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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

ENDF/B-VIII.1

The ENDF/B-VIII.1 release is the newest evaluated nuclear data library produced, distributed, and recommended by CSEWG for use in nuclear science and technology applications. Among the many key advances, relative to the previous version ENDF/B-VIII.0, are: re-evaluation of 239Pu file by a joint international effort; updated 16,18O, 19F, 28-30Si, 50-54Cr, 55Mn, 54,56,57Fe, 63,65Cu, 139La, 233,235,238U, and 240,241Pu neutron nuclear data by the IAEA-coordinated INDEN collaboration; significant changes for 3He, 6Li, 9Be, 51V, 88Sr, 103Rh, 140,142Ce, Dy, 181Ta, Pt, 206-208Pb, and 234,236U neutron data; new nuclear data for the photo-nuclear, being 196 adopted from the IAEA2019 Photonuclear Data Library and one new file from JENDL-5; and new evaluations for the charged-particle and atomic sublibraries. Numerous thermal neutron scattering kernels were re-evaluated or provided for the very first time. Additionally, new covariance testing was implemented. ENDF/B-VIII.1 reduced bias in the simulations of many integral experiments with particular progress noted for fluorine, copper and stainless steel containing benchmarks. Data issues which had hindered the deployment of ENDF/B-VIII.0 for commercial nuclear power applications in high burn-up situations, were addressed. ENDF/B-VIII.1 data are distributed in both ENDF-6 and GNDS formats.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Neutron Reaction Sublibrary

The neutrons sublibrary aims to describe nuclear reactions between incident neutron particles and different nuclei. For ENDF/B-VIII.1, many neutron files were re-evaluated or received major changes. A new 239Pu evaluation was jointly-produced by IAEA, LANL, LLNL and ORNL bringing important updates to fission neutron multiplicity, Prompt Fission Neutron Spectrum, resonance and fast regions. Around one third of the new neutron evaluations were performed as part of the INDEN collaboration, including 16,18 O, 19 F, 28,29,30 Si, 63,65 Cu, 50,51,52,53,54 Cr, 55 Mn, 54,56,57 Fe, 139 La, 233,235,238 U, 240,241 Pu. Important non-INDEN evaluations include 234,236 U, 206,207,208 Pb, 181 Ta, 88 Sr, 140,142 Ce, Pt and Dy isotopes, and many others. Also, dosimetry reactions from IRDFF-II were adopted for many materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Characterization of Uranium Foil Irradiations at the WSU TRIGA Reactor Using a New Reactor Model in Scale

A new reactor model of the Washington State University TRIGA was developed in the SCALE neutron transport code, and its fidelity was verified by comparison to MCNP and available data for several reactor parameters. The model was used to characterize irradiations designed to produce the short-lived actinides 237 U and 239 U, two key isotopes for nuclear forensics. These short-lived actinides, their decay daughters 237Np and 239Pu, and total fissions (via 99 Mo) were measured in irradiated foils at Los Alamos National Laboratory and other labs with good agreement among parent/daughter pairs and among labs. The laboratory-measured isotope ratios were used as a benchmark for the model determination of reaction rates. The flux distribution in the foils was also determined. Finally, using the continuous energy TSUNAMI-3D module of SCALE, a sensitivity/uncertainty analysis was performed to determine the effects on foil reactions caused by other system-wide reactions and by uncertainties in the known cross-sections.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Fission Product Yields of 233 U and 239 Pu by Neutron-Induced Fission at Neutron Energies from 0.18-140 MeV [Slides]

Nuclear fission is an important process with applications in astrophysics, nuclear reactors, and stockpile stewardship. Despite having been discovered over 80 years ago, fission is not fully described by a predictive model. This study will show fission product mass yields and total kinetic energies (TKE) across a large range of incident neutron energies for future refinement of fission models. 233U(n,f) data is used to verify features and structures on a previous experiment and compared against fission models. 239Pu(n,f) data will be analyzed to show new data at higher neutron energies using techniques to handle fission in the presence of alphas. 252Cf(sf) data will be used for calibration and benchmarking with a thin-backed target.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Interaction of Concentrated (68%) Aqueous Nitric Acid Solutions with Cheesecloth: A TGA-DSC-MS Study

The exposure of cotton cheesecloth to concentrated aqueous solutions of nitric acid reduced the onset of combustion by 30-40°C (from ~350°C to ~310°C) relative to that of both unexposed cheesecloth and cheesecloth that was rinsed immediately following exposure to the same solution. Cheesecloth exposed to concentrated (68%) nitric acid underwent three exothermic pre-combustion reactions which onset at 125-135°C, 160- 170°C and 200-210°C. These reactions give off mostly water vapor, with lesser amounts of NO and carbon dioxide. Exposure to concentrated nitric acid damages the cellulose polymers, oxidizing them, reducing them in length, and increasing their reactivity in response to thermal stress. Cheesecloth which was rinsed after exposure behaved similarly to unexposed cheesecloth, indicating that NC was not formed on exposure to the concentrated aqueous nitric acid solution. The addition of a 239Pu – 241Am spike in an aqueous nitric – hydrofluoric solution to the cheesecloth reduced the onset temperature of the first exothermic reaction to 126°C and lowered the onset temperature of combustion from ~310°C to ~275°C. When dried on to cellulose substrates in the presence of excess nitrate, metal ions may act as oxidizers, or may catalyze the oxidation of the carbonaceous material, lowering ignition temperatures. No evidence for the formation of nitrocellulose following the exposure of cheesecloth to nitric acid was observed during this study

36 MATERIALS SCIENCE↗
DOE OSTI2023

Ion Mobility – Mass Spectrometry Rapid Imaging of Special Nuclear Materials

Robot Nose will use new fast data throughput methods and experience with laser microfocus to dramatically improve the speed and spatial resolution of an established instrument, making it a useful tool for imaging special nuclear materials (SNM). Mass spectrometry imaging instruments have traditionally been too slow and mass alone is generally an insufficient measurement for organic analyses. This situation was markedly improved by Ionwerks Inc., who pioneered the combination of ion mobility with mass spectrometry imaging. In 2D mobility-mass plots, clearly identifiable trend lines can be selected and reduced to 1D mass spectra without isobaric interferences. This is a major advantage over other imaging mass spectrometers that generally cannot distinguish isobars without custom-built lasers or high voltage accelerators. Special nuclear material (SNM) contains substantial fissile isotopes, particularly 233U, 235U, and 239Pu. A complete forensic analysis of SNM therefore requires quantitative isotopic measurement – often through some form of mass spectrometry. SNM contaminated with inclusions such as pollen grains can help identify its prior locations. The project team is presently working on an approach to allow rapid screening of 1-3 cm diameter thin samples of such material.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

H-Canyon Flowsheet for the Neutralization of Dissolved Fast Critical Assembly (FCA) Fuel

The H-Canyon facility will be dissolving unirradiated stainless-steel clad Fast Critical Assembly (FCA) fuel in HNO3 and neutralizing the resulting solution with 50 wt% NaOH with no recovery operations prior to transfer to the Concentration, Storage, and Transfer Facility (CSTF). There are two types of FCA fuel, including a Pu-Al metal alloy and a mixed U and Pu oxide. H-Canyon anticipates dissolving 16 batches of the metal fuel and two batches of the oxide fuel. Potassium fluoride will be added to the dissolver solution at 0.05 M to promote dissolution. Gadolinium will be added after the dissolution as a thermal neutron poison for criticality control. H-Canyon requested that SRNL evaluate Gd:239Pu equivalent (239PuEq) ratios of 1:1 and 10:1. The neutralization process precipitates metals including actinides, Gd, and cladding components resulting in a slurry. Caustic neutralization is a routine H-Canyon operation, but the FCA dissolver solution will be unique relative to solutions that are typically processed due to the stainless-steel components, high HNO3 concentration of ~8.5 M, and an initial Pu concentration of up to 4 g/L. Previous neutralization studies have been performed for Pu containing solutions but at less than half the initial Pu concentration and much lower initial HNO3 concentrations. Experimental neutralizations targeted a final free hydroxide (OH-) concentration of 0.6 M as this was anticipated to be the final endpoint, but H-Canyon now expects the endpoint to be 1.2 M OH-. The purpose of this study was to characterize the distribution of the actinide, Gd, and cladding components between the precipitate and supernate, determine if the slurry will back up in the header during the transfer to the CSTF, and determine if solids will settle in the pipeline during the transfer.

Mills, Matthew S.↗
DOE OSTI2026

Improbability of Post-Closure Criticality in Array of Transuranic Waste Containers after Compaction by Salt Creep at Waste Isolation Pilot Plant

Based on the rationale presented, post-closure nuclear criticality is improbable when room closure compacts containers disposing transuranic (TRU) waste emplaced at the Waste Isolation Pilot Plant (WIPP), an operating repository in bedded salt in southeastern New Mexico. As described in the original WIPP certification, a qualitative estimate of the probability of post-closure criticality in TRU waste produced from atomic energy defense activities has been low either because remote-handled TRU waste canisters are neutronically isolated by the bedded salt or because the low fissile mass in an array of contact-handled TRU waste drums cannot be compacted sufficiently by room closure from salt creep. These situations are still valid for the majority of TRU waste that is disposed at WIPP without any disposal constraints, as updated herein. This report also qualitatively evaluates the probability of criticality after disposal of TRU waste in pipe overpack containers (POC) where every POC in a shipment may have the maximum 200 fissile gram equivalent of 239Pu content. The probability of criticality for a disposal room filled with POCs is estimated during four representative phases of repository evolution: (1) a large salt block falls onto POCs in the first 20 years, (2) salt creep closes a disposal room in the first 1000 years without brine seepage and subsequent gas generation, which permits maximum compaction, (3) some brine seepage occurs into the closed room, which initiates consumption of the fiberboard (cellulose) impact absorber in the POC in the second 1000 years, and (4) full brine inundation of a room and consumption of all fiberboard thereafter. Salt-block fall in the first phase does not greatly disrupt three tiers of POCs. The compacted spacing of POCs in the later three repository conditions is calculated through high-fidelity, geomechanical modeling. Criticality evaluation of compacted 200-g 239 Pu spheres at the compacted spacing shows that neither 12-inch nor 6-inch POCs are critical after the first 1000 years, the second 1000 years, or thereafter as the sea of reflector material changes to represent the three repository conditions. Specifically, fiberboard and iron isolates 239 Pu while dry, and brine reduces the reactivity when a room is partially and fully inundated. Because POC behavior bounds behavior of other standard TRU waste containers, post-closure criticality caused by room closure compacting containers is omitted in the performance assessments for the 2019 and 2026 WIPP compliance re-certification applications to the US Environmental Protection Agency.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2022

Lattice Confinement Fusion and Fusion-Fast-Fission Energy Source Development

NASA and the US Navy have demonstrated Lattice Confinement Fusion (LCF) and the Fusion-Fast-Fission of natural uranium and thorium. Both methods build upon decades of research, and the latter benefits from conventional nuclear fission and fusion results. However, this nascent technology doesn’t require fissile isotopes and avoids nuclear weapons proliferation concerns from uranium 235U isotopic enrichment or plutonium 239Pu separation. Of particular note, LCF doesn’t need power hungry magnets, lasers, or particle beams. Potentially, it could provide watts to hundreds of kilowatts of electrical power and process heat suitable for space power and high Isp nuclear electric propulsion (NEP) as well as distributed terrestrial power. Finally, there are indications that LCF fast-fission products are more benign than those of conventional fission reactors. Our goal is to scale the reactions and increase the power output through higher temperature operation and increased material masses.

lattice confinement fusion↗
DOE OSTI2019

Characteristics of TRU Fueled SFR Core during the Transition with Partially Loaded LEU Assemblies

The fast reactor (FR) fuels typically require higher fissile content than the thermal reactor fuels because the cross sections of the fissile isotopes are smaller relative to the fertile isotopes and higher neutron leakage from the reactor core. The medium-sized high-burnup sodium-cooled fast reactors (SFRs) in the framework of uranium-plutonium/transuranic (U/Pu or U/TRU) fuel-cycle can be optimized to operate with less than 17 % of Pu/TRU content in the driver fuel while having higher TRU conversion ratios (CR~1.0) [1]. However, possible shortage or inaccessibility to TRU based fuel could be accommodated by the use of uranium-based fuels on demand such as low enriched uranium (LEU) having significantly higher 235U content than used in the commercial light water reactors (LWRs), which is defined as high-assay LEU. The reactor systems having low TRU conversion ratio are commonly categorized as burner reactors, which include LWRs and even can be achieved by FRs with designed to utilize higher TRU fractions (e.g., increased neutron leakage and lower fuel volume fractions). Such reactor systems need to be initially loaded with high excess reactivity, which decreases toward the end of cycle, in order to achieve target burnup conditions. Therefore, in order to offset the excess reactivity, neutron absorbing materials need to be inserted into the core to compensate (e.g., control rods). By reducing the neutron leakage (e.g., increasing the fuel volume fraction or thicker blankets), the TRU conversion ratio in the FR can be increased. If the TRU conversion ratio becomes 1.0 (technically slightly greater than 1.0 to account for process losses during recycle), the reactor is in the condition known as “breakeven”, in which fuel contains the same amount of fissile material after recycling that was used to fuel the reactor. In such a condition, combined with the fact that fission products do not have strong poisoning effects in FRs, the reactivity swing becomes smaller than that of burner type reactors. For breeders, the reactivity can increase toward the end of the cycle because more fissile material is produced than consumed. In this regard, the breakeven FR core is implicitly optimized in a sense by eliminating the large insertion of control elements to offset the reactivity during the normal operation. However, the breakeven condition here is assumed to be achieved only by the TRU based fuel. Possible lack of TRU based fuel needs to be compensated by available LEU resources containing equivalent amount of fissile (i.e., 235U). The use of LEU will have some impact on important safety parameters that need to be evaluated. U-235 has a higher effective delayed neutron fraction than Pu-239. Also, the number of neutrons produced by fissions per neutron absorption (i.e. ?) of 235U in the fast energy range (>1e4 eV) is lower than that of 239Pu, which may lead to the poor fissile conversion when LEU is used as a high-conversion FR fuel. The full-paper presents the study of the impact of partially loaded LEU fuels within TRU fueled breakeven SFR on the transition of nuclide evolutions and core characteristics particularly focusing on the burnup reactivity swing. The full paper analyzes how the large reactivity swing occurs during the LEU loading and addresses possible options to reduce it to the level equivalent to that of the TRU fueled SFR core.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Artificial neural network based isotopic analysis of airborne radioactivity measurement for radiological incident detection

Responders need tools to rapidly detect and identify airborne alpha radioactivity during consequence management scenarios. Traditional continuous air monitor systems used for this purpose compute the net counts in various energy windows to determine the presence of specified isotopes, such as 235U, 239Pu, and 241Am. These calculations rely on having a well-calibrated detector, which is challenging in low-background environments. Here an alternative approach of using artificial neural networks to classify alpha spectra is presented. Two network architectures, fully connected and convolutional networks, were trained to classify alpha spectra into four categories: background and background plus the three isotopes above. Sources were injected into measured background at various fractions of the derived response level (DRL) corresponding to early-phase Protective Action Guides. The convolutional network identifies all sources at 1% of the DRL with average probability of detection of 95% and false alarm probability of 1%. Further, the network identifies sources ranging between 0.25% and 1% of the DRL with higher than 80% probability of detection and lower than 7% false alarm probability. Most significantly, the network performance improves in low-count background conditions, increasing its minimum probability of detection to 93% and reducing the false alarm probabilities to lower than 0.25%. These results show that, once trained on datasets representing a range of detection scenarios, artificial neural networks can accurately identify alpha isotopes of interest without the need for detector calibration.

Woldegiorgis, Surafel F.↗
DOE OSTI2022

Investigating Fission Reaction Rate Ratio Sensitivities [Abstract]

Reaction rate ratios are a measurable parameter for reactor and criticality applications. A number of foil irradiations and fission chamber measurements have been performed for critical assemblies at Los Alamos National Laboratory starting in the 1950’s including (i) Godiva, a bare HEU spherical assembly; (ii) Flattop-25, a spherical assembly consisting of an HEU core and a natural uranium reflector; (iii) Jezebel, a bare 239 Pu assembly; and (iv) Flattop-Pu, a spherical assembly consisting of a 239Pu core and a natural uranium reflector. Fission ratio data for 238 U(n,f)/ 235 U(n,f), 237 Np(n,f)/ 235 U(n,f), 233 U(n,f)/ 235 U(n,f) and 239 Pu(n,f)/ 235 U(n,f) were obtained and reported. The EUCLID (Experiments Underpinned by Computational Learning for Improvements in nuclear Data) project at Los Alamos National Laboratory (LANL) aims to constrain nuclear data by using a suite of measurement types beyond k-effective. Recent investigations include the use of pulsed spheres for nuclear data validation and other measurement methods of interest. One focus of the work is to determine if other methods are complimentary to the critical experiments utilized for nuclear data validation. It is anticipated the investigations will help inform methods that may be utilized in machine learning algorithms for nuclear validation. In order to use a measurement type for nuclear validation, it is necessary to obtain cross-section sensitivities for parameters. This work looks at reaction rate ratio sensitivities with SENSMG and Monte Carlo N-Particle R Code Version 6.21.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Reactivity Coefficient Measurements to Aid in Reducing Compensating Errors in Plutonium Nuclear Data

Compensating errors between several nuclear data observables in a nuclear data library can adversely impact application simulations. The primary goal of the EUCLID project (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) is to reduce compensating errors between fast (0.1–5 MeV) 239Pu nuclear data for prompt fission neutron spectra (PFNS), average prompt fission neutron multiplicities, and neutron induced fission, capture, elastic, and inelastic cross sections. This work will focus on the design and execution of void reactivity coefficient measurements in the EUCLID experiment, performed on the Planet vertical lift critical assembly machine at the National Criticality Experiments Research Center (NCERC). Two different base configurations were designed and measured, one with high neutron leakage, and one with low neutron leakage. Both were primarily made up of plutonium metal (Zero Power Physics Reactor plates) without interstitial moderators and reflected by half-inch aluminum. Design optimization showed that void reactivity coefficient measurements in three locations per configuration was most impactful to reduce nuclear data uncertainties due to the varying impacts from elastic and inelastic scattering, as well as fission and capture. The locations for measurements were chosen based on preliminary studies which balanced measurement uncertainty and measurement practicality. The measurements were also selected to have sensitivities maximally complementary to previous arrangements. Comparisons across nuclear data libraries highlight the potential impact.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Quasi-differential neutron induced neutron emissions from 235 U, and 239 Pu

Uncertainty in nuclear reaction cross sections, angular distributions, and other nuclear data directly impact how well simulations of nuclear systems represent physical observations. Here, to determine how well the nuclear data in ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0 evaluations describe the physical behaviour of 235 U, and 239 Pu when subjected to a neutron flux, the neutron emission spectrum was measured for carbon, 93.0% 235 U, and 93.9% 239 Pu samples, and compared against detailed MCNP6 simulations. The measurements were performed at the Los Alamos Neutron Science Center using a quasi-differential method previously developed at Rensselaer Polytechnic Institute. The measurement spanned 0.82–20 MeV and 30–150 degrees. The measurements show there are a significant number of discrepancies between library predictions of the neutron yield and physical observation. A few of the main discrepancies found are described in this paper. Based on these results a new evaluation utilizing these results for carbon, 235 U, and 239 Pu is recommended.

235U↗
DOE OSTI2023

Use of Compton suppression gamma ray spectrometry to determine 239 Pu

We report 239 Pu has been characterized by gamma-ray spectroscopy for a level higher than the usual environmental ones typically needing alpha spectrometry, ICP-MS or TIMS. The higher activities typically have applications in reprocessing, safeguards verification, and nuclear forensics. For activities of hundreds of Bq/g, it is possible to use gamma ray spectrometry for 239 Pu characterization. A soil reference standard of 185 Bq/g (75 ng/g) was measured on a 79% efficient HPGe with a digital Compton suppression system. The ratios of the counting statistics for suppressed and unsuppressed uncertainties for various 239 Pu photopeaks were determined. The effects of self-attenuation on the low energy gamma rays were evaluated, and it was determined that Compton suppression was an effective characterization method of 239 Pu after accounting for self-attenuation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Applying a Template of Expected Uncertainties to Updating 239 Pu(n,f) Cross-section Covariances in the Neutron Data Standards Database

Templates of uncertainties expected in specific measurement types were recently developed. One aim of these templates is to help evaluators in identifying (1) missing or suspiciously low uncertainties and (2) missing correlations between uncertainties of the same and different experiments, when estimating covariances for experimental data employed in their evaluations. These templates also provide realistic estimates of standard deviations and correlations for a particular uncertainty source and measurement type that can be used by evaluators in situations where they are not supplied by the experimenters. This information allows for a more comprehensive uncertainty analysis across all measurements considered in an evaluation and, thus, more realistic evaluated covariances. Here, in this work, we extend a template that is applicable to uncertainties expected in neutron-induced fission, (n,f), cross-section measurements. It is applied to improving covariances of 239 Pu(n,f) cross-section measurements in the database underlying the Neutron Data Standards evaluations. This particular example was chosen since this evaluation is primarily based on experimental information. Also, some uncertainties of individual 239 Pu(n,f) cross-section experiments in this database were suspected to be underestimated. The evaluated uncertainties obtained after updating the covariances in the database by means of the template indeed do increase compared to their original values. Even more importantly, the evaluated mean values change noticeably. These modified cross sections impact application calculations significantly, as is demonstrated by employing them in simulations of the effective neutron multiplication factor for a few selected critical assemblies. However, this updated evaluated 239 Pu(n,f) cross section should not be interpreted as the final one that should replace values of the current Neutron Data Standards project. Evaluations for the Neutron Data Standards of the 239 Pu(n,f) cross section must be linked to many other observables included in the associated database, most notably to cross sections for 235 U(n,f), but also to those for 10 B(n, α ), 6 Li(n,t), 238 U(n,f), and 238 U(n, γ ), because of included measurements of the 239 Pu(n,f) cross section that appear as ratios to these reactions. Some of these other reactions are correlated to further observables in the database. Hence, updating uncertainties of data sets of any of these observables can potentially impact the 239 Pu(n,f) cross section. Uncertainties for all measurements of these linked physical observables have to be updated before a comprehensive evaluation of the 239 Pu(n,f) cross section and its corresponding uncertainties can be provided.

239Pu Neutron-induced fission cross-section↗
DOE OSTI2025

Remeasurement of the 239 Pu(n,f)/ 235 U(n,f) Cross-Section Ratio with the NIFFTE fission Time Projection Chamber Using Vapor-deposited Targets

The NIFFTE fission Time Projection Chamber (fissionTPC) has been used to measure the 239 Pu(n,f)/ 235 U(n,f) cross-section ratio for neutron-induced fission in the range of 0.1–100 MeV, with high precision. A white neutron source was provided by the Los Alamos Neutron Science Center, where the experiment was conducted as a remeasurement to evaluate a roughly 2% discrepancy of the previous fissionTPC results with ENDF/B-VIII.0. Further, a detailed accounting of measurement uncertainties was performed, based on the fissionTPC's novel ability to provide three-dimensional reconstruction of fission-fragment ionization profiles. Current results obtained using a vapor-deposited, highly uniform 239 Pu target, in comparison to the measurement published in 2021, where a 239 Pu electroplated target was used, are presented and discussed. The remeasurement presented here is in agreement with the previous fissionTPC result within measurement uncertainties.

235U↗
DOE OSTI2020

Evaluating 239 Pu(n,f) cross sections via machine learning using experimental data, covariances, and measurement features

In this paper, the neutron-induced 239 Pu fission cross section, 239 Pu(n,f), is evaluated from 1–20 MeV using experimental data and associated covariances while also considering information on the measurement, termed features here. For instance, methods to determine the background, sample backing material, or impurities in the sample, are explicitly taken into account in the evaluation process. To this end, outliers in the experimental data are identified with a modified version of the Hybrid Robust Support Vector Machine. In a second step, two machine learning methods (logistic regression with elastic net regularization and random forest regression with SHAP feature importance metric) are used to highlight measurement features that are common among many of the outlying data points. Based on this analysis, penalty uncertainties are added to the experimental covariances of outlying data points that have outlier measurement features and are put through the generalized-least-squares evaluation. The resulting evaluated mean values and covariances differ distinctly from those data evaluated without the penalty uncertainties. These results highlight that certain measurement features should be more closely examined.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗