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DOE OSTI2022

Easy_PERT: a Python tool for writing PERT cards and parsing PERT card results [Slides]

This presentation begins by providing an overview of the PERT card. The PERT card uses differential operator method to compute first- and second-order tally variations due to density, composition, and reaction cross-sections. It is possible to have multiple PERT cards in one MCNP input deck to study tally variations for several sets of nuclides, reactions, and energy ranges. Furthermore, the METHOD option tells MCNP to calculate either the perturbed tally (METHOD=-1, -2, -3) or the change in the unperturbed tally (METHOD=1, 2, 3). In summation, a powerful use-case for the MCNP code PERT card is that it facilitates calculating tally sensitivities to nuclear data. Writing PERT card entries and parsing output MCTAL files is tedious and error prone. however, Easy_PERT makes use of existing tools (Faust and MCNPTools) to handle writing PERT card entries and parsing the output MCTAL files. The PERT card is early in the development process and planned upcoming capabilities include calculating sensitivities and combining MCTAL files from separate runs into one JSON file.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

IER 516 - Zirconium Test Assembly (ZTA) Preliminary Experimental Design [Slides]

Zirconium present in large quantities in criticality safety applications such as fuel manufacturing and processing facilities, and in commercial thermal nuclear reactors. ICSBEP lacks benchmarks optimized for zirconium sensitivity to validate zirconium evaluations. Evaluators lack isotopic differential nuclear data in regions of high importance to fission systems.

K-PSO 2.0↗
DOE OSTI2026

The ENDF/B Nuclear Data Library and Its Impact on Reactor Simulations

The ENDF/B library, which is developed, maintained, and distributed by the Cross Section Evaluation Working Group, is the main source of nuclear data for analyses and computational simulations in nuclear applications, like different nuclear reactors concepts, radiation shielding, medical applications, astrophysics, etc. The library is constantly being improved and updated, with each release bringing an optimal representation of nuclear interactions as they are understood in their time. The most recent release, ENDF/B-VIII.1, represents a significant improvement in terms of the performance and consistency of the measured differential data relative to previous versions, as it combines the most recent experimental differential data and advanced theoretical nuclear models. As one of the many highlights, ENDF/B-VIII.1 restores a high-burnup depletion performance, comparable to ENDF/B-VII.1, that had been degraded in ENDF/B-VIII.0, while further improving the performance in criticality benchmarks, such as those in the Mosteller’s suite. Additionally noteworthy is the improved performance of ENDF/B-VIII.1 in radiation shielding and thick-target leakage spectrum integral experiments, which are also important for fusion and reactor applications. In this work we present in a very summarized way the main updates implemented in the ENDF/B-VIII.1 release and its main impacts, and also begin to delineate the path forward as to what to expect in the future for the next ENDF/B release, which, based on the timeline of the past few releases, is estimated to happen around 5 years from now. We emphasize that the ENDF/B-VIII.1 release was the product of an enormous collaborative effort among many authors and that for a complete detailed picture, the reader is strongly encouraged to refer to the article accompanying the release, which is currently in the publication process, but available as preprint [G. P. A. Nobre et al. arXiv Preprint arXiv: 2511.03564 (2025)].

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Enhancing nuclear data validation analysis by using machine learning

In this work, we demonstrate how machine learning (ML) techniques can be used as an auxiliary tool for nuclear data validation analysis. The output of the ML analysis can inform evaluators and validators on the quality, or lack thereof, of specific nuclear data and benchmark experiments with respect to simulating these experimental benchmark values. To this end, measured and simulated effective neutron multiplication factors, k eff , of 875 selected ICSBEP critical assemblies and the latter's sensitivities with respect to nuclear data as well as benchmarks' features (e.g., material nuclide, core geometry) are used as input for a random forest (RF) regression model. The RF is used to encode the complex inter-dependencies between thousands of nuclear data features (i.e., sensitivity profiles and aspects describing the measurements) and 875 simulated and experimental k eff values in order to understand what nuclear data features are most informative for predicting bias. The complexity of relationships and high-dimensional space is difficult-to-impossible to search fully using simply expert judgment. As a first proof-of-concept—step, it is demonstrated that this technique is able to correctly trace large discrepancies between simulated and experimental k eff back to fabricated shortcomings in nuclear data that were perturbed to simulated k eff values. In a second, real-case scenario, step, the RF algorithm is used to validate the ENDF/B-VIII.0 library in comparison to ENDF/B-VII.1 nuclear data. One case is showcased where the chosen ML algorithms highlighted nuclear data (the 19 F(n,inl) cross section from 0.4–0.9 MeV) that are shown to be problematic by comparing them to associated differential experimental data and nuclear data from other libraries. In addition to that, it is shown that the RF results point towards poor benchmark experiments and associated underestimated uncertainties (e.g., the PU-SOL-THERM-028 series). However, using the RF algorithm for validating nuclear data with respect to k eff is currently limited to pinpointing groups of questionable nuclear data due to the inherent correlations between features introduced by the nuclear data themselves and how k eff is simulated. Due to this, we recommend that the ML methods presented be used to augment—rather than replace—the expert knowledge of evaluators and validators.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2023

New Measurements to Resolve Discrepancies in Evaluated Model Parameters of 181 Ta

To resolve discrepancies in evaluated cross sections among major nuclear data libraries, energy-differential neutron transmission and radiative capture yield of 181 Ta were measured from 0.15 to 100 keV using multiple sample thicknesses. The new measurements provide resolution such that the resolved resonance region (RRR) can be evaluated up to at least 2.5 keV and the unresolved resonance region can be evaluated up to at least 100 keV. The transmission and capture yield measurements were modeled using resonance parameters from three major libraries to assess the predictive capability of each. It was found that JENDL-5.0 performed best in the RRR. Because of the poor performance of the U.S. ENDF/B evaluation, it is recommended that ENDF/B be reevaluated for 181 Ta.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Continuing Development of the Nuclear Data Processing Code AMPX [Poster]

The ENDF/B-VIII.1 evaluation library has seen a great growth in the thermal neutron scattering sub-library. The SCALE code system has traditionally approached CE transport by assuming that the CE library on disk represented the fully expanded cumulative probability distributions, conditional on exiting angle and marginal on exiting energy. While this is a complete description of the data, it comes at the potential cost of large amounts of on-disk storage. This approach was strained by several TSL files in ENDF/B-VIII.1, such as graphite, which contained data for a large number of Bragg edges. In the fully expanded probability distributions, this was found to be a disproportionately large fraction of the SCALE CE library.

GNDS↗
DOE OSTI2020

Nuclear Data Sheets for A=233

The present evaluation of the known A=233 nuclides ( 233 Fr, 233 Ra, 233 Ac, 233 Th, 233 Pa, 233 U, 233 Np, 233 Pu, 233 Am, 233 Cm) and 233 Bk represents an update and revision, after 15 years of the previous evaluation of A=233 published by B. Singh and J.K. Tuli (2005Si15). Laboratory identification of 233 Bk nuclide is still considered as tentative. Various decay and reaction data are evaluated and compared. Adopted data, levels, gammas, level half-lives, spins, parities, and configuration assignments are provided. For 233 Fr, 233 Ra, 233 Ac, 233 Pu, 233 Am and 233 Cm, information is available for the respective ground states only. The decay schemes from β - decays of 233 Fr and 233 Ra, and from ε decay of 233 Am and 233 Cm are unknown, while those from β - decay of 233 Ac, and from ε decay of 233 Np are considered incomplete. The α decay of 237 Np to 233 Pa has been studied in detail by many investigators, yet the decay scheme is considered incomplete as several very low-energy transitions are involved, without the knowledge of their intensities, resulting in mismatch of directly measured α feedings and those deduced from γ-transition intensity balances. The decay schemes from α decays of 237 Am, 237 Cm and 237 Cf to respective A=233 daughters are considered incomplete. Particle-transfer and high-spin data are available for 233 Th, 233 Pa and 233 U nuclei, which remain the most extensively studied in this mass chain. However, data for half-lives for the excited states are available only for one level in 233 Pa, and for five levels in 233 U, thus limiting the knowledge of reduced transition probabilities in this mass chain.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

B(E2) Evaluation for 0{sub 1}{sup +}→2{sub 1}{sup +} Transitions in Even-Even Nuclei

A collaborative study by Brookhaven-McMaster-Central Michigan is underway to evaluate B(E2)↑ for 0{sub 1}{sup +}→2{sub 1}{sup +} transitions. This work is a continuation of a previous USNDP evaluation and has been motivated by a large number of recent measurements and nuclear theory developments. It includes an extended compilation, data evaluation procedures and shell model calculations. The subset of B(E2)↑ recommended values for nuclei of relevance to the double-beta decay problem is presented, and evaluation policies of experimental data and systematics are discussed. Future plans for completion of the B(E2;0{sub 1}{sup +}→2{sub 1}{sup +}) evaluation project are also described.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Initial Delayed Critical Configuration of SHEBA I and Subcritical Measurements by Californium Source-Driven Noise Analysis

In September 1980, the initial delayed critical configuration of the Solution High-Energy Burst Assembly (SHEBA I) was assembled, and the US Department of Energy’s Oak Ridge National Laboratory (ORNL) staff performed near critical and subcritical measurements using the californium source-driven noise analysis (CSDNA) method at the Los Alamos National Laboratory (LANL) Critical Experiments Facility. An unreflected 56 cm outside diameter stainless-steel cylindrical tank was partially filled with uranyl fluoride solution (235U enrichment was 4.95 wt %) until delayed criticality was achieved. Then, measurements were performed for various fuel solution heights from delayed critical to 60% of the height required for delayed criticality. The stainless-steel tank had an inside diameter of 54.6 cm and a height of 105 cm and was partially (20–36.5 cm) filled with an aqueous solution of uranyl fluoride (with a density of 2.162 g/cm 3 ). The density of the uranium in solution was 1.042 g/cm 3 , and the solution had a H/U atomic ratio of 20.43. The tank had a 6.35 mm wall thickness and an axial reentrant tube with an inside diameter of 6.02 cm and wall thickness of 0.165 cm. The tank was essentially unreflected on the top and sides because it was in a thin metal shed. The reactivity of the near–delayed critical configuration was −10.4 cents, which corresponds to a k eff value of 0.99922. In addition to the CSDNA measurements, the prompt neutron decay constants were determined from break frequency noise analysis (BNFA) measurements. The subcritical neutron multiplication factors from CSDNA and BFNA compared extremely well. These data can be used as the basis of International Nuclear Criticality Safety Evaluation benchmark for the near-critical configuration, and the k eff values at various subcritical configurations can be used as subcritical benchmarks. Furthermore, the measured prompt neutron decay constants can be used as reactor physics benchmarks. This report documents the experimental information for the measurements performed so that, at a later date, researchers could perform the required uncertainty and calculational analyses and documentation to use these data for an International Criticality Safety Benchmark Evaluation Program (ICSBEP) or Nuclear Energy Agency benchmarks. The measured prompt neutron decay constants can be used as the basis of a benchmark for the International Reactor Physics Evaluation Program (IRPhEP). The data from these measurements are available from the ORNL Records Management Services Department, and the logbook is available from ICSBEP at Idaho National Laboratory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Nuclear Data Sheets for A=249

Here, the evaluated spectroscopic data are presented for these nuclides: 249 U, 249 Np, 249 Pu, 249 Am, 249 Cm, 249 Bk, 249 Cf, 249 Es 249 Fm, 249 Md and 249 No. This work supersedes the earlier full evaluation of A=249 published by 2011Ab07.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2014

Format requirements of thermal neutron scattering data in a nuclear data format to succeed the ENDF format

In November 2012, the Working Party on Evaluation Cooperation Subgroup 38 (WPEC-SG38) began with the task of developing a nuclear data format and supporting infrastructure to replace the now nearly 50 year old ENDF format. The first step in this process is to develop requirements for the new format and infrastructure. In this talk, I will review the status of ENDF's Thermal Scattering Law (TSL) formats as well as support for this data in the GND format (from which the new format is expected to evolve). Finally, I hope to begin a dialog with members of the thermal neutron scattering community so that their data needs can be accurately and easily accommodated by the new format and tools, as captured by the requirements document. During this discussion, we must keep in mind that the new tools and format must; Support what is in existing data files; Support new things we want to put in data files; and Be flexible enough for us to adapt it to future unanticipated challenges.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Quality Quantification of Evaluated Cross Section Covariances

Presently, several methods are used to estimate the covariance matrix of evaluated nuclear cross sections. Because the resulting covariance matrices can be different according to the method used and according to the assumptions of the method, we propose a general and objective approach to quantify the quality of the covariance estimation for evaluated cross sections. The first step consists in defining an objective criterion. The second step is computation of the criterion. In this paper the Kullback-Leibler distance is proposed for the quality quantification of a covariance matrix estimation and its inverse. It is based on the distance to the true covariance matrix. A method based on the bootstrap is presented for the estimation of this criterion, which can be applied with most methods for covariance matrix estimation and without the knowledge of the true covariance matrix. The full approach is illustrated on the {sup 85}Rb nucleus evaluations and the results are then used for a discussion on scoring and Monte Carlo approaches for covariance matrix estimation of the cross section evaluations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Producing ENDF/B-quality Evaluations of 239 Pu(n,f) and 235 U(n,f) Average Prompt Neutron Multiplicities using the CGMF Model

This report documents evaluations of the neutron-induced 239 Pu and 235 U average prompt neutron multiplicities, $\overline{v}_p$, using the CGMF model developed at LANL. These evaluations are not updates of existing ENDF/B-VIII.0 nuclear data, but were both re-done from "scratch". That is, all experimental data were extracted from EXFOR, re-normalized to the newest nuclear data representing monitor observables, and detailed uncertainties were estimated from information in EXFOR as well as from templates of expected measurement uncertainties. We also included experimental data that were not yet available for ENDF/B-VIII.0, for instance, the data of Marini et al. for 239 Pu(n,f) $\overline{v}_p$ or the data of Khoklov et al. for 235 U(n,f) $\overline{v}_p$. Finally, we include the CGMF model; it computes $\overline{v}_p$ based on model parameters that at the same time calculate fission fragments as a function of mass, Y (A), the average total kinetic energy as a function of incident-neutron energy, $\langle TKE \rangle$(E inc ), etc. Such a detailed fission model that ties together many fission quantities has not been used to date for any $\overline{v}_p$ evaluation in ENDF/B. Here, we show that we can get evaluated 235 U(n,f) and 239 Pu $\overline{v}_p$ that not only correspond well to experimental $\overline{v}_p$, but that the associated evaluated model parameters also yield parameterizations of Y (A), $\langle TKE \rangle$(E inc ), etc., that correspond well to their respective experimental data. The new evaluated 239 Pu(n,f) $\overline{v}_p$ was also combined into a 239 Pu test file with the newest nuclear data for the 239 Pu prompt-fission neutron spectrum and fission cross sections. This new 239 Pu file performed reasonably well in predicting PU-MET-FAST assemblies, reaction rates in Jezebel and Flattop and three LLNL pulsed spheres. Due to that, the new 239 Pu(n,f) $\overline{v}_p$ evaluation presented here is being considered for ENDF/B-VIII.1. Hence, we can show here that CGMF is able to produce ENDF/B quality $\overline{v}_p$ nuclear data.

239Pu, 235U, Average Prompt-fission Neutron Multip↗
DOE OSTI2018

Nuclear Data Sheets for A=39☆

Experimental nuclear structure data from various reactions and decays have been evaluated for all known nuclides of mass 39 (Mg, Al, Si, P, S, Cl, Ar, K, Ca, Sc, Ti). Detailed evaluated nuclear structure information is presented for each reaction and decay and the best values combining all available data are recommended for level energies, half-lives, γ-ray energies and intensities, decay properties (energies, intensities and placement of radiations), and other spectroscopic data. The {sup 39}K, {sup 39}Ca and {sup 39}Ar are the most extensively studied nuclides; no excited states are known in {sup 39}Mg and {sup 39}Ti. This work supersedes the earlier full evaluation of A=39 by B. Singh and J. Cameron (2006Si02).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Subcritical Copper-Reflected-α-phase Plutonium (SCRαP) Integral Experiment Design

A Subcritical Copper-Reflected α-phase Plutonium (SCRαP) integral benchmark experiment is proposed where multiplication will be approximated using correlated neutron data from a detector system consisting of {sup 3}He tubes inside polyethylene. Measurements will be performed on various subcritical experimental configurations consisting of a weapons-grade plutonium sphere surrounded by different Cu thicknesses. In addition to the proposed base experimental configurations with Cu, additional configurations are planned with the plutonium ball nested in various thicknesses of interleaved polyethylene spherical shells mixed in with the Cu shells. The polyethylene is intended to provide fast neutron moderation and reflection, resulting in additional measurements with differing multiplication, spectra, and nuclear data sensitivity. The proposed experiments will performed at the National Criticality Experiments Research Center (NCERC). NCERC is home to a 4.5-kg α-phase stainless steel clad plutonium sphere that is referred to as the BeRP (Beryllium-Reflected Plutonium) ball due to its historical use in a beryllium-reflected critical experiment. In 2014, similar subcritical measurements were performed with the BeRP ball surrounded by nickel and tungsten. The nickel-reflected measurements were documented as a benchmark evaluation which was accepted by the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Similar to past measurements, the proposed work will help both identify deficiencies and quantify uncertainties in nuclear data and validate computational methods related to neutron multiplication inference for subcritical benchmark evaluations. Monte Carlo simulations were performed for the preliminary design of a subcritical Copper-Reflected - phase Plutonium (SCRP) integral benchmark experiment. Specifically, MCNP{sup R} version 6 was used to determine the k{sub eff} and associated nuclear data sensitivities for different preliminary configurations (reflector thickness, material type). The final experimental design process will include continuous-energy nuclear data sensitivity analysis. This will help assess the potential impact of this integral measurement as it relates to international efforts to continuously improve current libraries (ENDF, JEFF, JENDL etc) that still tend to over/under-estimate the measured results of integral experiments, sometimes significantly. Deficiencies in underlying nuclear data quantities such as Nubar have been shown to have an effect on inferred values from subcritical measurements as well. The proposed experiments and subsequent computational validation will help identify such deficiencies. It will also help validate new nuclear data evaluations, including one for Cu that was recently performed in the resolved region up to 100 MeV for {sup 65}Cu and {sup 63}Cu which is expected to improve Cu-related benchmark simulation performance. The final experimental design phase will also include simulation results that will incorporate the detectors that will be used for the subcritical measurements. For the benchmark portion of the measurements, the main detectors that are being considered are portable neutron multiplicity counters made up of {sup 3}He tubes embedded in polyethylene and lined with cadmium. These detectors employ a data acquisition system that produces list-mode data (correlated information where each event is listed sequentially) that can be analyzed using the Feynman variance-to-mean method. Other detector systems including SNAP (single {sup 3}He tube embedded in polyethylene), and a high-purity germanium (HPGe) detector are being considered for supplementary measurements only and are not intended to be used for the final benchmark measurements. Decoupling the supplementary detectors from the system eliminates any contribution to the total system multiplication, and greatly simplifies the final computational benchmark model. Therefore, an optimization of the portable neutron multiplicity counter placement in the benchmark measurements will be performed to ensure that there is minimal contribution to the total system multiplication. In addition, new temperature measurement techniques and hardware will be explored. (authors)

42 ENGINEERING↗
DOE OSTI2018

Correlated Production and Analog Transport of Fission Neutrons and Photons Using Fission Models FREYA, FIFRELIN, and the Monte Carlo Code TRIPOLI-4

Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by the international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, the precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: Fission Reaction Event Yield Algorithm (FREYA) and Fission Fragment Evaporation Leading to an Investigation of Nuclear Data (FIFRELIN). TRIPOLI-4 simulations can now be performed, either by connecting via an application programming interface to the Lawrence Livermore National Laboratory fission library including FREYA, or by reading the external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from the Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, the broad underlying principles of the two fission models are recalled. We then present the experimental measurements of neutron angular correlations for 252 Cf(sf) and 240 Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of this paper, simulation results are compared with the experimental data. Spontaneous fissions in 252 Cf and 240 Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in the analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using either FREYA or FIFRELIN, are compared with the experimental results. In conclusion, for 240 Pu(sf), the measured correlations were used to tune one of the FREYA model parameters. Similarly, different sets of parameters have been tested in FIFRELIN to try to improve the agreement with the experimental data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Impacts of processing decisions on TNSL cross sections and their applications

Thermal neutron scattering law (TNSL) data describe low-energy neutrons scattering off of bound materials, and can have a significant impact on modeling any system with slow neutrons, including nuclear reactors. Previous work to introduce TNSL data to neutron transport codes at LLNL focused on COG and TART, with the limitation that these codes require highly specialized data processing and formatting. We have recently increased efforts to process TNSL data with the LLNL nuclear data processing code FUDGE. FUDGE reads and writes the evaluated and processed files using the generalized nuclear database structure (GNDS). This process uncovered some significant difficulties in processing TNSL data, and unearthed assumptions made in current TNSL data processing that we have found inadequate. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Nuclear Data Sheets for A = 87

The evaluated experimental data are presented for 14 known nuclides of mass 87 (Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru). New datasets have been added for {sup 87}Ga, {sup 87}Se, {sup 87}Kr, {sup 87}Rb, {sup 87}Sr, {sup 87}Nb, and {sup 87}Tc. The observation of and new lifetime measurements for {sup 87}As at the NSCL Coupled Cyclotron Facility and the Oak Ridge Isochronous Cyclotron at the HRIBF–ORNL Facility are included in the adopted levels for this isotope. The new evaluation also includes an isomeric level at unknown energy for {sup 87}Tc, changes to the adopted levels for {sup 87}Nb based on newly published data including an extension of the high spin levels and new lifetime measurements, and new heavy ion data for {sup 87}Kr. New high spin levels for {sup 87}Rb were observed, indicating possible neutron core excitations. In addition, the ground state lifetime of {sup 87}Rb has been revised and a new lifetime measurement for an excited level in {sup 87}Rb indicates a possible proton spin flip. New excitation levels have been observed for the first time in {sup 87}Se. New direct and precise measurement of atomic masses of {sup 87}Tc, {sup 87}Mo, {sup 87}Nb, {sup 87}Zr, {sup 87}Mo and {sup 87}Tc have greatly improved the Q value landscape in this mass region. This work supersedes the previous A=87 evaluation published in 2002He09, and Q values for all isotopes were updated to incorporate the mass evaluation in 2012Wa38.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗