Nuclear Data Evaluation Research Articles

A revision of recent taxonomic changes to the eyelash palm pitviper, Bothriechis schlegelii (Serpentes, Viperidae)

The taxonomy of the Bothriechis schlegelii species group has recently been expanded, nearly doubling the number of recognized species within this genus. However, the validity of these new species, defined primarily by mitochondrial DNA and without a robust evaluation of nuclear DNA and morphological data, warrants critical reassessment. This study re-examines the integrative data used in the initial analysis and employs additional visualizations of the genetic and morphological datasets. The results suggest that the genetic differences previously interpreted as species boundaries may instead reflect clinal variation, not independent lineages. Furthermore, DELINEATE species delimitation analysis and morphological assessments challenge the distinction of the proposed taxa. The findings indicate that only three species in the B. schlegelii group should be maintained as valid (B. nigroadspersus, B. schlegelii, and B. supraciliaris), while most taxa proposed by Arteaga et al. (2024) should be considered junior synonyms, or at most, as subspecies. This has significant implications for conservation priorities and resource allocation. In highlighting the need for a cautious and thorough approach to species delimitation, this study contributes to the broader discussion of taxonomic practices, the issue of taxonomic inflation, and their impact on biodiversity conservation.

Bayesian model averaging (BMA) for nuclear data evaluation

Bayesian model averaging (BMA) for nuclear data evaluation

Angular distribution measurements of neutron elastic scattering on natural carbon

The angular distributions of neutron elastic scattering on natural carbon were studied in the fast neutron energy region between 1 and 8 MeV. The experiments were carried out at the white neutron source of the Geel Electron Linear Accelerator (GELINA) facility by using thin and thick natural carbon samples. This work demonstrates the need for using thin samples to avoid strong multiple scattering effects. Neutrons and γ rays from scattering were detected using the ELastic and Inelastic Scattering Array (ELISA), a setup consisting of 32 liquid organic scintillators. The n−γ separation was achieved via pulse-shape analysis. For each sample a different approach in methodology is studied, one based on a global response function model and another one based on a per-detector model. The detectors are placed at eight different detection angles between 16∘ and 164∘ with respect to the neutron beam direction, allowing the simultaneous calculation of both the differential and the integral cross section by implementing the Gauss-Legendre quadrature rule. The neutron flux was measured with a U235 ionization chamber. The angular distributions were extracted relative to the U235(n,f) cross section. The results are compared with other experimental data available in the EXFOR library, along with the most recent nuclear data evaluations. The angle-integrated cross sections are in excellent agreement with the nuclear data evaluations and for the angular distributions, ENDF/B-VIII.0 is better reproducing the experimental data in all eight detection angles. Published by the American Physical Society 2024

Calculation method of particle angular distribution based on directional probability approach in the Monte Carlo simulation

Abstract To address the limitations of narrow applicability and low computational efficiency in the existing methods for calculating particle angular distribution using Monte Carlo simulations, this paper proposes a method based on the directional probability approach. In this method, when calculating the fluence at a recording point using the directional probability approach, virtual particles are binned based on the cosine of the angle between their direction and different reference directions, enabling statistical calculation of the particle angular distribution at the recording point. The simulated one-dimensional angular distribution of particles at the recording point agrees well with the angular distribution at the recording surface, with a total fluence deviation of less than 2% and an angular distribution deviation of less than 5%. Additionally, to facilitate the coupling of particle transport simulation with other physical fields, corresponding methods have been established to statistically calculate the fluence components in different directions of the xyz coordinate system. This method can calculate one-dimensional, two-dimensional angular distributions, and different directional components of particles at any location within the radiation field, providing technical support for research in ray detection, radiation protection, nuclear radiation field analysis, and nuclear data evaluation.

Fast neutron leakage spectra of the EUCLID experiment

Special nuclear material in sub-critical and critical configurations measured in integral experiments are important for validation and adjustment of nuclear data. Many different evaluations of nuclear data exist, and these different evaluations can provide different values for individual cross sections that vary due to the uncertainties in differential experiments or lack of such data. For integral experiments, differences in these individual cross sections can have compensating errors, which lead to the same answer. One example of this is the Jezebel critical assembly, where keff of the system is correctly computed by both ENDF/B-VIII.0 and JEFF-3.3, despite having substantially different underlying evaluated values for specific reactions (such as elastic and inelastic cross sections).To reduce compensating errors in nuclear data, the Experiments Underpinned by Computational Learning for Improvements in Nuclear Data (EUCLID) project has utilized machine learning to design a set of sub-critical and critical experiments. These experiments include slab- and cube-like configurations of 239Pu in the form of the ZPPR plates. Six different responses were measured on a total of thirteen different configurations. One of these responses, the neutron leakage spectrum, was measured using an EJ301D detector. The results of the neutron leakage spectra show good agreement (within 1–2 σ) with the expected spectrum from simulations and will be used in the subsequent nuclear data adjustment done by the EUCLID team.

Measurement of the neutron total cross section of 9Be at the Back-n white neutron source of CSNS* *Supported by the National Key Research and Development Plan (2016YFA0401603), the National Natural Science Foundation of China (11675155, 11790321) and Foundation of President of China Academy of Engineering Physics (YZJJLX2016003)

The neutron total cross section data of 9Be are essential in the nuclear structure model research of light nuclei and nuclear power installations. The neutron total cross section of 9Be in the 0.3 eV−120 MeV energy region has been measured using time-of-flight and transmission methods with the Neutron Total Cross Sectional Spectrometer (NTOX) based on the multi-cell fast fission chamber at the China Spallation Neutron Source (CSNS)-Back-n white neutron source (Back-n). The fission count-neutron energy distributions of 235U and 238U without samples and with Be samples with three thicknesses were measured in the double-bunch operation mode for a beam power of 100 kW. The Bayesian method was used to eliminate the influence of the double-bunch problem on neutron measurement in the energy region above 10 keV. The neutron total cross section of 9Be results was consistent with ENDF/B-VIII.0 evaluation library data in the 0.3 eV−20 MeV energy region. In the energy ranges of 0.3 eV to 10 keV and 0.01 to 20 MeV, the deviations between our results and the evaluation results of ENDF/B-VIII.0 were within 2.5% and 15%, respectively. In the resonance energy region, the measured resonance energies in our experiment were 0.63, 0.82, and 2.8 MeV, respectively. The results showed that the total cross section uncertainties of three Be samples were within 2.2% in the energy region below 1 MeV. The total cross section uncertainty of 30 mm Be from 235U was the smallest and less than 5% in the energy region of 0.3 eV−120 MeV. The results of this experiment can provide technical support for further data analysis and related nuclear data evaluation.

A Nuclear Data Evaluation Pipeline for the Fast Neutron Energy Range – using heteroscedastic Gaussian processes to treat model defects

A Nuclear Data Evaluation Pipeline for the Fast Neutron Energy Range – using heteroscedastic Gaussian processes to treat model defects

Dimensionality reduction through tensor factorization: Application to ab initio nuclear physics calculations

The construction of predictive models of atomic nuclei from first principles is a challenging (yet necessary) task towards the systematic generation of theoretical predictions (and associated uncertainties) to support nuclear data evaluation. The consistent description of the rich phenomenology of nuclear systems indeed requires the introduction of reductionist approaches that construct nuclei directly from interacting nucleons by solving the associated quantum many-body problem. In this context, so-called ab initio methods offer a promising route by deriving controlled (and systematically improvable) approximations both to the inter-nucleon interaction and to the solutions of the many-body problem. From a technical point of view, approximately solving the many-body Schrödinger equation in heavy open-shell systems typically requires the construction and contraction of large mode-4 (mode-6) tensors that need to be stored repeatedly. Recently, a new dimensionality reduction method based on randomized singular value decomposition has been introduced to reduce the numerical cost of many-body perturbation theory. This work applies this lightweight formalism to the study of the Germanium isotopic chain, where standard approaches would be too expansive to run. Inclusion of triaxiality is found to improve the overall agreement with experimental data on differential quantities.

Data assimilation of decay heat experiments for feedback on nuclear data

Integral decay heat experiments can provide interesting feedback on particular nuclear data (decay data and fission yields mainly). After ensuring that the C/E discrepancies were mostly due to nuclear data discrepancies, a Bayesian inference approach can be applied. Nevertheless, the results strongly depend on the quality of the experiment and on our capability to estimate realistic experimental correlation matrices when considering several integral experiments in the assimilation process. A former study performed in 2019 was dedicated to the data assimilation of a large C/E dataset from the experimental validation database of fuel inventory calculations with the DARWIN2.3 package in order to provide feedback to the nuclear data evaluators. This paper is an attempt to exploit the General Electric decay heat experiments performed in the USA in the 1980s in order to confirm or not the trends on four particular cumulated fission yields: 235U(nth,f)133Cs, 235U(nth,f)137Cs, 239Pu(nth,f)106Ru and 239Pu(nth,f)144Ce.

Need for precise nuclear structure data for reactor studies

This paper highlights the strong need for precise nuclear structure and decay data measurements in order to perform high-quality modelling on nuclear reactors and other applications. The context of nuclear data evaluation, as well as the importance of low uncertainty evaluations, will be first presented. The importance of such data for interpreting nuclear data experimental measurements is stressed throughout. To demonstrate this, we will explain how mass and charge-dependent fission yields, decay data (in particular for the purpose of residual heat calculations), and inelastic neutrons scattering cross section rely on nuclear structure and decay information and how new and higher quality in such data can lead to improved accuracy in the precision of evaluated nuclear data.

Development of HELIOS/BIPR/PARCS/MCNP6 Computation Route for WWER RPV Neutron Fluence Analysis and Validation Against Ex-Vessel Detector Measurement Data

A computational route was developed for precise calculation of fast neutron fluence on a WWER-type reactor pressure vessel (RPV). The method is based on the transfer of neutronics data from HELIOS-2 lattice calculations and nodal diffusion neutronics data (power, density, and temperature) from BIPR7.1 and PARCS 3.36/PATHS core calculations into a three-dimensional (pinwise axially distributed) fixed neutron source for modeling of transport of fast neutrons from the reactor core to the outer surface of the RPV using MCNP6.2. Validation of the proposed computational method was carried out based on comparative analysis of MCNP6.2-predicted and neutron dosimetry–measured reaction rates [54Fe(n,p)54Мn, 93Nb(n,nʹ)93mNb, and 58Ni(n,р)58Со] on the outer surface of the Armenian Nuclear Power Plant (ANPP) Unit 2 RPV. Validation revealed that the MCNP6.2-predicted fast neutron fluence results are very sensitive to the ENDF-B neutron data. Particularly, MCNP6.2 with ENDF/B-VIII.0 significantly underpredicts (20% to 30%) fast neutron fluence while using ENDF/B-VII.1 data overpredicts it. Adding revised beta-released evaluations of 54Fe, 56Fe, 57Fe, and 16O from the International Nuclear Data Evaluation Network (INDEN) to ENDF/B-VIII.0 allows one to obtain reasonable agreement with measurement results for all types of measured reaction rates.

A new determination of the 68Ga half-life and evaluation of literature data

Inconsistencies within the 68Ga half-life dataset have been repeatedly identified in nuclear decay data evaluations performed over the past decade. In this paper, a half-life has been determined using ionisation chambers measurements and HPGe gamma-ray spectrometry (T1/2 = 67.783(27) min). The current literature has been reviewed and possible causes of discrepancy have been identified, such as neglected contaminant effects, robustness of analysis, and underestimation of systematic effects. This could lead to a significant improvement of the half-life in future evaluations.

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.

Studying differential cross section for elastic proton scattering by a tensor model

Previous studies have shown the reliability of the Bayesian Gaussian CANDECOMP/PARAFAC tensor decomposition (BGCP) algorithm in predicting the independent fission yields and evaluating global uncertainty of the data. This work proposes to apply the BGCP algorithm for the purpose of predicting and evaluating the differential cross section for elastic proton scattering. The tensor model is developed by training with the experimental data of elastic proton scattering from the EXFOR database. The Root Mean Square Error is used to evaluate the experimental data and search for potential errors and omissions in EXFOR, in order to identify outlying data of 98Mo, 100Mo, 144Sm and 209Bi. After the revision for the outlying data in EXFOR, the reproduced values for 144Sm are compared with the remaining controversial entries. Besides, the predicted data for 9C are compared with the published data not available in EXFOR. The results demonstrate that the tensor model can make reasonable predictions and has a role to play in evaluating the quality of elastic proton scattering data. The tensor model may become a powerful tool for nuclear data evaluation in EXFOR.

Prompt fission neutron multiplicity in the 239Pu(n, f) reaction and its energy dependence

Measuring prompt fission neutrons to high precision is an experimental challenge, especially for radioactive fissioning nuclides. However, accurate average multiplicities, , and kinetic energy distributions of prompt fission neutrons are essential for fundamental and applied nuclear physics. We present here a recent measurement of the 239Pu (n,f) as a function of the incident-neutron energy, over the range 1-700 MeV. The measurement was performed with a cutting-edge setup and an innovative technique, which allowed to minimize and account for the main sources of bias. An unprecedented precision was therefore achieved. Our data are compared to GEF predictions as well as to evaluated libraries. For the first time, at low energies, the ENDF/B-VIII.0 nuclear data evaluation is validated with an independent measurement and the evaluated uncertainty reduced by up to 60%. This work paves the way to precisely measure prompt fission neutron multiplicities on highly radioactive nuclei.

Consistency Among the Results of the ASPIS Iron88, PCA Replica, and PCA ORNL Benchmark Experiments

Iron is an essential element in the construction materials for fission and fusion reactors. Due to its complexity, the evaluation of iron cross sections continues to represent a challenge for the international nuclear data community. A comprehensive validation of any new nuclear data evaluation (and the computational procedure) against experimental benchmarks is therefore needed. The shielding benchmark database SINBAD includes relatively numerous experiments with iron as a shielding material; altogether, 27 benchmarks and several more are known but have not yet been evaluated in the database. However, in order to use the benchmark information with confidence and to rely on the predictions based on integral benchmark calculations, it is crucial to verify the quality and accuracy of the measurements themselves, as well as the (completeness of) available experimental information. This is done in the scope of the benchmark evaluation process. A further check of the reliability of the experimental information can be achieved by intercomparing the results of similar types of benchmark experiments and checking the consistency among them.

Structural material nuclear data basic research

The nuclear data evaluation for deuteron-induced reactions andα-particle emission by neutron interactions is addressed within “Nuclear data for fusion technology, from basic research to full-scale applications.” The status and open questions related to these subjects in the area of nuclear data for fusion technology, specifically for the nuclear design of the ITER fusion device, the European DEMO fusion reactor, and the IFMIF-DONES Irradiation Facility, are briefly reviewed. A firm demand for accurate cross-sections of reactions induced by neutrons and deuterons exists, in this respect, within a more enlarged energy range up to 50 MeV than for fission applications. The current requirements are closely met by the TENDL Evaluated Nuclear Data Library, settled using the TALYS nuclear model code, which is one of the most widely used codes in basic research and applications including nuclear fusion technology. However, further improvement of this data library has recently been suggested, while, with respect to fission applications, not only the aforementioned energy range but also the diversity of nuclear data for fusion technologies is plainly stretched. Consequently, the progress of nuclear data activities conducted more recently on deuteron-induced reactions andα-emission by neutron interactions, throughout the European Fusion Program and subsequent to previous achievements within F4E and EUROfusion programs, is wholly summarized.

Experimental and theoretical study of the Cu65(n,p)Ni65 reaction cross section from reaction threshold up to 25 MeV

The cross section of the $^{65}\mathrm{Cu}(n,p)^{65}\mathrm{Ni}$ reaction was studied experimentally at three different neutron energies using an activation technique. The quasimonoenergetic neutrons were produced via the $^{7}\mathrm{Li}(p,n)$ reaction at the 14UD BARC-TIFR Pelletron facility in Mumbai, India. Al monitor foils along with Cu samples were activated to determine the incident neutron flux. The activities of the reaction products were measured using a high resolution high purity germanium spectrometry system. Statistical model calculations were performed using the reaction codes talys (ver. 1.9) and empire (ver. 3.2.3) from the reaction threshold to the neutron energy of 25 MeV. Additionally, the effects of various combinations of the theoretical nuclear level densities (NLDs), optical model potentials (OMPs), preequilibrium models (PEs), and \ensuremath{\gamma}-ray strength functions (\ensuremath{\gamma} SFs) were considered for the reproduction of experimental data. The input parameters needed in theoretical calculations to reproduce the present and previous measurements were taken from the RIPL-3 database. The present results are compared with the previous measurements, with the latest evaluations of the ENDF/B-VIII.0, JEFF-3.3, JENDL-4.0/HE, CENDL-3.2, TENDL-2019, and FENDL-3.2 libraries, and with the theoretically calculated values based on talys and empire codes. Furthermore, the cross section of the $^{65}\mathrm{Cu}(n,p)^{65}\mathrm{Ni}$ reaction was estimated within the neutron energies of 14--15 MeV using different systematic formulas. These estimated cross sections by various systematic formulas were compared with the available experimental data. The present data will help to understand the nuclear reaction theory (models) in higher energy regions and improve the evaluated nuclear data evaluation that is needed for fundamental nuclear applications.

Nuclear data for fusion: inventory validation successes and future needs

Nuclear data, describing neutron reaction probabilities (cross sections) and decay behaviour, are critical to the design and operation of fusion experiments and future fusion power plants. Equally vital, are the inventory codes that use the data to predict neutron-induced activation and transmutation of materials, which will define the radiological hazards that must be managed during reactor operation and decommissioning. Transmutation, including gas production, combined with the neutron-induced displacement damage, will also cause the properties of materials to degrade, for example through swelling and embrittlement, eventually limiting the lifetime of components. Thus validated and accurate nuclear data and inventory codes are essential. For data validation there are decay heat measurements performed at FNS in Japan more than 20 years ago. The experiments produced an invaluable database for benchmarking of nuclear data libraries; the latest versions of several international libraries perform well against this data during tests with the FISPACT-II inventory code, although there is still scope for improvement. A recent attempt to provide fusion-relevant validation based on γ-spectroscopy data from neutron-irradiated material samples tests produced predictions for short-lived (several hours or less) radionuclides. The detailed analysis performed for molybdenum demonstrates how these data could eventually provide a new benchmark, and also illustrates the potential benefits of further experiments targeting the longer-lived radionuclides relevant to maintenance and decommissioning timescales. There are also some successful tests of transmutation predictions with FISPACT-II. These direct validations of inventory simulations are critical for lifetime predictions and future experiments should learn lessons from the examples described for tungsten, which demonstrate the importance of an accurate description of the neutron spectrum in experiments. More novel experimental techniques are needed to measure helium production in materials such as Fe and C, but the need to validate the nuclear data evaluations used by simulations should motivate future experimental efforts.

STEK: A potential fast spectrum benchmark for fission product cross sections

The reactivity worth of many fission product samples was measured in fast spectrum conditions in the STEK facility during the early 1970s. These results were then used to improve and validate nuclear data evaluations for fission products, but in the last 2 decades STEK has not been used: the nuclear data evaluations for fission products in the current versions of libraries like ENDF/B or JEFF have not been tested against STEK results. Here we argue that the STEK data are still valuable, because there is no other data set that can replace them, and the interest in fast spectrum conditions is picking up strongly of late. It should be considered to evaluate whether a benchmark can be defined for the International Reactor Physics Experiments Handbook.