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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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At least 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

History of Pu-238 Production Restart Efforts at Idaho National Laboratory and Oak Ridge National Laboratory

In the early 2010s, efforts to restart production in the U.S. of plutonium-238 heat source (HSPu) material for NASA deep space missions were initiated. Processes, procedures, hardware, and chemical separations were developed and implemented to enable the production of heat source material at Oak Ridge National Laboratory (ORNL) and Idaho National Laboratory (INL). Here, this review provides an overview of the timeline and efforts associated with the restart of production, as well as upcoming efforts to increase production.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Influence of Borate on Sorption of Strontium and Barium to Calcium Silicate Hydrate - 20148

Radioactive waste with the decommissioning process at the Fukushima Daiichi Nuclear Power Plant (FDNPP), contains borate with relatively high concentration. As with general low-level wastes, such waste is also assumed the solidification with cementitious materials which calcium silicate hydrate (C-S-H) is one of the main components. In this study, the sorption behavior of strontium (Sr) ions and barium (Ba) ions to C-S-H in the presence of borate was examined by the sorption experiments without drying process. In the sorption experiments, C-S-H was synthesized by the mixture of calcium oxide, fumed silica and a solution with setting the Ca/Si molar ratios to 0.4, 0.8, 1.2 and 1.6. The total amount of a solid phase was 1.5 g, and the total volume of the solution containing borate and adsorbed ions (Sr or Ba) was 30 mL. The concentration of borate was set to 0 M or 0.6 M and the sodium/boron (Na/B) ratios was set to 0.2 or 1.0 by mixing boric acid and sodium hydroxide. The concentration of Sr ions or Ba ions was adjusted to 1 mM. The sorption period was set to 7 days. In the results, the sorption ratios of Sr ions and Ba ions for C-S-H exceeded 80% under all experimental conditions, although the sorption competition of cations with Na ions and the inhibition of the C-S-H formation with generating borate compounds observed by Raman spectroscopy might slightly reduce the sorption of them. These suggest that cationic radionuclides will be immobilized by C-S-H contained in the cements even in the presence of borate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

An introduction to the significance of sample size in particle analyses for nuclear forensics and radiological investigations

Particulate isotopic analysis in nuclear forensics has developed rapidly during the past two decades due to technical advances in determining the isotopic composition of individual particles. This paper introduces basic statistical concepts that can be applied by analysts to understand the importance of statistical adequacy when interpretating particle data. While these basic statistical methods provide a useful point-of-entry to particle data analysis, more sophisticated statistical and modeling approaches are needed to extract maximal information from such datasets in the future.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

New measurements of gamma-ray energies and their absolute intensities from the decay of 231 Pa

Accurate nuclear data for the decay of 231 Pa is highly important for non-destructive analysis in the areas of radiochronometry and nuclear forensics. However, the evaluated nuclear data for this isotope is lacking; many previous measurements have large uncertainties and only two measurements are used to determine the absolute intensities. In this work, we present modern measurements of the major gamma-ray emissions from the decay of 231 Pa to 227 Ac including the energies and absolute intensities. For many of these measurements, the error envelopes have been significantly decreased compared to previous measurements, while remaining in good agreement. The absolute intensity of the 283 keV reference line was measured as 0.0163(2).

231Pa↗
DOE OSTI2021

Exploring the use of thorium isotope compositions and concentrations as nuclear forensic signatures for uranium ore concentrates

This interlaboratory study measured thorium concentrations and isotope compositions in uranium ore concentrates from different geographical locations to examine whether thorium impurities may be useful forensic signatures for uranium ore concentrates found out of regulatory control. Measured 230 Th/ 232 Th in fifteen uranium ore concentrates record over three orders of magnitude of compositional variation. Results demonstrate that 230 Th/ 232 Th used in combination with U/Th ratios resulted in a unique signature for individual uranium ore concentrates from different processing locations. Data presented here suggest potential for 230 Th/ 232 Th and U/Th to be used as comparative signatures to investigate the provenance of seized uranium ore concentrates.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

ASMS 2024 Investigation of Uranyl Perchlorate Anion Complexes in the Gas Phase via Infrared Multiphoton Dissociation and Collision Induced Dissociation

Investigation of Uranyl Perchlorate Anion Complexes in the Gas Phase via Infrared Multiphoton Dissociation and Collision Induced Dissociation Brittany D. M. Hodges, Christopher A. Zarzana, JungSoo Kim, Jonathan Martens, and W. C. M. Berden Introduction (120 words max) Effects of electronic structure on chemical bonding and reactivity play critical roles shaping the chemical bonding and reactivity behaviors of heavy elements. Understanding the role of f electrons in bond formation between the actinide-series elements like uranium with other ligands is critical for solving technical challenges associated with these heavy elements, important to nuclear fuel cycles, efficient separations of rare earth metals, and understanding the chemistry of stored nuclear fuels and related environmental management sites. In this study, we further examine the interactions between uranyl and the perchlorate ion in order to understand the structures of these ions through the use of IRMPD. Here we report the IRMPD spectra of [UO2(ClO4)3]-, [UO3(ClO4)2]-, and a proposed transition state. Methods (120 word max) IRMPD spectra and CID spectra were acquired using a Bruker amaZon QIT/MS instrument at the Free-Electron Lasers for Infrared eXperiments (FELIX) laboratory at Radboud University. The FELIX QIT/MS is modified to allow for the high-intensity tunable IR beam from FELIX to be directed into the ion packet, resulting in multiphoton dissociation that is measured only when the IR frequency is in resonance with an adequately high absorption vibrational mode of the mass-selected complex. DFT geometry optimizations and frequency calculations using the Gaussian suite of programs were performed using B3LYP, TPSSh, and PBE0 level of theory with 6-31+G(d) basis for the O, C, H, and N atoms and the SDD basis set for U. The SDD basis set employs the Stuttgart/Dresden effective core potential. Preliminary Data or Plenary Speakers Abstract (300 words max) Metal ion clusters of uranyl perchlorate were formed via direct electrospray ionization. For each metal ligand complex of interest, the parent ion was isolated and collision induced dissociation fragmentation and Infrared Multiphoton Dissociation (IRMPD) fragmentation spectra were acquired. Results presented here are the first look at the IRMPD spectra of [UO3(ClO4)2]-, [UO2(ClO4)3]-. Structures were examined using Gaussian at different levels of theory B3LYP level of theory, TPPSh and PBE0 levels, to reflect the behaviors of uranium metal ligand complexes most accurately. In these structures, we identified an overlap between each of these uranyl stretches resulting in their largely being obscured by a perchlorate mode. The CID product spectra agree with similar structures reported by Groenewold for uranyl nitrate in 2006 (10.1021/ja058106n).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Radiochemical transport analysis of gamma spectroscopic data to support estimation of molten salt reactor off-gas inventories

This work introduces a novel application of radiochronometry to estimate nuclide inventories in molten salt reactor off-gas systems based on gamma spectroscopic data from the Molten Salt Reactor Experiment. By analyzing isotopic, isobaric, and isomeric activity ratios, key depletion model parameters related to species transport within the reactor system could be inferred. The findings demonstrate the potential of leveraging a limited subset of gamma spectroscopy measurements to accurately estimate nuclide inventories throughout the off-gas system. The approach can be useful in reactor design activities and support analyses relevant to operations, safety, security, and safeguards.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Energy dependence of fission product yields in the second-chance fission region

An extensive dataset of cumulative fission product yields has been generated under a joint collaboration between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and the Triangle Universities Nuclear Laboratory (TUNL). The energy dependence of the cumulative yield for a select number of high-yield fission products has been measured using quasimonoenergetic neutrons with energies between 5.5 and 11.0 MeV. This is in addition to previously published data covering 0.5–4.5 and 14.8 MeV. The absolute number of fissions was determined during the irradiation period using dual-fission ionization chambers, and the fission products were measured postactivation by whole target γ-ray spectroscopy. Here this paper presents the absolute cumulative fission product yields as a function of incident neutron energy from the neutron-induced fission of 235 U, 238 U, and 239 Pu isotopes at four incident energies in the second-chance fission region and compares them with existing literature values. Corrections relevant to this collaboration's previously published data are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Selection of Vadose Zone Flow and Transport Properties with Gravel Fraction Corrections for the Hanford Site Composite Analysis and Cumulative Impact Evaluation

This environmental calculation file (ECF) is a compilation of hydrologic and radiological/chemical properties to be used for the updated Hanford Site Composite Analysis (CA) and Cumulative Impact Evaluation (CIE) vadose zone (VZ) modeling. The CA will provide an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at the Hanford Site following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1 Chg 1, Radioactive Waste Management, related to managing low-level waste disposal facilities at the Hanford Site. The CIE evaluates the effects of cleanup decisions regarding groundwater quality in the Hanford Site Central Plateau (DOE/RL-2018-69, Cumulative Impact Evaluation Technical Approach Document). Due to the complexity and large number of waste sites in source operable units (OUs), the computational tools used for the CIE must be capable of representing a range of site conditions and source terms in the VZ while also efficiently computing the impact that cleanup decisions have on the underlying aquifer. Waste-site proximity between and within source OUs has resulted in contaminants commingling in the vadose and saturated zones in complex ways. Plume commingling requires cleanup decisions to be evaluated considering the surrounding waste sites and existing groundwater contamination, therefore demonstrating the need to evaluate cumulative impacts from the VZ to groundwater.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Solidification of REDC Organics and Airborne Release Fraction Measurements

The Radiochemical Engineering Development Center (REDC) generates various liquid organic wastes from processing irradiated targets to recover heavy elements. In the past, these organics were discharged to the Oak Ridge National Laboratory (ORNL) liquid low-level waste (LLLW) system, along with the aqueous waste. Because of a reduction in aqueous LLLW from other ORNL generators and an increase in the radionuclide concentration, particularly 238 Pu, in the organic waste stream, the organics can no longer be discharged to the LLLW system. The plan is to solidify the liquid organic waste for disposal as solid debris waste. The documented safety analysis for REDC uses an airborne release fraction of 5 x 10 -4 with a respirable fraction of 0.2 for plutonium and other actinides from packaged solid debris waste during a fire scenario, for an overall airborne release fraction/respirable fraction of 10 -4 . Solidifying the organic liquids, either by using PM-199 Organoclay ® or by grouting, would produce solid waste forms with no free liquid, which should qualify for disposal at the Waste Isolation Pilot Plant. The use of Organoclay would be much simpler to implement inside a hot cell. Granular Organoclay can be added to the organic waste solutions until there is a dry layer on top of the liquid and then allowed to cure for a few hours. Grouting the liquids would require using multiple ingredients and mixing steps. Both methods produced waste forms that trapped cerium, as a surrogate for plutonium, inside the waste, with airborne release fractions during fire scenario tests of well below the 10 -4 limit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Characterization and CST Batch Contact Equilibrium Testing of Modified Tank 9H Process Supernate Samples in Support of TCCR

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate (CST) media to process radioactive waste solutions for the removal of ¹³⁷Cs. TCCR currently focuses on dissolving Savannah River Site (SRS) radioactive tank waste (primarily sodium saltcake solids) within Tank 10H followed by at-tank ion exchange column treatment. Two supernate batches from Tank 10H have been processed through the TCCR unit and processing of a third batch is expected soon. After processing of this supernate batch, plans are to replace the CST columns and process dissolved salt solution from Tank 9H through Tank 10H and then through the new CST columns installed in the TCCR unit. The new columns are expected to contain either a media similar to an archived CST batch (IE-911) or the R9120-B CST media used in the current TCCR columns (the two materials are fundamentally the same; just different specifications, product names, and preconditioning steps). Samples of Tank 9H dissolved saltcake were received at the Savannah River National Laboratory (SRNL) and characterized. The Tank 9H supernate contained a high sodium concentration (~9.6 M Na⁺) and will require dilution to near 6 M [Na⁺] prior to processing through the TCCR unit. The cesium concentration in Tank 9H is currently significantly higher than was observed with Tank 10H. Three dilutions of the Tank 9H supernate were conducted to mimic possible dilutions that could be conducted in the tank farm prior to TCCR processing using inhibited water, sodium hydroxide, and sodium nitrate solutions. Dilution #1 was prepared by diluting the Tank 9H supernate by a factor of 1.6 with inhibited water, Dilution #2 was prepared by diluting the Tank 9H supernate by a factor of 3.7 with a mixed sodium hydroxide/sodium nitrate diluent, and Dilution #3 was prepared by also diluting by a factor of 3.7 but with sodium hydroxide only. All three dilutions have similar Na⁺ concentrations (~6 M), but Dilutions #2 and #3 contain significantly less cesium. Major and key minor components of the diluted Tank 9H supernate samples are provided in Table ES-1. Batch contact equilibrium tests were conducted with the Tank 9H dilutions and the two different CST media batches being considered for use in the new TCCR columns. Results are summarized in Table ES-2 and compared to ZAM model predictions. The highest cesium distribution coefficient and percent removal were observed with Dilution #3. IE-911 CST (an archived CST media batch) was more effective at removing cesium than the more recently prepared R9120-B CST media, though both media samples removed >88% of the cesium and the differences may not be statistically different considering the overall uncertainty. Based on the results, maximum cesium loadings were calculated for each CST media type and Tank 9H dilution. In general, maximum cesium loadings from dilutions of this supernate batch are quite high (approaching 0.1 mmol total Cs⁺/g CST for Dilution #1). The highest calculated maximum ¹³⁷Cs loading for the Tank 9H dilutions using the ZAM model with input of the tank compositions and batch contact results was 207 Ci/kg CST (Dilution #1 with IE-911 CST). In all cases, higher maximum cesium loading values were predicted for IE-911 CST versus R9120-B, although the differences varied considerably between the dilutions. The maximum loading value for IE-911 CST with Tank 9H Dilution #1 was only 7% higher than the maximum loading for R9120-B. The maximum loading value for IE-911 CST with Dilution #3 was 24% higher than the maximum loading for R9120-B. Dilution #2 was intermediate between these values. Note that these maximum loading values are the theoretical calculated values, and actual operating conditions will cause differences. A CST binder dilution (correction) factor near 0.7 (relative to pure powder CST) was required for each batch contact test with IE-911 engineered CST using the three Tank 9H dilutions. Correction factors calculated for R9120-B CST ranged from 0.56 to 0.66 for the three dilutions. Following the batch contact tests, the CST samples were isolated from the Tank 9H solution, washed, dried, and digested in acid following established procedures. The analysis results are provided in Table ES-3. Total cesium loading values determined by CST digestion were similar to the calculated loading values based on solution analysis (Table ES-2) for all samples. In addition, the CST was observed to load calcium (R9120-B sample only), iron, strontium, lead, uranium, and plutonium after contact with the waste supernate, as has been observed previously.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Permanganate Oxidation of Actual Defense Waste Processing Facility (DWPF) Slurry Mix Evaporator Condensate Tank (SMECT) and Offgas Condensate Tank (OGCT) Samples to Remediate Glycolate

Savannah River Remediation (SRR) tasked Savannah River National Laboratory (SRNL) to examine glycolate destruction in radioactive samples emulating the proposed process. Experiments with radioactive Defense Waste Processing Facility (DWPF) condensates were performed to demonstrate the oxidation of glycolate using sodium permanganate. One test each was performed for Off-Gas Condensate Tank (OGCT) and Slurry Mix Evaporator Condensate Tank (SMECT) samples received. Condensates were blended with Recycle Collection Tank (RCT) heel simulant and Slurry Mix Evaporator (SME) product containing glycolate to represent typical RCT compositions with estimated nominal entrainment of glycolate. A permanganate to glycolate (P/G) molar ratio targeting 5.7 decreased sample glycolate concentration below Ion Chromatography (IC), glycolate method reporting limits in less than 2 hours for the OGCT sample and in less than 8 hours for the SMECT sample.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

NA-22 Write for Consortium Efforts

We are asked to look into the improvements for the measurement of Pu of a variety of material types (MT): MT41 (Pu-242, 20 – 60% enriched), MT42 (Pu-242 > 60% enrichment), and MT-82 (Np-237). Many of these are legacy items that were packaged years ago (Fig 1&2), and when the items are re-packaged for long-term storage, they must be re-measured using either NDA (nondestructive analysis methods or radiochemical methods. It is of interest to improve the NDA and radiochemistry to reduce the un-certainty in the measurements, as in the case of larger items (250 g to 2 kg items) the uncertainty in measurements makes it difficult to certify the mass of the material down to the 0.5 g accountability requirement per DOE orders by only doing NDA measurement. It would be of interest to look into methods to improve the detection of these materials, which would not only benefit the weapon manufacturing community, but also has the ability to significantly improve the non-proliferation community.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

M4SF-20LL010302082- Thermodynamic Database for Generic Disposal System Assessment

This progress report (Level 4 Milestone Number M4SF-20LL010302082) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline International Collaborations Activity Number SF-20LL010302081. The activity is focused on our long-term commitment of engaging our partners in international nuclear waste repository research. This includes participation in the Nuclear Energy Agency Thermochemical Database (NEA TDB) Project (Cindy Atkins-Duffin) and development of methodologies for integrating US and international thermodynamic databases for use in SFWST Generic Disposal System Assessment (GDSA) efforts.

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