Verified, Archived Library of Inputs and Data (VALID) Online Repository
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Nuclear Criticality Safety Program (NCSP)↗
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Results for “nuclear data library”
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Impact of T - and ρ -dependent decay rates and new (n, γ ) cross-sections on the s process in low-mass asymptotic giant branch stars
Aims. We study the impact of nuclear input related to weak-decay rates and neutron-capture reactions on predictions for the slow neutron-capture process (s process) in asymptotic giant branch (AGB) stars. We provide the first database of surface abundances and stellar yields of the isotopes heavier than iron from the Monash models. Methods. We ran nucleosynthesis calculations with the Monash post-processing code for seven stellar structure evolution models of low-mass AGB stars with three different sets of nuclear inputs. The reference set has constant decay rates and represents the set used in the previous Monash publications. The second set contains the temperature and density dependence of β decays and electron captures based on the default rates of nuclear NETwork GENerator (NETGEN). In the third set, we further update 92 neutron-capture rates based on re-evaluated experimental cross sections from the ASTrophysical Rate and rAw data Library. We compare and discuss the predictions of the sets relative to each other in terms of isotopic surface abundances and total stellar yields. We also compare the results to isotopic ratios measured in presolar stardust silicon carbide (SiC) grains from AGB stars. Results. The new sets of models result in a ∼66% solar s-process contribution to the p-nucleus 152 Gd, confirming that this isotope is predominantly made by the s process. The nuclear input updates result in predictions for the 80 Kr/ 82 Kr ratio in the He intershell and surface 64 Ni/ 58 Ni, 94 Mo/ 96 Mo, and 137 Ba/ 136 Ba ratios that are more consistent with the corresponding ratios measured in stardust; however, the new predicted 138 Ba/ 136 Ba ratios are higher than the typical values of the SiC grains. The W isotopic anomalies are in agreement with data from the analyses of other meteoritic inclusions. We confirm that the production of 176 Lu and 205 Pb is affected by too large uncertainties in their decay rates from NETGEN.
79 ASTRONOMY AND ASTROPHYSICS↗
Decarbonizing Industrial Heat and Electricity Applications Using Advanced Nuclear Energy
Idaho National Laboratory (INL) is investigating the technical pathways to assist industrial heat and electricity users to meet their decarbonization goals through integration with advanced nuclear power plants (NPPs). This project will deliver a library of process models and accompanying documents that guide specific industries in choosing potential nuclear technologies based on their needs. Considerations in providing this guidance include specific hazards from the industrial facility, heat transport requirements and associated technologies, and feasibility with site-specific demand profiles. The library of facility process models will be based on real data from industrial facilities in the United States. The industrial processes will be identified in this project based on the following: (1) operational heat characteristics that nuclear systems can provide, (2) sufficient energy requirements to merit the capital investment for nuclear plant construction, and (3) environmental benefits of replacing existing energy production with carbon-free nuclear power. Other decarbonization opportunities considered are the addition of nuclear-powered electrolysis processes or high-temperature electric heating where the thermal requirements exceed nuclear generation conditions. In addition to assessing the technical feasibility, INL is evaluating the impact of hazards introduced by the industrial facilities on the siting requirements of advanced NPPs. Site characterization of an industrial plant is essential to determine the feasibility and suitable integration methods for each industry. The assessment of siting and technical data will reveal opportunities for a single-use nuclear integration as well as integration of multiple industrial facilities with a single NPP.
02 PETROLEUM↗
Extending the Nuclide Inventory Validation Basis for High-Burnup Fuel with New Radiochemical Assay Data
Efforts are underway at Oak Ridge National Laboratory to improve the nuclide inventory validation basis for spent nuclear fuel at high burnups. Recently conducted radiochemical assay experiments provided new measurement data for nine samples of fuel irradiated in a pressurized water reactor, with estimated sample burnups in the 30 to 70 GWd/t range. This type of destructive assay data is essential for validating computational methods, tools, and nuclear data applied in nuclear safety analyses and for improving our understanding of the bias and uncertainty in code predictions. The measurement data include key actinides and fission products that span a gamut of needs and interests for nuclear science and engineering applications in criticality safety, reactor physics, nuclide inventory, decay heat, and radiation shielding. The SCALE 6.3 code system with ENDF/B-VII.1 cross-section libraries was used to simulate the irradiation histories of the measured fuel samples. The calculated nuclide concentrations are compared to corresponding measurement data. The significance of the comparisons is discussed, emphasizing how the addition of the new measurement data fills gaps in the validation basis at high burnups and contributes to the decrease in bias and uncertainty for predicted nuclide concentrations. The discussion addresses the effect of the sample burnup used in the simulation—which is based on reactor operator records or on calibration to measured data for burnup indicator fission products—on the validation results.
Nuclide inventory↗
Beta-delayed gamma spectra compilation and analysis following the thermal-neutron induced fission of 235 U, 239,241 Pu
The integral gamma and electron spectra emitted by fission products, also known as delayed gamma and electron spectra, were measured at Oak Ridge National Laboratory in the 1970s for the thermal-neutron induced fission of 235 U and 239,241 Pu. Scintillator detectors were used to measure these spectra, data used later on to obtain decay heat values - that is, the spectra mean values per unit time as function of time - work that was published in the Nuclear Science and Technology journal; the spectral data, however, was only published in laboratory reports. Here, in this work, we analyze the gamma spectra data using modern methods and nuclear databases to reveal the signature of individual fission products as well as to gauge the performance of the ENDF/B-VIII.0 decay data sub-library, concluding about possible future enhancements in predictive capabilities.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Adapting CLUTCH methodology to multigroup TSUNAMI-3D for eigenvalue sensitivity calculations
The sensitivity of the eigenvalue to uncertainties in nuclear data and its evaluation are important for nuclear criticality safety. TSUNAMI-3D sequences within the SCALE code system offer several options to the user community for calculating eigenvalue sensitivity coefficients with multigroup (MG) and continuous energy (CE) 3D transport capabilities. TSUNAMI-3D sequences implement the adjoint-based perturbation theory with MG KENO code, the Contributon Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance CHaracterization (CLUTCH) method with CE KENO code, and the Iterated Fission Probability (IFP) method with CE KENO and Shift codes. Each method has benefits and limitations depending on the problem that is run. The work presented here aims to adapt the CLUTCH method, which enables the Contributon method's mesh-free, memory-efficient approach for calculating adjoint-weighted tallies for sensitivity calculations, to the MG TSUNAMI-3D sequence. This application would eliminate the explicit adjoint KENO calculation, as well as the memory-consuming mesh flux moment tallies required by the conventional MG TSUNAMI-3D. Smaller memory footprints in the CLUTCH methodology and relatively shorter runtimes in MG KENO transport can make MG TSUNAMI-3D a viable method for some complex problems. Moreover, this adaptation allows MG sensitivity calculations with Shift, ORNL's next-generation high-performance Monte Carlo transport code, which currently does not offer any sensitivity capabilities with MG particle transport simulations. Initial implementation of the new MG TSUNAMI-3D sequence and its preliminary results with a selected critical benchmark experiment in the Verified, Archived Library of Inputs and Data (VALID) are presented in this study.
KENO↗
Formatting and V&V of Consistent 238,240−24 2Pu $\overline{v}_p$ Evaluated Mean Values and Covariances
This report is in answer to the Nuclear Criticality Safety Program FY24 quarter 4 milestone that requires: “Format and V&V nu-bar means and covariances” for 238,240-242 Pu average prompt fission neutron multiplicities, $\overline{ν}$ p , that were obtained by a consistent evaluation leveraging the fission-event generator CGMF and a detailed uncertainty quantification of experimental data. It is described how nuclear data mean values and covariances were formatted using ENDFtk. Implementing the new 238,240-242 Pu $\overline{ν}$ p into the ENDF/B-VIII.1β 4 library leads to only small overall changes in criticality values of the Jezebel, Dirty Jezebel, Jupiter-001, Jupiter-002, EUCLID 3x2 and EUCLID 8x1 critical assemblies. Simulated k eff uncertainties due to $\overline{ν}$ p covariances change only little if cross-isotope covariances are considered or not for those assemblies with low percentage content of minor Pu isotopes. However, for the Dirty Jezebel critical assembly, that has a sizeable 240 Pu and non-negligible 241 Pu content, the simulated k eff uncertainties due to considering or neglecting cross-isotope $\overline{ν}$ p covariances is 443 versus 374 pcm.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Low-energy 17 O(𝑛,𝛾) 18 O reaction within the microscopic potential model and its role for the weak 𝑟 process
The neutron radiative capture reaction 17 O (𝑛,𝛾) 18 O plays a pivotal role in both nuclear structure studies and astrophysical nucleosynthesis, particularly in the formation of elements during hydrostatic and explosive stellar environments. We calculated the 17 O (𝑛,𝛾) 18 O cross section within the Skyrme Hartree-Fock potential model and analyzed electric dipole 𝐸1 transitions to both positive- and negative-parity states below the α-decay threshold in 18 O. Our cross sections are significantly different from the data available in commonly used libraries. We further investigate the impact of the new calculated cross section on weak 𝑟-process nucleosynthesis using large-scale reaction network calculations across a wide range of electron fractions and entropies. Our results show that the 17 O (𝑛,𝛾) 18 O reaction rate significantly influences the production of first 𝑟-process peak elements, such as strontium, under specific astrophysical conditions. This study highlights the importance of accurate nuclear data for light isotopes in modeling heavy-element synthesis and provides updated reaction rates for future nucleosynthesis simulations.
6 ≤ A ≤ 19↗
NuMaCo
The Nuclear Material and Composition (NuMaCo) server is a web-based platform for storing, managing, and serving material composition data. It includes the SCALE Standard Composition Library and the PNNL Compendium, along with their associated references and uncertainty data. NuMaCo also provides tools to generate, compare, and validate material input definitions for both SCALE and MCNP.
Skutnik, Steve (000000016441135X)↗
Survey of Neutron Induced Fission Experimental and Evaluated Data for 233 U, 238 Pu, 240 Pu, and 242 Pu between 100 keV and 20 MeV
We review data and evaluations for neutron-induced fission for the actinides 233 U, 238 Pu, 240 Pu and 242 Pu. These isotopes are part of common nuclear fuel cycles, especially for modern fast reactors. We focus on incident neutrons in the energy range of 100 keV to 20 MeV and compare the experimental data from the literature to the major evaluated libraries ENDF/B-VIII.1, JENDL-5, and JEFF-3.3, as well as to LLNL’s ENDL-2009.5 library. Based on the assessment of the fission cross sections, we provide recommendations on which library to use. Additional assessments covering other reaction channels will be provided separately.
07 ISOTOPE AND RADIATION SOURCES↗
Coupling RELAP5-3D to BISON
This report details two approaches for coupling the BISON nuclear fuel performance code with RELAP5-3D. Both approaches are shown to work well. The first is a Python-based approach, and the second combines the two applications into one executable with data passed in memory from one library to the other. This second coupling approach forms the basis for analysis of complex loss of coolant scenarios and enables future calculations of modeling uncertainties.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Uncertainty Quantification for Data-Driven Machine Learning Models in Nuclear Engineering Applications: Where We Are and What Do We Need?
Machine learning (ML) has been leveraged to tackle a diverse range of tasks in almost all branches of nuclear engineering. Many of the successes in ML applications can be attributed to the recent performance breakthroughs in deep learning, the growing availability of computational power, data, and easy-to-use ML libraries. However, these empirical successes have often outpaced our formal understanding of the ML algorithms. An important but under-rated area is uncertainty quantification (UQ) of ML. ML-based models are subject to approximation uncertainty when they are used to make predictions, due to sources including but not limited to, data noise, data coverage, extrapolation, imperfect model architecture and the stochastic training process. The goal of this paper is to clearly explain and illustrate the importance of UQ of ML. We will elucidate the differences in the basic concepts of UQ of physics-based models and data-driven ML models. Various sources of uncertainties in physical modeling and data-driven modeling will be discussed, demonstrated, and compared. We will also present and demonstrate a few techniques to quantify the ML prediction uncertainties, including Monte Carlo dropout, deep ensemble, Bayesian neural networks, Gaussian Processes and conformal prediction. Lastly, we will discuss the need for building a verification, validation and UQ framework to establish ML credibility.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Benchmark Calculation for BEAVRS Cycles 1 and 2 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package
This study performed the benchmark calculation for BEAVRS to validate the SCALE 6.3/Polaris–PARCS v3.4.2 code package with the ENDF/B-VII.1 56-group library, 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 pressurized water reactor physics analysis for key nuclear parameters such as reactivity, control bank worth, temperature coefficients, and pin and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris as well as 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 in evaluating uncertainties for key nuclear parameters with other benchmark results.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Benchmark Calculation for Surry Unit 1 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package
The benchmark calculations were performed for Surry Unit 1 cycles 1–3 to validate the SCALE 6.3/Polaris–Purdue Advanced Reactor Core Simulator (PARCS) v3.4.2 with the ENDF/B–VII.1 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 pressurized water reactor physics analysis for key nuclear parameters such as reactivity, control bank worth, temperature coefficients, and pin and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris, GenPMAXS, and PARCS. Additional details are provided for the input and output files produced for the benchmark calculations. The benchmark results were summarized such that they can be used in evaluating uncertainties with other benchmark results for key nuclear parameters.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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↗
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↗
Benchmark Calculation for the Hatch Unit 1 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package
This study was the performance of the benchmark calculation for the Hatch Unit 1 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 in evaluating uncertainties of the SCALE/Polaris–PARCS code package for boiling water reactor (BWR) 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, and 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 in evaluating uncertainties for key nuclear parameters with other BWR benchmark results.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Evaluating 240 Pu Prompt Fission Neutron Spectra Including Chi-Nu Experimental Data
This report documents an evaluation of 240 Pu prompt fission neutron spectra (PFNS) including Chi-Nu experimental data. This work is a FY2025 Q4 NCSP (Nuclear Criticality Safety Program) milestone and named a “strategic priority”. This evaluation is new; it differs substantially from the most recent evaluation included in the ENDF/B library, which was for ENDF/B-VII.1 and then carried over unchanged to ENDF/B-VIII.1. The key difference between ENDF/B-VIII.1 and the new evaluation is that we have, for the first time, realistic experimental 240 Pu PFNS covering a broad incident and outgoing neutron energy range.
240Pu↗