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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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511 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Reduction of CO 2 and CS 2 with Uranium(III) Metallocene Aryloxides

Herein, the reactivity of two metallocene aryloxide U(III) complexes, [(C 5 Me 5 ) 2 U(O-Ar)], Ar = 4- t BuC 6 H 4 , 1; Ar = 2,6- t Bu 2 -4-CH 3 C 6 H 2 (BHT), 3, with CO 2 and CS 2 has been investigated. The reaction of 1 with CO 2 produces a bridging oxo complex, [{(C 5 Me 5 ) 2 (4- t BuC 6 H 4 -O)U} 2 (μ 2 -O)], 4, while 3 with CO 2 results in reductive disproportionation to form the bridging carbonate species, [{(C 5 Me 5 ) 2 (BHT)U} 2 (μ 2 -κ 2 :η 1 -CO 3 )], 5. The difference in reactivity can be attributed to the steric properties of the ligand because the reaction of 3 with an oxo-delivering agent yields a U(V) terminal oxo complex, [(C 5 Me 5 ) 2 (BHT)U=O], 6. Reduction of CS 2 to form a bridging (CS 2 ) 2- ligand, [{(C 5 Me 5 ) 2 (tBuC 6 H 4 -O)U}2(μ 2 -CS 2 )], 7, is observed with 1, while the reaction of 3 with CS 2 also produces a bridging (CS 2 ) 2- reduced ligand complex, followed by C-H bond activation of a methyl group from one (C5Me 5 ) 1- ring, [(C5Me 5 ) 2 (BHT)U{μ 2 -C(H)S 2 }U(C 5 Me 4 CH 2 )(C 5 Me 5 )(BHT)], 8. All compounds are characterized by NMR and IR spectroscopy, and their solid-state structures are determined by X-ray crystallography.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Synthesis and Reactivity of Heteroleptic U 4+ Alkyl, Benzyl, and Hydride Imidophosphorane Complexes

A series of heteroleptic U 4+ benzyl, neopentyl, and methyl complexes supported by the imidophosphorane ligand, [N = P(N,N′-ditert-butylethylenediamide)(diethylamide)] 1− (NP*), were synthesized from the monoiodide precursor, [UI(NP*) 3 ]. These heteroleptic complexes were synthesized through the selective formation of [UI(NP*) 3 ] under transmetalation conditions in the reaction between [UI 4 (1,4-dioxane) 2 ] and K[NP*]. Formation of the homoleptic complex [U(NP*) 4 ] was not observed even in the presence of excess K[NP*]. The oxidation and hydrogenolysis reactivity of the neopentyl complex, [U(Npt)- (NP*) 3 ] (Npt = neopentyl) was explored. While cyclic voltammetry indicates a potentially isolable U5+ alkyl cation, chemical oxidation of the neopentyl complex results in the isolation of a cationic U 4+ complex with a bound diethyl ether in the primary coordination sphere, [U4+(NP*)) 3 (Et 2 O)][(BArF 24 )] (BArF 24 = tetrakis(3,5-bis(trifluoromethyl)phenyl)borate). Notably, hydrogenolysis of [U(Npt)(NP*) 3 ] with H2 gas at −20 °C results in the formation of a terminal hydride intermediate confirmed by in situ NMR spectroscopy and deuterium labeling with D 2 . The connectivity and structural parameters of this hydride intermediate, [UH(NP*) 3 ], which rapidly thermally decomposes to the homoleptic complex, [U(NP*) 4 ], can be confirmed by single-crystal X-ray diffraction studies of a crystal grown by chilling the reaction mixture. The identity of [U(NP*) 4 ] was confirmed by its direct, bulk synthesis from [U(Me)(NP*) 3 ] and HNP* in a protonolysis reaction.

Alkyls↗
DOE OSTI2023

A Design and Fabrication Options Trade Study for Metallic Fuel without Internal Sodium Bonding

Nuclear fuels using alloys of uranium, or metallic fuels, have many beneficial properties. The classical metallic fuel design uses a loose fitting cylindrical “slug” of fuel placed inside stainless-steel cladding tubes where the gap is filled with sodium. This sodium bond is liquid at operating temperature and conducts heat from the slug to the cladding, especially in early life before fuel swells into contact with the cladding. Despite the benefits of sodium bonding, there is a desire to develop metallic fuel technologies without it chiefly to reduce chemical reaction hazards in spent fuel storage from sodium fast reactors operating on once-through fuel cycles. Elimination of the sodium bond may also help unlock potential benefits for fuel fabrication, reactor neutronics, and compatibility with other types of reactors. Creating a sodium-free metallic fuel revolves around the problem of manufacturing fuel slug geometries which are in close contact with the cladding at beginning of life to facilitate heat transport while alleviating fuel-cladding chemical interactions (FCCI) at this interface and providing enough free volume to accommodate fuel swelling. Accelerating development and qualification of this fuel system will require careful selection of design and manufacturing options. To this end, a design trade-off study was performed to evaluate candidate options. Several design and manufacturing options were assessed, weighted, scored, and ranked with respect to fabrication, normal reactor operation, off-normal scenarios, and back-end considerations. This effort was performed both for “baseline” needs, which represented a once-through fuel cycle at temperatures and burnups known to be viable for sodium-bonded metallic fuel, and for “enhanced” needs to represent opportunities for closed fuel cycles and/or more aggressive temperatures/burnups. The outcomes of this study prioritized a baseline technology using U-Zr alloy with additives to mitigate FCCI, produced in annular slug geometry by continuous casting, clad in austenitic stainless-steel alloy, and followed by a final step to swage the cladding down to close the gap. This study prioritized an enhanced fuel technology using U-Mo alloy, also produced by continuous casting into an annular geometry, followed by coating/plating with an FCCI barrier on the slug, again with a final step to swage the cladding diameter down using oxide dispersion strengthened steel. It was noted that development of the enhanced fuel technology would entail more risk, thus U-Zr alloy was put forth as a backup to U-Mo if challenges are encountered with FCCI barriers, and advanced ferritic/martensitic steels are put forth as a backup to oxide dispersion strengthened steels if swaging and welding are found unworkable.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

f-Element complexes with benzyl and cyclohexyl substituted trihydroborates

Actinide complexes containing the simplest borohydrides (BH 4 ) 1- and (MeBH 3 ) 1- can exhibit remarkably highly volatility, which creates unique hazards and handling challenges, especially when making measurements on solid samples under vacuum. Here we describe efforts to prepare new actinide borohydride complexes with attenuated volatility by adding bulkier benzyl (Bn) and cyclohexyl (Cy) substituents to boron. Reactions of ThCl 4 , UI 3 (thf) 4 , and NdI 3 with the mixed alkali metal salt Li/K(BnBH 3 )(thf) n yielded Th(BnBH 3 ) 4 (thf) 2 , U(BnBH 3 ) 4 (thf) 2 , and K[Nd(BnBH 3 ) 4 ], respectively. Notable amongst these, the reaction with UI 3 (thf) 4 proceeds via oxidation of U(III) to U(IV) despite the presence of reducing borohydride ligands. Similarly, reactions of the same metal halides with four equivalents of Li(CyBH 3 )(Et 2 O) n yielded Th(CyBH 3 ) 4 , U(CyBH 3 ) 4 (thf) 2 , and [Li(Et 2 O) 3 ][Nd(CyBH 3 ) 4 ]. Single crystal X-ray diffraction studies of the M(BnBH 3 ) 4 (thf) 2 complexes with M = Th and U confirmed their formulations. Furthermore, the complexes have approximate D 2d point group symmetry and adopt bicapped hexagonal antiprismatic coordination geometries with axial thf ligands and κ 3 -BnBH 3 ligands bound in the equatorial plane. K[Nd(BnBH 3 ) 4 ] and [Li(Et 2 O) 3 ][Nd(CyBH 3 ) 4 ], which were prepared for comparison to U(III) complexes that were unsuccessfully targeted, were also structurally characterized to reveal complex anions with tetrahedral arrangements of trihydroborate ligands bound to Nd(III). Crystals obtained for Th(CyBH 3 ) 4 and U(CyBH 3 ) 4 (thf) 2 were not suitable for XRD studies, but 1 H and 11 B NMR spectra were consistent with their formulations. Collectively, these complexes represent rare examples of structurally characterized f-element trihydroborate complexes with carbon substituents other than methyl.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

CoDCon Project (Final Report)

The co-decontamination (CoDCon) project was established in FY 2016 with the objectives of (a) evaluating the uncertainty in the uranium (U)/plutonium (Pu) ratio in a mixed U/Pu product from a tributyl phosphate (TBP)–based solvent extraction flowsheet, and (b) developing and demonstrating on-line optical spectroscopy for real-time monitoring of key components (e.g., Pu, U, and HNO 3 concentrations) in the process solutions. We were interested in assessing the accuracy and precision to which a specific uranium-to-plutonium (U/Pu) ratio can be achieved, which for the purposes of this project was set at a U/Pu mass ratio of 7/3. The uncertainty associated with achieving this specific target U/Pu ratio was investigated during five flowsheet tests using laboratory-scale solvent extraction equipment. In addition, optical spectroscopic techniques were incorporated into the CoDCon solvent extraction testing system, allowing real time monitoring of all input and output process streams. Two CoDCon flowsheet tests were performed in FY 2018 using a simple dissolved fuel simulant containing only U (~1 M) and Pu (~15 mM) in nitric acid (HNO 3 ; ~3 M). In FY 2019, two additional flowsheet tests were performed. For the first of these (CoDCon Run 3), the dissolved fuel simulant was similar to that used in the first two tests, with the inclusion of 1 mM neptunium (Np). The second test conducted in FY 2019 (CoDCon Run 4) used a more representative dissolved fuel simulant, including addition of non-radioactive fission product elements. A fifth CoDCon flowsheet test (CoDCon Run 5) was conducted in FY 2020, with the following additional objectives: (1) routing of the technetium (Tc) in the simulated dissolved fuel solution to the solvent extraction raffinate, and (2) routing of the Np in the simulated dissolved fuel solution to the U/Pu product. All tests used a bank of sixteen 2 cm centrifugal contactors. The tests involved first loading the solvent (30 vol% TBP dissolved in n-dodecane) with U and Pu (and Np, for Run 5), then the Pu (and Np) was stripped from the loaded solvent with a U(IV) solution (~50 mM) and the flowsheet conditions were adjusted such that some U partitioned into the Pu-containing product stream. The amount of U accompanying the Pu was monitored in real time using optical spectroscopic techniques coupled with chemometric modeling. Based on the real-time spectroscopic measurement of the U/Pu ratio, adjustments were made to the flowrate of the fresh TBP solvent phase used to scrub U from the aqueous Pu-containing product. This proved to be a very effective way to control the U/Pu mass ratio in the product. This report presents the results of the CoDCon Run 5 test. The flowsheet tested in Run 5 was substantially different than that run in the prior tests, especially the solvent loading section of the flowsheet. Two key changes were made. First, based on the objective to extract all the Np and route it with the U/Pu product, pentavalent vanadium [V(V)] was added to the feed and scrub solutions. The purpose of the V(V) was to convert all the Np to the +6 oxidation state, which is extractable by TBP. Second, a high acid (8 M HNO 3 ) scrub was added to the flowsheet to scrub the Tc from the solvent. This was followed by a low acid scrub (0.05 M HNO 3 ) to reduce the residual HNO 3 concentration in the solvent prior to the Pu stripping step. The output from the low acid scrub was collected separately, rather than routing towards the raffinate. The modifications to the solvent loading part of the flowsheet were only partially successful. The treatment with V(V) was effective at converting the Np to Np(VI). Only 1.3% of the Np remained in the raffinate solution. However, ~40% of the Np stripped out of the solvent in the low acid scrub step; nearly 20% of the Pu also was stripped from the solvent during the low acid scrub. For further development, either modifications to the flowsheet, or concentration and recycle of the low acid stream into the

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Radiation Effects on Uranyl Tetrachloro Coordination Compounds: Impact of Lattice Water

Nuclear materials, such as uranium-bearing solids, are exposed to high levels of ionizing radiation throughout the nuclear fuel cycle; thus, it is important to develop a molecular-level understanding of how these materials behave and degrade in the presence of gamma (γ) irradiation. In the current study, three U(VI) tetrachloride complexes, M 2 [UO 2 Cl 4 ]·xH 2 O (where M = K + , Rb + , or Cs + and x = 0 or 2), and their respective chloride salts were exposed to 1–50 kGy of γ radiation using a 60 Co source. Irradiated materials were evaluated by using electron paramagnetic resonance (EPR) and Raman spectroscopy and were further explored by using density functional theory (DFT) methods. EPR spectra of the irradiated materials suggest the formation of a Cl-based radical for both the alkali salts and the uranyl tetrachloride compounds, and DFT calculations provide evidence that the Cl 2 –• radical is formed within these materials. The presence of water in the K + and Rb + compounds leads to additional spectroscopic signatures that could be traced back to water radiolysis and the formation of peroxide and superoxide species. DFT results support the formation of HO 2 • in the lattice and potentially the formation of a [UO 2 Cl 3 (O 2 )] 3– species, highlighting the impact of water within the hydrated material to alter U(VI) speciation by radiolysis.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Fracture behavior of irradiation induced nanocrystalline UO 2 studied by in-situ mechanical testing in transmission electron microscopy

Uranium Dioxide (UO 2 ) is widely used as a fuel in current light water reactors (LWRs). Upon accumulation of radiation damage, LWR UO 2 fuel pellets start to develop a different microstructure at the pellet periphery when fuel burnup exceeds 45–50 GWd/tHM. The resulting porous, nanocrystalline microstructure is one of the most prominent microstructural changes occurring in such fuel. Its fracture mechanisms, which causes fuel fine fragmentation, could impact safety limits when the cladding breaches. Direct measurements of these properties are challenging, therefore a surrogate obtained via ion irradiation can be used. In this study, multiple microcantilevers were fabricated by focused ion beam from both fresh UO 2 and UO 2 irradiated with 84 MeV Xe 26+ ions to a peak dose of 1357 displacements per atom (dpa). Further, the irradiation produced a pseudo high burnup structure approximately 2 µm below the surface. In-situ nano-mechanical bending tests were conducted to investigate the fracture behavior and the effect of the surrogate UO 2 high burnup structure on local fracture properties. Fresh UO 2 fuel was observed to fracture in transgranular mode without nucleation or movement of dislocations. However, the Xe-irradiated nanocrystalline microcantilevers fractured along the grain boundaries, with no influence from the pre-existing micro-cracks in the microcantilever. Fracture toughness for this type of surrogate high burnup UO 2 structure is reported for the first time in literature. Both the fracture stress and toughness show degradation for UO 2 as a result of Xe-irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗