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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Thiol redox proteomics: Characterization of thiol-based post-translational modifications

Redox post-translational modifications on cysteine thiols (redox PTMs) have profound effects on protein structure and function, thus enabling regulation of various biological processes. Redox proteomics approaches aim to characterize the landscape of redox PTMs at the systems level. These approaches facilitate studies of condition-specific, dynamic processes implicating redox PTMs and have furthered our understanding of redox signaling and regulation. Mass spectrometry (MS) is a powerful tool for such analyses which has been demonstrated by significant advances in redox proteomics during the last decade. A group of well-established approaches involves the initial blocking of free thiols followed by selective reduction of oxidized PTMs and subsequent enrichment for downstream detection. Alternatively, novel chemoselective probe-based approaches have been developed for various redox PTMs. Direct detection of redox PTMs without any enrichment has also been demonstrated given the sensitivity of contemporary MS instruments. This review discusses the general principles behind different analytical strategies and covers recent advances in redox proteomics. Several applications of redox proteomics are also highlighted to illustrate how large-scale redox proteomics data can lead to novel biological insights.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2021

A multimillion-year-old record of Greenland vegetation and glacial history preserved in sediment beneath 1.4 km of ice at Camp Century

Understanding the history of the Greenland Ice Sheet (GrIS) is critical for determining its sensitivity to warming and contribution to sea level; however, that history is poorly known before the last interglacial. Most knowledge comes from interpretation of marine sediment, an indirect record of past ice-sheet extent and behavior. Subglacial sediment and rock, retrieved at the base of ice cores, provide terrestrial evidence for GrIS behavior during the Pleistocene. Here, we use multiple methods to determine GrIS history from subglacial sediment at the base of the Camp Century ice core collected in 1966. This material contains a stratigraphic record of glaciation and vegetation in northwestern Greenland spanning the Pleistocene. Enriched stable isotopes of pore-ice suggest precipitation at lower elevations implying ice-sheet absence. Plant macrofossils and biomarkers in the sediment indicate that paleo-ecosystems from previous interglacial periods are preserved beneath the GrIS. Cosmogenic 26 Al/ 10 Be and luminescence data bracket the burial of the lower-most sediment between <3.2 ± 0.4 Ma and >0.7 to 1.4 Ma. In the upper-most sediment, cosmogenic 26 Al/ 10 Be data require exposure within the last 1.0 ± 0.1 My. The unique subglacial sedimentary record from Camp Century documents at least two episodes of ice-free, vegetated conditions, each followed by glaciation. The lower sediment derives from an Early Pleistocene GrIS advance. 26 Al/ 10 Be ratios in the upper-most sediment match those in subglacial bedrock from central Greenland, suggesting similar ice-cover histories across the GrIS. We conclude that the GrIS persisted through much of the Pleistocene but melted and reformed at least once since 1.1 Ma.

58 GEOSCIENCES↗
DOE OSTI2024

STAT7 v1.2 Verification and Validation Report

This report documents the software testing which has been performed for the STAT7 Version 1.2 (hereafter STAT7 v1.2) software, a code for the statistical propagation of uncertainties in steady-state thermal hydraulics (TH) analysis of plate-fueled research and test reactors. The key capabilities addressed in this testing were selected based on the use of the software for analysis work supporting the conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors (USHPRR) such as the Massachusetts Institute of Technology Research Reactor (MITR-II). This testing scope includes comparison of the code’s statistical processing capabilities using statistical tests, comparison of coolant property evaluations to NIST data tables, and systematic comparison of key TH solver capabilities to the results of hand calculations and to predictions from PLTEMP/ANL Version 4.4. The good agreement achieved in this testing campaign, within the acceptance criteria of a 1% relative difference, also demonstrates that these capabilities have been implemented in STAT7 v1.2 correctly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

STAT7 v2.0 Verification and Validation Report

This report documents the software testing which has been performed for STAT7 Version 2.0 (hereafter STAT7 v2.0), a code for the statistical propagation of uncertainties in steady-state thermal hydraulics (TH) analysis of plate-fueled research and test reactors. The key capabilities addressed in this testing were selected based on the use of the software for analysis work supporting the conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors (USHPRR) such as the Massachusetts Institute of Technology Research Reactor (MITR-II). This testing scope includes comparison of the code’s statistical processing capabilities using statistical tests, comparison of coolant property evaluations to NIST data tables, and systematic comparison of key TH solver capabilities to the results of hand calculations and to predictions from PLTEMP/ANL Version 4.4. The

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Production of National Nuclear Material Archive Subsamples for High Precision Chemical Analysis

With its historical mission as the focal point for production of uranium components for the NNSA nuclear weapons program, the National Nuclear Materials Archive (NNMA) makes use of Y-12's wide range of uranium materials processing knowledge and onsite materials to identify, collect, and preserve sample materials for nuclear forensics purposes. In 2019, Y-12 took on the responsibility for identifying, subsampling, and shipping a total of 24 NNMA samples to the Lawrence Livermore National Laboratory (LLNL) for further forensics analysis. These were highly enriched uranium metal pieces representative of different weapon systems components produced by Y-12 from approximately 1963 to 1993. The purpose was to provide the NNMA program with higher precision analyses of these materials than what is currently available from historical Y-12 production stream data.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

The 2022 Russian Invasion of Ukraine: Nuclear Supply Chains vs. Sanctions

The United States and European countries are unlikely to collectively approve of sanctions against Russian exports of nuclear fuel assemblies and uranium in the near term despite the 2022 invasion of Ukraine. Rather than bearing the costs that cutting off Russian imports would have on domestic nuclear energy production, the United States and European countries are more likely to prefer developing alternative supply chains of nuclear fuel and enriched uranium to reduce dependence on Russian resources. This outlook is primarily based on two areas of evaluation: (1) comparison of the Russian-made reactor fuel supply chain within the European Union (EU) in 2014 following the annexation of Crimea and in 2022 following the invasion of Ukraine, and (2) the dependence of France and the United States on Russia’s exports of enriched uranium. Additionally, the potential impacts to Kazakh uranium exports if trade routes for nuclear resources through Russia became unavailable were considered as an additional factor in the push for disentanglement from Russian nuclear resources. This work utilized BACI, a database of harmonized international trade data at the product level drawn from the United Nations Comtrade Database and created by the French international economics research institute CEPII.1 Comparisons of imports and exports were based on the trade value in U.S. dollars as reported in BACI instead of the quantity of the traded product. This evaluation only considers import and export data up until 2021, since the trade data for 2022 is incomplete within the Comtrade Database for several of the key countries presented in this work, including Australia, Canada, Namibia, Niger, and Russia.

Political science↗
DOE OSTI2021

Approaches to investigate crop responses to ozone pollution: from O 3 -FACE to satellite-enabled modeling

Ozone (O 3 ) is a damaging air pollutant to crops. As one of the most reactive oxidants known, O 3 rapidly forms other reactive oxygen species (ROS) once it enters leaves through stomata. Those ROS in turn can cause oxidative stress, reduce photosynthesis, accelerate senescence, and decrease crop yield. To improve and adapt our feed, fuel, and food supply to rising O 3 pollution, a number of Free Air Concentration Enrichment (O 3 -FACE) facilities have been developed around the world and have studied key staple crops. In this review, we provide an overview of the FACE facilities and highlight some of the lessons learned from the last two decades of research. We discuss the differences between C 3 and C 4 crop responses to elevated O 3 , the possible trade-off between productivity and protection, genetic variation in O 3 response within and across species, and how we might leverage this observed variation for crop improvement. We also highlight the need to improve understanding of the interaction between rising O 3 pollution and other aspects of climate change, notably drought. Finally, we propose the use of globally modeled O 3 data that are available at increasing spatial and temporal resolutions to expand upon the research conducted at the limited number of global O 3 -FACE facilities.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

Clustering Acoustic Background Noise in the Stratosphere Using Machine Learning

Infrasound, characterized by low-frequency sound inaudible to humans (<20 Hz), emanates from natural and anthropogenic sources. Its efficacy for monitoring phenomena necessitates robust sensing networks. Traditional ground-based infrasound sensors have limitations due to atmospheric dynamics and noise interference. Balloon-bore sensors have emerged as an alternative, offering reduced noise and improved capabilities. This study bridges clustering algorithms with balloon borne infrasound data, a domain yet to be explored. Employing K-Means, DBSCAN, and GMM algorithms on normalized and reshaped data and only normalized data from a New Zealand-based NASA balloon flight, insights into background noise at stratospheric altitudes were revealed. Despite challenges arising from distinguishing signals amid unique background noise, this research provides vital reference material for noise analysis and calibration. Beyond infrasound event capture, the dataset enriches comprehension of background noise characteristics in the southern hemisphere.

47 OTHER INSTRUMENTATION↗
DOE OSTI2021

AGR-5/6/7 Irradiation Test Final As-run Report

This document presents the as-run analysis of the Advanced Gas Reactor (AGR) 5/6/7 irradiation experiment. AGR-5/6/7 is the last of a series of experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of the development and qualification of tri-structural isotropic (TRISO) low enriched fuel for use in high-temperature gas-cooled reactors (HTGR). The test train contained five separate capsules that were independently controlled and monitored. Each capsule contained multiple 24.91-mm-long and 12.25-mm-dimeter compacts filled with low-enriched uranium carbide/oxide tri-structural isotropic fuel particles. The objectives of the AGR 5/6/7 experiment were to: 1. Irradiate reference-design fuel particles to support fuel qualification. 2. Establish operating margins for the fuel, beyond normal operating conditions. 3. Provide irradiated-fuel performance data and irradiated-fuel samples for post irradiation examination and safety testing. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) was to verify the successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures. Its primary objective is to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions, in support of plant design and licensing. AGR 5/6/7 will also provide irradiated-fuel performance data based on the fission gas release from particles during irradiation. To achieve the test objectives, the AGR-5/6/7 experiment was irradiated in the northeast (NE) flux trap of the ATR with a planned duration of 500 effective full-power days (EFPDs). The NE flux trap was selected because its larger diameter provided greater flexibility for test-train design compared to the Large B positions used for the AGR-1 and AGR-2 irradiations, significantly enhancing test capabilities for the combined irradiation campaigns. Due to delays in the ATR schedule, the AGR-5/6/7 irradiation was significantly shorter than the originally planned 13-cycle schedule. Irradiation began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles (162B?168A) over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 EFPDs. Final burnup values, on a per-compact basis, ranged from 5.66% to 15.26% fissions per initial heavy metal atom, while fast fluence values ranged from 1.62 to 5.55 × 1025 n/m2 (E >0.18 MeV). Time-averaged volume-averaged fuel temperatures on a capsule basis at the end of irradiation ranged from 756°C in Capsule 5 to 1313°C in Capsule 3 excluding days with significantly lower temperature during the two short powered axial locator mechanism cycles, 163A and 167A. By the end of irradiation, 48 out of 54 installed thermocouples (TC) had failed (the bottom three capsules lost all TCs). During the first five cycles (162B ? 165A), fission-gas isotope release-rate-to-birthrate (R/B) ratios were stable in the 10-8?10-6 range, and no in-pile particle failures were observed based on the gross gamma counts. During this time, the high exposed kernel fraction and high fuel particle temperatures in Capsule 1 led to the maximum R/B value of around 2 ? 10-6 for Kr-85m. The fission gas release in all capsules started to increase from the second half of Cycle 166A, when a large number of in-pile particle failures occurred in Capsule 1 and a gas line problem in this capsule caused fission gas leakage at various degrees into the other four capsules. This gas line problem also prevented fission gas release measurement for Capsule 1 during the last three cycles due to gas flow isolation. By the end of irradiation, it is estimated that approximately 15 particle failures in Capsule 3, which was considered possible at the start of the irradiation, since the experiment was designed to operate beyond the high-temperature gas-cooled reactor (HTGR) normal operating temperature range. A few hundred in-pile particle failures were estimated for Capsule 1 by the end of Cycle 166A, but the total number of failures is unknown due to lack of fission gas release data in the later cycles. Additionally, four potential in-pile failures were identified for Capsule 2 during the last cycle, 168A. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on absence of the typical spikes in gross gamma counts and low failure estimates using the AGR-3/4 R/B per exposed kernel model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A Wide View of the Galactic Globular Cluster NGC 2808: Red Giant and Horizontal Branch Star Spatial Distributions

Wide-field and deep DECam multiband photometry, combined with HST data for the core of the Galactic globular cluster NGC 2808, allowed us to study the distribution of various stellar subpopulations and stars in different evolutionary phases out to the cluster tidal radius. We used the C ugi = (u - g) - (g - i) index to identify three chemically distinct subpopulations along the red giant branch and compared their spatial distributions. The most light-element-enriched subpopulation (P3) is more centrally concentrated; however, it shows a more extended distribution in the external regions of the cluster compared to the primordial (P1) and intermediate (P2) composition populations. Furthermore, the P3 subpopulation centroid is off-center relative to those of the P1 and P2 groups. We also analyzed the spatial distribution of horizontal branch stars and found that the relative fraction of red horizontal branch stars increases for radial distances larger than ≈1$^{'}_{.}$5, while that of the blue and hotter stars decreases. These new observations, combined with literature spectroscopic measurements, suggest that the red horizontal branch stars are the progeny of all the stellar subpopulations in NGC 2808, i.e., primordial and light-element enhanced, while the blue stars are possibly the result of a combination of the "hot-flasher" and the "helium-enhanced" scenarios. A similar distribution of different red giant branch subpopulations and horizontal branch stars was also found for the most massive Galactic globular cluster, ω Cen, based on combined DECam and HST data, which suggests that the two may share a similar origin.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2020

Gamma ray Imaging Based Methods to Improve the Accuracy of Uranium Holdup Quantification

Deposits in commercial uranium processing facilities of interest to nuclear safeguards are predominantly low enriched uranium (LEU). The shapes and sizes of holdup deposits can vary significantly, thus making it difficult to model the deposits accurately for calibration purposes. This presents a challenge to currently employed methods such as the Generalized Geometry Holdup (GGH) that rely on quantifying 235U mass by simplifying deposit shapes as a point, a line, or an area. In this work, gamma ray imaging using high energy resolution Germanium Gamma Imagers (GeGI™) is employed to determine the distribution of uranium inside the source containment. Two types of imaging methods are employed; coded-aperture imaging and Compton imaging. Gamma-ray emissions seen in different regions of the image are quantified using an inverse gamma-ray transport solver being developed at ORNL. Using measured data, the inverse method will be used to solve for unknown source parameters such as source matrix thickness, density, and attenuation due to container wall and shielding. The intrinsic efficiency of the detector is determined based on a library of built-in response functions and key dimensions of the detector (e.g., thickness and radius of the detector crystal). Measurements are performed in both imaging modalities using uranium sources of well-known masses and enrichments configured inside mocked-up holdup fixtures. Results from the two imaging modalities are intercompared. The paper presents preliminary results of the imaging measurements and the progress made in the computational methods to quantify uranium mass.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Computational determination of a primary diffusion mode in γ U-10Mo under irradiation

Low enriched uranium (< 20 % 235 U)-molybdenum (U-Mo) monolithic fuel is the primary candidate for high performance research and test reactors, and is in the process of being qualified to replace highly enriched uranium (≥ 20 % 23 5U) fuel. As part of the qualification process, it is critical to understand and predict the behavior of fission gas bubbles under irradiation, which affects fuel swelling and fuel failure. Mechanistic fuel models are being developed that can both reproduce the existing experimental data for fuel swelling, and be further applied to irradiation conditions beyond the experimental scope. Diffusion of species under irradiation conditions is an important parameter in the mechanistic fuel models; however, no temperature-relevant experimental diffusion data exists. In the present work, radiation-enhanced diffusion coefficients of U, Mo, and Xe in λU-10wt.%Mo were calculated in the temperature range between 300 K and 1400 K via rate-theory models and molecular dynamics simulations with an embedded-atom method interatomic potential for the U-Mo-Xe system. Accordingly, total diffusion coefficients under relevant irradiation conditions are determined using previously obtained intrinsic thermal diffusion and radiation-driven diffusion coefficients, as well as the newly calculated radiation-enhanced diffusion coefficients presented herein. Radiation-enhanced diffusion of U and Mo was dominant in the intermediate temperature range, whereas radiation-enhanced diffusion of Xe did not significantly contribute to total diffusion of Xe at the relevant fission rate densities. Radiation-enhanced diffusion of Xe became faster than both intrinsic thermal diffusion and radiation-driven diffusion at a fission rate density of 5 x 10 22 fissions/m 3 /s, which is higher than the typical fission rate density range in research reactors. The temperature regime that radiation-enhanced diffusion of each element dominated was dependent on the fission rate density. Finally, the total diffusion coefficients of U, Mo, and Xe, updated in this work, will be utilized as parameters in the mechanistic fuel models to help predict the behavior of fission gas bubbles under irradiation more accurately.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Foliar Application of an Inositol-Based Plant Biostimulant Boosts Zinc Accumulation in Wheat Grains: A μ-X-Ray Fluorescence Case Study

There has been much interest in the incorporation of organic molecules or biostimulants into foliar fertilizers with the rationalization that these compounds will enhance the uptake, or subsequent mobility of the applied nutrient. The objective of this research was to investigate the effects of an inositol-based plant stimulant on the mobility and accumulation of foliar-applied zinc (Zn) in wheat plants ( Triticum aestivum L.). High-resolution elemental imaging with micro-X-ray fluorescence (μ-XRF) was utilized to examine Zn distribution within the vascular bundle of the leaf and whole grains. The inclusion of myo -inositol with Zinc sulfate, significantly increased Zn concentration in shoots in contrast to untreated controls and Zn sulfate applied alone. Foliar Zn treated plants increased Zn in grains by 5–25% with myo -inositol plus Zn treated plants significantly increasing grain Zn concentration compared to both Zn treated and non-treated controls. XRF imaging revealed Zn enrichment in the bran layer and germ, with a very low Zn concentration present in the endosperm. Plants treated with Zn plus myo -inositol showed an enhanced and uniform distribution of Zn throughout the bran layer and germ with an increased concentration in the endosperm. While our data suggest that foliar application of myo -inositol in combination with Zn may be a promising strategy to increase the absorption and mobility of Zn in the plant tissue and subsequently to enhance Zn accumulation in grains, further research is needed to clarify the mechanisms by which myo -inositol affects plant metabolism and nutrient mobility.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2023

Modeling, analysis, and optimization of complex nuclear processes and facilities via computational methods: The HALEU process case study

Improving and adapting industrial systems to timely meet changing programmatic and market demands is an important goal to achieve, including when operating and maintaining complex nuclear processes and facilities. However, changes to these complex systems are costly, particularly when they are already in place and bounded to stringent requirements and constraints such as when handling radioactive material and contaminated equipment. These conditions often exist when treating spent nuclear fuel remotely within shielded nuclear radiation chambers, commonly referred as hot cells, to condition nuclear material and/or fabricate products for utilization in other nuclear enterprises such as in the manufacture of advanced nuclear fuel. The illustrative case considered here is the production of high assay low enriched uranium (HALEU) products supporting the deployment of advanced nuclear reactors. For the HALEU program, resources invested were and are being systematically analyzed so that these investments are maximized in a facility that is nearly 60 years old. A methodology that has effectively enabled optimized and improvements in the Spent Fuel Treatment (SFT) program, and consequently the HALEU program, involves discrete event simulation as addressed in this article. Here, the quantification of multiple productivity metrics, including material processing rates, cycle times, bottlenecks, number of material transfers as well as equipment, workstation, and material handling utilization, has resulted in a myriad of diverse discoveries and data-informed decisions regarding process layout and constituent, labor levels and schedules, selection of new process units, storage needs, and other critical process configurations. This article describes such a computational capability being applied for decision-making, illustrates its application to an actual process and program, provides illustrative results, and argues how computational methods for the modeling, analysis, and optimization of complex processes and facilities does lead to informed decisions derived from data and not only from intuition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Reaction Rate Ratios for Recent Fast Metal Experiments with Large Plutonium Masses

Reaction rate ratios are integral responses that are used within the criticality experiments field because they contain spectral information. While these types of measurements have been utilized for nuclear data validation with historic experiments, few experiments of this type have been utilized for recent experiments, as few exist. This work focuses on measured reaction rate ratios for two nearly bare plutonium critical assemblies with different geometries: one that is cube like (with a Pu mass of 40 kg) and one that is slab like (with a Pu mass of 109 kg). Irradiations were performed with both configurations in which foils were placed near the center of the assembly. Plutonium, highly enriched uranium, depleted uranium, and Au foils were included in the irradiation and counted via high-purity germanium detectors. From these measurements, reaction rate ratios were calculated. Measured and simulated values and uncertainties are presented for the reaction rate ratios. Ratios utilizing the following reactions are given in this work: 197 Au(n, γ), 197 Au(n,2n), 235 U(n,fission), 238 U(n, fission), 238 U(n,2n), 238 U(n,γ), and 239 Pu(n,fission). Uncertainties for the measured reaction rate ratios ranged from 4% to 7%, and the contribution of various parameters to this uncertainty was investigated. The results are compared to historical experiments and should be used for nuclear data validation for future nuclear data library releases. These measurements are part of the EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) project, which utilizes measurement responses in addition to k eff (such as these reaction rate ratios) to help reduce uncertainties in 239 Pu nuclear data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Cross sections and calculated yields of some radionuclides of yttrium, strontium and rubidium formed in proton-induced reactions on enriched strontium-86: possibility of production of 85g Sr, 83 Rb and 82m Rb in no-carrier-added form

Here, cross sections of the 86 Sr(p,3n) 84m Y, 86 Sr(p,αn) 82m Rb, and 86 Sr(p,x) 85g Sr reactions were measured from their respective thresholds up to 16.2 MeV and from 23.0 to 44.1 MeV at FZJ, and from 14.3 to 24.5 MeV at LBNL, using 96.4% enriched 86 SrCO 3 as target material. Thin targets prepared by sedimentation were irradiated with protons in a stacked-form, and the induced radioactivity was measured by high-resolution γ-ray spectrometry. Nuclear model calculations based on the code TALYS reproduced our experimental cross section data well. From the excitation functions, the integral yields of the above three radionuclides were calculated. The yield of 85g Sr via the natSr(n,γ) process was also measured using the TRIGA Mark-II reactor at AERE, Savar. A comparison of the reactor and cyclotron production of carrier-added 85g Sr is given. The production possibilities of the three investigated radionuclides in no-carrier-added forms at a 30 MeV cyclotron via new routes are discussed.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Critical Experiment Design Phase 2 Report for Integral Request 304

This report documents the second phase of the critical experiment design process (CED-2) conducted as part of integral experiment request (IER) 304. The purpose of IER-304 is to develop a temperature-dependent critical experiment capability at the Sandia National Laboratories critical experiment facility using low-enriched uranium oxide fuel. Only a few benchmark quality critical experiments are currently available to validate criticality safety computational methods at temperatures other than room temperature. Recent advancements in computational techniques such as implementation of on-the-fly Doppler broadening techniques and advanced treatment of thermal scattering data necessitate the development of additional experimental capabilities in this area. The work described in this report developed representative configurations using 7uPCX and BUCCX fuels, analyzed the physics exercised by the different arrays, and analyzed the estimated experimental uncertainties. This report also scopes the information necessary for performing modifications to the reactor to accommodate operation at elevated and reduced temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

SCALE 6.3 Validation: Spent Nuclear Fuel

This report is the fifth volume in a series documenting the validation of SCALE 6.3 with ENDF/B-VII.1 libraries for nuclear criticality safety, reactor physics, radiation shielding, and spent nuclear fuel applications. This fifth volume, which focuses on validating SCALE capabilities that impact spent nuclear fuel applications, provides an update of the similar validation reported for SCALE 6.2.4. The experimental data used as basis for validation herein consist of measurement data for nuclide inventories and decay heat, including the following: 1. radiochemical assay (RCA) measurements of nuclides important to burnup credit, decay heat, and radiation shielding in 205 light-water reactor (LWR) spent nuclear fuel samples that cover burnups ranging up to 80 GWd/MTU and initial fuel enrichments up to 4.9% 235U; 2. full-assembly decay heat measurements for 236 LWR assemblies with initial fuel enrichments ranging up to 4% 235 U, assembly burnups of 5–51 GWd/MTU, and decay times after fuel discharge in the 2- to 27-year range (of importance to spent nuclear fuel storage, transportation, and disposal); and 3. pulse fission irradiations for fissionable materials at cooling times of interest to severe accident analyses (<10 5 s).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗