactions with bismuth. These experiments were conducted using a 14 MeV deuterium-tritium (D-T) neutron source generated by a 400 kV nanosecond-pulse neutron generator. The experimental setup included a dual-axis electric sample stage for precise positioning and alignment of bismuth slabs, a multi-layer collimation system to reduce background noise, and a comprehensive detection system for accurate measurement of neutron leakage spectra.Measurements were carried out using bismuth slabs with effective thicknesses of 5 cm, 10 cm, and 15 cm, achieved by combining two large bismuth samples with dimensions of 30 × 30 × 5 cm and 30 × 30 × 10 cm. Experiments covered six angles: 47°, 58°, 73°, 107°, 122°, and 133°. The bismuth used in the study had a purity exceeding 99.997%, ensuring minimal impurity effects on the results. Time-of-flight (TOF) spectra were obtained for neutron energies ranging from 0.8 to 16 MeV, with pulse time distributions reconstructed using the MLEM algorithm.The detection system included a primary Φ2 × 2-inch EJ301 liquid scintillator for measuring neutron leakage, two Φ0.5 × 0.5-inch EJ301 neutron monitors for source neutron pulse recording, and a silicon carbide (SiC) detector to differentiate between alpha particles from the D-T reaction and protons from D-D reactions. The digital data acquisition system, based on the GDDAQ platform, utilized 16-channel Pixie-16 modules with a 500 MSPS sampling rate and 12-bit resolution, enabling precise and efficient data collection.Monte Carlo simulations using the MCNP-4C code were performed to compare experimental results with nuclear data libraries, including CENDL-3.2, ENDF/B-VIII.0, JENDL-5, and JEFF-3.3. These simulations and corresponding experimental measurements form the core of the dataset, providing a valuable resource for validating nuclear data and advancing the development of bismuth-based nuclear applications.

In order to evaluate the accuracies of the kinetics parameters predicted with the continuous-energy Monte Carlo calculation code, MVP3, core analysis was performed for light-water moderated UO2 and mixed oxide (MOX) fuel cores for which the ratios of effective delayed-neutron fraction (βeff) to prompt-neutron life time (ℓ) – represented hereafter by βeff/ℓ – and βeff values were measured. The results obtained with the JENDL-4.0-based neutron library showed that (the calculation values (C)/the measurement values (E)−1) ranged from −1.0% to 4.0% for the βeff/ℓ values with an average of 1.0% and a standard deviation of 1.1% for the 17 cores (the UO2, and UO2-MOX mixed cores) tested in the Tank-Type Critical Assembly (TCA). With respect to the βeff of one UO2 core in TCA, C/E − 1 was 1.2%. For the βeff values of the UO2 and MOX cores tested in the EOLE critical facility, C/E − 1 was −1.5% for the former and −4.6% for the latter. The calculated βeff value of the MOX core using the JEFF-3.2-based neutron library was larger by 4% than that calculated using the JENDL-4.0-based neutron library and showed better agreement to the measurement.

We describe how the prompt fission neutron spectrum (PFNS) was determined for the Manhattan Project at Los Alamos. Early work before World War II at American and British universities is described, together with theoretical work by Feather at Cambridge and Bethe at Los Alamos. As the Manhattan Project was being planned in 1942, two experiments on natural uranium were commissioned that proved to be influential: 1) An integral experiment at Chicago by Christy and Manley that accurately determined the average PFNS spectrum energy, 2.2 ± 0.2 MeV; 2) Bloch and Staub’s Stanford cyclotron measurement of the PFNS spectrum, which obtained an average energy of 1.70 ± 0.34 MeV. These two papers, previously unavailable outside of Los Alamos, are reproduced in the Supplementary Appendix. From these data, at the beginning of the project in 1943 Serber estimated an average 235U PFNS energy of 2 MeV, and indeed this agrees with today’s best estimate. The challenges facing the scientists involved both the availability of only very small samples of enriched uranium and plutonium targets, and fast neutron detection technologies. During the project, 235U and 239Pu PFNS were measured by Nicodemus and Staub. These also proved to be quite accurate and gave an average spectrum energy of 2 MeV for 235U. [This is not reproduced in the Appendix because it was published after the war in Physical Review 89, 1288 (1953)]. New methods were developed to enable more accurate measurements, and this paper describes how the PFNS was determined surprisingly well by 1945. We end by describing the post-war measurements in the 50s, including the PFNS data used by Ford and Wheeler in their simulations in 1951, the Bonner 1952 data, the seminal 1952 Watt paper with a new empirical parametrization of the PFNS, and the accurate PFNS measurement undertaken at Los Alamos by Cranberg et al. in 1956. We compare the measurements with our best understanding today as embodied in the Evaluated Nuclear Data File ENDF/B-VIII.0. Some images from historical documents in our Los Alamos National Security Research Center (NSRC) archives are shown.

The data related to the article "New experimental measurement of natSe(n, γ) cross sections between 1 eV to 1 keV at the CSNS Back-n facility" published in the journal Chinese Physics B. 74Se is one of 35 p-nuclei, and 82Se is a r-only nucleus, and their (n,γ) cross section are vital input parameters for nuclear astrophysics reaction network calculations. The neutron capture cross section in the resonance range of isotopes and even natural selenium samples has not been measured. Prompt γ-rays originating from neutron-induced capture events were detected by four C6D6 liquid scintillator detectors at the Back-n facility of China spallation neutron source (CSNS). The pulse height weighting technique (PHWT) was used to analyze the data in the 1 eV to 100 keV region. The deduced neutron capture cross section was compared with ENDF/B-VIII.0, JEFF-3.2, and JENDL-4.0, and some differences were found. Resonance parameters were extracted by the R-Matrix code SAMMY in the 1 eV - 1 keV region. It contains 2 text files, which are described below：1) natSe_CS_data.txt: The experimental data of natSe(n, γ) cross sections between 1eV and 100 keV.2) R_matrix_par.txt: resonance parameters of natSe extracted by R-matrix from experimental data.