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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Consistency of $^{16}O(n,α)$ cross sections

The evaluated $^{16}O(n,α)$ cross sections in the ENDF/B-VIII.0 nuclear data library remain uncertain because of systematic discrepancies in the measured data. In the energy region below the first excited state, R-matrix analyses rely heavily on the inverse reaction, and in particular, they rely on the measured $^{13}C(α,n)^{16}O$ cross sections reported by Bair et al. in 1973 and Harissopulos et al. in 2005. The Harissopulos cross section values are systematically lower than those previously reported by Bair. Here, drawing on the available experimental information, this paper briefly describes and demonstrates that the two sets of measured cross sections are consistent.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Estimation of the NiCu Cycle Strength and Its Impact on Type I X-Ray Bursts

Abstract Type I X-ray bursts (XRBs) are powered by thermonuclear burning on proton-rich unstable nuclides. The construction of burst models with accurate knowledge of nuclear physics is required to properly interpret burst observations. Numerous studies that have investigated the sensitivities of burst models to nuclear inputs have commonly extracted the strength of the NiCu cycle in the rp process, determined by the 59 Cu( p , α ) 56 Ni and 59 Cu( p , γ ) 60 Zn thermonuclear reaction rates, as critical in the determination of reaction flow in the burst. In this study, the strength of the cycle at the XRB temperature range was estimated based on published experimental data. The nuclear properties of the compound nucleus 60 Zn were evaluated for the 59 Cu( p , α ) 56 Ni and 59 Cu( p , γ ) 60 Zn reaction rate calculations. Monte Carlo rate calculations were conducted to include the large uncertainties of nuclear properties in the calculations. In the current work, a weak NiCu cycle is expected, whereas the rates adopted by the previous studies suggest a strong NiCu cycle. Model simulations were performed with the new rates to assess the impact on Type I XRBs. The results show that the estimated cycle strength does not strongly influence the model predictions of the burst light curve or synthesized abundances.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2023

Nuclear Data Sheets for A=245

Available information pertaining to the nuclear structure of ground and excited states for all known nuclei with mass numbers A=245 have been compiled and evaluated. In this work, the adopted level and decay schemes, as well as the detailed nuclear properties and configuration assignments based on experimental data, are presented for these nuclides. When there are insufficient data, expected values from systematics of nuclear properties and/or theoretical calculations are utilized. Unexpected or discrepant experimental results are also noted. Since the last evaluation, new rotational bands in 245 Pu have been observed to high spin from transfer reaction 244 Pu( 209 Bi, 208 Biγ) by 2014Ho16. While in 245 Am, the 249 Bk α decay measurements by 2013Ah03 added a new rotational band based on π3/2[521] configuration. The 245 Pu β - decay scheme is basically what was proposed in the late 60's and 47 γ rays have still not been placed. In 245 Cm, precise alpha energies and intensities were measured by 2015Ah03 in 249Cf α decay. Their reported alpha energies of the main α group were about 2 keV lower than in the previous evaluation. The spin and parity of the g.s. of 245 Es is still ambiguous as more detailed information on the decay is needed. With the production of 249 No via fusion evaporation reaction by groups in Dubna and GSI, it's alpha decay was studied and some properties of the daughter 245 Fm were published and included in this evalaution. Information on the other nuclides is scarce and not much has been investigated since the last evaluation. A summary and compilation of the discovery of the various nuclides in this mass region is given in: 2013Fr02 ( 245 Pu, 245 Am, 245 Cm, 245 Bk, 245 Cf), 2011Me01 ( 245 Es), and 2013Th02 ( 245 Md, 245 Fm).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Impact of nuclear data covariance libraries on uncertainty quantification of sodium cooled fast reactor simulation

Uncertainty quantification in core modelling to assess reactor performance is a regular practice to identify design safety limits. Best-estimate plus uncertainty (BEPU) methods are commonly used to propagate uncertainties to core responses from input parameters, characterized by variance-covariance libraries. This study evaluates the impact of different covariance libraries on the uncertainty associated with core safety parameters and their propagation through reactor transient modelling for a metallic core (ABR-1000). The COMAC-1.0, COMMARA-2.0, and COMMARA-2.1 covariance libraries are used to propagate nuclear data uncertainties to neutron feedback coefficients using the Generalized Perturbation Theory, then through transient safety analyses to establish the confidence interval for safety performance of the reactor using stochastic sampling. The considered transients include Unprotected Transient Over Power and Unprotected Loss Of Flow. Most influential nuclide and reaction pairs in different variance-covariance libraries are identified and their impact on the uncertainty of safety parameters are evaluated. Significant differences were observed between uncertainties from COMAC compared to COMMARA libraries. Uncertainties from COMMARA libraries were consistently larger than COMAC for all cases. For structure, coolant density feedback coefficients and control rod worth, total nuclear data uncertainties were approximately 50% larger for COMMARA than COMAC. Differences were also observed in identification of top contributors of uncertainty in terms of reaction-cross sections. For instance, largest contribution to total uncertainty of radial expansion coefficient from COMMARA-2.0 and 2.1 originates from Na{sup 23} P1 Elastic cross section (0.81%) but from Pu{sup 239} Fission (0.53%) with COM C-1.0. Such differences are evaluated in this research for all quantities of interest. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Nuclear Data Sheets for A=47*

Spectroscopic data for all nuclei with mass number A=47 have been evaluated and the corresponding level schemes from radioactive decay and reaction studies are presented. Highlights from this evaluation include first observation of gamma-ray transitions in 47 Ar, a new high precision measurement of 47 K β decay, and first observation of excited levels and gamma-ray transitions in 47 Mn.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Nuclear Data Sheets for A=47

Spectroscopic data for all nuclei with mass number A=47 have been evaluated and the corresponding level schemes from radioactive decay and reaction studies are presented. Furthermore, highlights from this evaluation include first observation of gamma-ray transitions in 47 Ar, a new high precision measurement of 47 K β decay, and first observation of excited levels and gamma-ray transitions in 47 Mn.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Impact of Glass Irradiation on Laser-Induced Breakdown Spectroscopy Data Analysis

Increased absorption of optical materials arising from exposure to ionizing radiation must be accounted for to accurately analyze laser-induced breakdown spectroscopy (LIBS) data retrieved from high-radiation environments. We evaluate this effect on two examples that mimic the diagnostics placed within novel nuclear reactor designs. The analysis is performed on LIBS data measured with 1% Xe gas in an ambient He environment and 1% Eu in a molten LiCl-KCl matrix, along with the measured optical absorption from the gamma- and neutron-irradiated low-OH fused silica and sapphire glasses. Significant changes in the number of laser shots required to reach a 3σ detection level are observed for the Eu data, increasing by two orders of magnitude after exposure to a 1.7 × 1017 n/cm2 neutron fluence. For all cases examined, the spectral dependence of absorption results in the introduction of systematic errors. Moreover, if lines from different spectral regions are used to create Boltzmann plots, this attenuation leads to statistically significant changes in the temperatures calculated from the Xe II lines and Eu II lines, lowering them from 8000 ± 610 K to 6900 ± 810 K and from 15,800 ± 400 K to 7200 ± 800 K, respectively, for exposure to the 1.7 × 1017 n/cm2 fluence. The temperature range required for a 95% confidence interval for the calculated temperature is also broadened. In the case of measuring the Xe spectrum, these effects may be mitigated using only the longer-wavelength spectral region, where radiation attenuation is relatively small, or through analysis using the iterative Saha–Boltzmann method.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Validation of Jezebel Reactivity Coefficients and Sensitivity Analysis

Nuclear data validation is often performed today using criticality measurements. The gold standard for criticality measurements is the International Criticality Safety Benchmark Experiment Project (ICSBEP). The validation specifically focuses on the effective multiplication factor ($k_{eff}$). $K_{eff}$ is a relatively easy parameter to infer and has reduced uncertainty due to being at or above critical. However, while $k_{eff}$ is the most documented parameter and its uncertainties and sensitivities have been evaluated in great detail, it cannot be used as a standalone metric to determine inaccuracies in nuclear data (e.g., cross section data, PFNS, nu), which is based on theory, physics, and differential measurements. The Experiments Underpinned by Computational Learning for Improvements in nuclear Data (EUCLID) project aims to identify compensating errors in specific isotope nuclear data by optimally designing experiments that are, or are not sensitive to a suite of measurement parameters beyond $k_{eff}$. By identifying parameters that are sensitive to each other, oppositely sensitive, or have substantial magnitude differences in sensitivity, experiments can be designed to constrain questionable nuclear data. One sensitivity that is of particular interest to this project includes the sensitivity of reactivity coefficients. Reactivity coefficients compare reactivity, which is related to $k_{eff}$ at two different states therefore being sensitive to small changes in the system. The most common type of reactivity coefficient measurements is comparison to void for a small sample within the assembly. It is key that the sample sizes are small enough to not affect the flux of the full system. Reactivity coefficients were evaluated for many early experiments to better understand transport corrected cross sections. In fact, ICSBEP includes reactivity coefficient results as “Supplemental Measurements” in appendices for a handful of older benchmarks. One of those benchmarks is Jezebel, the bare Pu critical assembly. This paper compares new simulations of reactivity coefficients for Jezebel, and explores the sensitivity of reactivity coefficients to small changes in nuclear data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

SNL-NJOY-2016

This is a wrapper around the LANL NJOY-2016 code that interfaces with extended code capabilities and minor modification of the LANL NJOY-2016 code that supports modeling radiation damage to materials. The NJOY Nuclear Data Processing System is a modular computer code designed to read evaluated data in ENDF format, transform the data in various ways, and output the results as libraries designed to be used in various applications. The wrapper provided here permits Sandia-specific control parameters to be used in the input data file. The modifications incorporated here enhance the ability of NJOY to address material damage response functions for many materials, e.g. to include the NRT and arc-dpa forms of the damage energy in addition to the default NJOY-2016 implementation of the sharp-threshold Kinchin-Pease threshold energy treatment. All of the modifications provided here are being made available to the GitHub-based NJOY-2016 code. As useful enhancements found here are incorporated into the baseline NJOY-2016 code, they will be eliminated from this version so as to maintain our compatibility with the baseline NJOY-2016 code. SAND2020-13060 M Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA00035

Griffin, Patrick↗
DOE OSTI2024

Benchmark Calculation for the Peach Bottom Unit 2 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package

In this study, benchmark calculations were performed for Peach Bottom Unit 2 cycles 1–3 to validate the SCALE 6.3/Polaris–PARCS v3.4.2 with the ENDF/B-VII.1 AMPX 56-group library by comparing the simulated results with the measured data. The benchmark results will be used to evaluate uncertainties of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis for key nuclear parameters such as reactivity and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris, GenPMAXS and PARCS. Additionally, detailed information is provided for all the input and output files produced for the benchmark calculations. The benchmark results were summarized such that they can be used to evaluate uncertainties with other benchmark results for key nuclear parameters.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Benchmark Calculation for the Quad Cities Unit 1 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package

In this study, benchmark calculations were performed for the Quad Cities Unit 1 cycles 1–3 to validate the SCALE 6.3/Polaris–PARCS v3.4.2 code package with the ENDF/B-VII.1 AMPX 56-group library by comparing the simulated results with the measured data. The benchmark results will be used in evaluating uncertainties of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis for key nuclear parameters such as reactivity and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris, GenPMAXS, and PARCS; additionally, detailed information is provided for all the input and output files produced for the benchmark calculations. The benchmark results are summarized herein so that they can be used to evaluate uncertainties with other benchmark results for key nuclear parameters.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Evaluating the 238 U PFNS Including Chi-Nu Experimental Data

This report documents an evaluation of 238 U prompt fission neutron spectra (PFNS) which is a deliverable for a FY2024 Q4 NCSP (Nuclear Criticality Safety Program) milestone. This evaluation is new; its prior input is based on extended Los Alamos and exciton models implemented in the code CoH. Experimental covariances were estimated for five experimental data sets. One of these data sets that was measured by the Chi-Nu team of LANL and LLNL. It covers the 238 U PFNS for continuous incident-neutron energies of 1–20 MeV and outgoing-neutron energies from 10 keV– 10 MeV with high precision. Contrary to Chi-Nu data, previous data sets were measured in a limited energy range. The resulting evaluated data correspond well to the experimental PFNS taken into account for the evaluation. The evaluated PFNS also produce average mean energies in agreement with associated Chi-Nu data. If one uses the new evaluated data to predict the neutron multiplication factor, k eff , of the Flattop, Flattop-Pu and BigTen ICSBEP critical assemblies (which all have thick reflectors with high percentages of 238 U), the differences of simulated values compared to those using ENDF/B-VIII.1β3 is modest (less than 25 pcm). In addition to that, the new PFNS predict on average 238 U LLNL pulsed-sphere neutron-leakage spectra slightly better than ENDF/BVIIII.0 and ENDF/B-VIII.1β3 PFNS. The differences are, however, well within the experimental uncertainties.

238U↗
DOE OSTI2021

Nuclear Science User Facilities FY 2020 Program Performance Summary

The Nuclear Science User Facilities (NSUF) first gathered and reported metric data for Fiscal Year (FY) 2014. The goal of the FY 2014 NSUF Metric Report was to establish metric measures and to provide baseline data to evaluate future performance of the NSUF. Each metric goal and objective was established through NSUF discussions with Department of Energy (DOE) Idaho Operations Office and Office of Nuclear Energy staff. This report provides the data, analysis, and conclusions for the FY 2020 metrics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Nuclear Science User Facilities FY 2019 Program Performance Summary

The Nuclear Science User Facilities (NSUF) first gathered and reported metric data for Fiscal Year (FY) 2014. The goal of the FY 2014 NSUF Metric Report was to establish metric measures and to provide baseline data to evaluate future performance of the NSUF. Each metric goal and objective were established through NSUF discussions with Department of Energy (DOE) Idaho Operations Office and Office of Nuclear Energy staff. This report provides the data, analysis, and conclusions for the FY 2019 metrics.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

IRDFF-II: A New Neutron Metrology Library

High quality nuclear data is the most fundamental underpinning for all neutron metrology applications. This paper describes the release of version II of the International Reactor Dosimetry and Fusion File (IRDFF-II) that contains a consistent set of nuclear data for fission and fusion neutron metrology applications up to 60 MeV neutron energy. The library is intended to support: a) applications in research reactors; b) safety and regulatory applications in the nuclear power generation in commercial fission reactors; and c) material damage studies in support of the research and development of advanced fusion concepts. The paper describes the contents of the library, documents the thorough verification process used in its preparation, and provides an extensive set of validation data gathered from a wide range of neutron benchmark fields. The new IRDFF-II library includes 119 metrology reactions, four cover material reactions to support self-shielding corrections, five metrology metrics used by the dosimetry community, and cumulative fission products yields for seven fission products in three different neutron energy regions. In support of characterizing the measurement of the residual nuclei from the dosimetry reactions and the fission product decay modes, the present document lists the recommended decay data, particle emission energies and probabilities for 68 activation products. It also includes neutron spectral characterization data for 29 neutron benchmark fields for the validation of the library contents. Additional six reference fields were assessed (four from plutonium critical assemblies, two measured fields for thermal-neutron induced fission on 233U and 239Pu targets) but not used for validation due to systematic discrepancies in C/E reaction rate values or lack of reaction-rate experimental data. Another ten analytical functions are included that can be useful for calculating average cross sections, average energy, thermal spectrum average cross sections and resonance integrals. The IRDFF-II library and comprehensive documentation is available online at www-nds.iaea.org/IRDFF/. Evaluated cross sections can be compared with experimental data and other evaluations at www-nds.iaea.org/exfor/endf.htm. The new library is expected to become the international reference in neutron metrology for multiple applications.

IAEA IRDFF↗
DOE OSTI2025

U.S. Efforts in Support of Examinations at Fukushima Daiichi - September 2024 Meeting Notes

Information obtained from Fukushima Daiichi Nuclear Power Station (Daiichi) is required to inform future Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. Commercial nuclear power plants. A collaborative U.S. and Japanese effort was initiated in 2014 by the Department of Energy Office of Nuclear Energy to identify Daiichi examination needs and evaluate recent Daiichi examination data to address these needs. This document summarizes information presented at and findings, action items, and recommendations by U.S. and Japanese experts in reactor safety and plant operations during the September 2024 Forensics Effort meeting. Significant safety insights were obtained in several areas: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties and knowledge gaps in severe accident modeling progression, these insights continue to be used to assess whether additional updates are needed in guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, Daiichi-related activities, such as code modeling improvements and analysis, testing, and new technology deployment efforts, have the potential to offer additional safety and economic benefits to the operating fleet and new light water reactor (LWR) and non-LWR designs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Chlorine Worth Study Nuclear Data

The Chlorine Worth Study (CWS) was a series of experiments that took place at the National Criticality Experiments Research Center (NCERC), operated by Los Alamos National Laboratory (LANL). The focus of the experiments was to develop new integral benchmarks for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook with high sensitivity to chlorine in the thermal neutron energy region and which match sensitivities of aqueous chloride operations at LANL. This work discusses the experiment and how the experimental and simulated results using different nuclear data libraries compare to each other.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗