Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “nitric acid”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 91 records · Page 5

TRISO Particle Isolation and Graphite Removal Using SRNL Vapor Digestion Technology

Tri-structural isotropic fueled reactors are planned to come online in the next decade, but there is not currently a widely agreed upon disposition pathway for this new fuel stream. For long term used nuclear fuel storage, there would be a significant benefit if the waste volume could be reduced or mitigated. Most of the volume from used TRISO fuel contains graphite. SRNL currently has an active patent which describes one pathway to digest the graphite. One benefit in the utilization of this pathway is that multiple cylindrical pieces can be stacked end-to-end in a reaction tube, then nitric acid and water vapor can be flowed through the tube, and the weight change of the individual cores can provide both a reaction profile and overall oxidant use efficiency. The reaction of nitric acid and graphite passes through a series of intermediate products – NO 2 , NO, and N 2 O – from reaction and decomposition to eventually form N2. The data shows that NO2 grows in with increasing temperature between 500-600°C and then decreases by way of either reaction with graphite or decomposition. Nitric oxide, a reaction and thermal decomposition product of NO 2 , exhibits a consistent decline as a function of temperature, which is consistent with the literature. Similarly, the concentrations of N 2 O and N 2 increase as a function of temperature as their reactions with graphite become more favorable.

Advanced Reactors↗

Photocatalytic Nitrate Destruction Studies in Complex Environments

One method for denitration of nitric acid used in nuclear facilities is to use formic acid as a reductant. The major problem with formic acid denitration is an induction period of varying duration that may result in excessive accumulation of formic acid at the reaction onset. This accumulation poses an off-gas process control issue. In this talk, we will describe the use of titaniabased photocatalysts for the treatment of nitric acid and nitrate wastes. We find that the photocatalytic process is a simple and straightforward method to completely destroy nitrate ions at room temperature without any initiation period.

nanomaterials↗

Investigating Radiation-Induced Actinide Species in Solution

Our fundamental understanding of actinide radiation-induced redox chemistry is crucial due to their unavoidable exposure to ionizing radiation fields, both inherent and from in-process applications. Plutonium (Pu) and americium (Am) both possess multiple oxidation states, the careful manipulation of which are essential in the study and utilization of their rich chemistry, developing new nuclear technologies, and securing the long-term sustainability of nuclear power. However, knowledge in this area is far from complete. Consequently, we have studied the radiation-induced chemistry of both Pu and Am through a variety of techniques. Temperature-controlled electron pulse radiolysis has been used to study Am for the first time, determining the feasibility of Am redox reactions under used nuclear fuel reprocessing conditions, (e.g. nitric acid, non-ambient temperature). Additionally, we developed an experimentally evaluated multi-scale computer model for the prediction of gamma radiation-induced Pu redox chemistry due to radiolysis and disproportionation reactions in concentrated nitric acid solutions.

actinide↗

Investigating Radiation-Induced Actinide Species in Solution

Our fundamental understanding of actinide radiation-induced redox chemistry is crucial to nuclear fuel cycle development, due to the unavoidable exposure of these elements to ionizing radiation fields, both inherent and from in-process applications. Plutonium (Pu) and americium (Am) both possess multiple oxidation states, the careful manipulation of which are essential in the study and utilization of their rich chemistry, developing new nuclear technologies, and securing the long-term sustainability of nuclear power. However, knowledge in this area is far from complete. Consequently, we have studied the radiation-induced chemistry of both Pu and Am through a variety of techniques. Temperature-controlled electron pulse radiolysis has been used to study Am for the first time, determining the feasibility of Am redox reactions under used nuclear fuel reprocessing conditions, (e.g. nitric acid, non-ambient temperature). Additionally, we have developed an experimentally evaluated multi-scale computer model for the prediction of gamma radiation-induced Pu redox chemistry due to radiolysis and disproportionation reactions in concentrated nitric acid solutions.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Sensitivity study of coupled chemical-CFD simulations for analyzing aluminum-clad spent nuclear fuel storage in sealed canisters

We report the United States Department of Energy (DOE) manages over 50 Metric Tons Heavy Metal (MTHM) of aluminum-clad spent nuclear fuel. One main source for DOE’s Aluminum-clad spent nuclear fuel (ASNF) inventory is the advanced test reactor (ATR) at the INL site, which makes this fuel of particular interest for storage scenarios. Road-ready and final disposition packaging configurations for the ATR fuel dictates storage within helium-backfilled, sealed DOE standard canisters. The conditions within these sealed canisters for extended (greater than50 year dry) storage is of interest. To further this goal, a three-dimensional (3D) multi-physics computational fluid dynamics (CFD) model is developed of the sealed DOE standard canisters. This 3D CFD model is one-way coupled with bulk gas radiolysis reactions considering sealed canisters with inert gas and possible trace amounts of air and water vapor. This study looks at the evolution of the thermal history of the canisters over a 50 year time period with a coupling to the chemical reactions occurring from radiolytic breakdown of residual water. A sensitivity study is then carried out over the parameters of the model including the fuel decay heat, residual water content, sealed pressure, canister external temperature, and canister emissivity. In pure helium, hydrogen generation rates are low, under 10 ppm, but hydrogen generation rates are affected greatly by the presence of even 1% residual air, increasing by 50-plus-fold, and nitric acid generation with residual air also occurs ranging from 500 to 4000 ppm after 50 years. The fuel decay heat and the residual water content show the most importance in the generation of hydrogen gas in pure air, and for nitric acid with a residual air condition. External temperature, canister emissivity and sealed pressure all show minor sensitivity effects to the generation of potentially harmful species.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Electrochemical Behavior of Cerium at an Indium Tin-Doped Oxide Electrode in Acidic Media

The redox behavior and speciation of cerium at mesoporous thin films composed of nanoparticles of indium tin-doped oxide (nITO) electrodes were characterized in pH 4.8, 0.1 M acetate buffer and both 0.1 and 1 M HNO 3 using electrochemical techniques and X-ray photoelectron spectroscopy. Anodic deposition of ceria species from Ce(III) to the nITO electrode was achieved under all solvent conditions via spontaneous condensation of electrochemically generated ceric hydroxide species. In 1 M nitric acid, the rate of CeO 2 dissolution is on the same order as CeO 2 deposition, resulting in negligible amounts of CeO 2 electrodeposited at the nITO surface. The cathodic stripping of CeO 2 from the nITO substrate deposited in 0.1 M nitric acid or pH 4.8 acetate buffer follows a 2-step process where Ce(IV)-oxide is initially reduced to an unstable Ce(III)-oxide species that rapidly undergoes acid catalyzed dissolution to yield soluble Ce(III) (aq) . These findings provide a foundation for the pH and anodic potential controlled deposition of CeO 2 thin films to ITO substrates, which can aid in the development of materials composed of ceria. As a result, they can also be used to infer likely analogous actinide redox behavior and speciation at these electrodes.

Cerium↗

Irradiation Impact on Uranium Recovery Under Direct Extraction Conditions

Reducing the quantity of high-level radioactive waste is essential for minimizing environmental impact and improving efficiency of using natural resources for nuclear power. The current standard, Plutonium Uranium Solvent EXtraction (PUREX), uses tributyl phosphate (TBP) ligands to extract complexes of uranium and plutonium from a nitric acid (HNO3) phase. Although this method is effective, large volumes of HNO3 and the non-incinerable phosphate ligands increase the amount of hazardous waste produced. Alternative extractants and flowsheets have been proposed that allow for more selective extraction of radioactive metals, reduced nitric acid use, and easier incineration by only containing carbon, hydrogen, oxygen, and nitrogen (CHON). One candidate, N,N-di(2-ethylhexyl)-isobutyramide (DEHiBA) exhibits promising properties for direct extraction. A HNO3 pre-equilibrated DEHiBA phase selectively extracts U(VI), leaving plutonium, transuranics, and fission products behind as precipitate and reducing the volume of radioactive HNO3 produced.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗

Actinide oxide dissolution in tributyl phosphate

An alternative to dissolving used nuclear fuel (UNF) in an acidic solution during reprocessing is direct dissolution in an organic solution, which would eliminate an aqueous dissolution step, decrease the amount of nitrate needed, and reduce the facility size. The flowsheet for this potentially less expensive alternative is first to voloxidize the UNF to remove fission product gases and form an oxide. After voloxidation, the UNF is then dissolved in an organic solution containing an extractant mixed with an aliphatic diluent and pre-equilibrated with nitric acid. The organic solution then goes through a solvent extraction process to recover the uranium and/or other desired radionuclides. This work qualitatively studied the dissolution of actinide oxides (UO2, NpO2, and PuO2) in tributyl phosphate using UV-Vis-NIR absorbance spectroscopy to ascertain dissolution behavior. Initial studies included material that is otherwise difficult to dissolve in only nitric acid, specifically CeO2, that is sometimes used as a dissolution surrogate for PuO2. This work confirmed that CeO2, NpO2, and PuO2 are difficult to dissolve in 30 vol% TBP-dodecane pre-equilibrated with 10 M HNO3 and will readily dissolve when co-precipitated with U (i.e., the mixed oxides U-Ce, U-Np, and U-Pu), surrogates for voloxidized nuclear fuel.

Gogolski, Jarrod [Savannah River National Laborato↗

Zirconium Sludge Criticality Calculations in Large Process Tanks

The Savannah River Site’s (SRS) H-Canyon recently re-activated its 6.3D electrolytic dissolver to dissolve stainless steel clad research reactor fuels from Japan. A potential new mission was identified to dissolve other long-cooled fuel stored at SRS that is not aluminum clad. In the basic process of dissolution, the spent fuel dissolves into nitric acid (bulk solution). The eight ft. high and four ft. radius 6.3D dissolver is unique in that it has a platinum-coated niobium insert to resist corrosion, and the niobium basket forms an anode and cathode arrangement within the charge chute (Fig 1,2). DC current flows from anode to cathode, at up to 10,000 amps at 45 volts, through the potential gradient in the nitric acid solution, which allows the stainless steel (or other non-aluminum cladding) of the fuel to dissolve.

Wade, Brindley↗

Process Intensification for Recovery of Uranium from Spent Fuel Using DEHiBA

Two approaches are being pursued to intensify the DEHiBA process for recovery of U from used nuclear fuel. For the traditional solvent extraction approach in which the fuel is first dissolved in hot nitric acid, the DEHiBA concentration was adjusted to 1.5 M. This allows for increased loading in the organic phase, but the organic phase U concentration should remain below 100 g/L to avoid unfavorable physicochemical properties that would upset the hydrodynamics in contacting equipment such as centrifugal contactors. Direct extraction of U into the 1.5 M DHEiBA solvent is another intriguing approach to intensifying the process. In this case, the hot nitric acid dissolution step is avoided, a potential significant simplification of the process. Strategies for routing Tc, Np, and Pu to the HLW stream will likely need to be developed to avoid contamination of the U product with these undesirable species.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Adapting LEU-Modified Cintichem Process for Processing Up to 750 Grams of Uranium Targets

Argonne National Laboratory (Argonne) developed a low-enriched uranium (LEU)-modified Cintichem (LMC) process to enable production and purification of fission-produced molybdenum-99 (Mo-99) from LEU foil targets. This separation scheme starts with nitric acid (HNO 3 ) dissolution of irradiated LEU foil targets that yields an approximate final volume of 2 L. The dissolved uranium solution containing fission products, including Mo-99, is then treated by loading the solution on a titania-based recovery column to separate Mo-99 from bulk uranium. After several washing steps, Mo and other fission products are stripped from the titania column using sodium hydroxide (NaOH). The eluate is then acidified using nitric acid and subsequently processed for final purification using the LMC process. Uranium and most fission products are not adsorbed on the titania column but are eluted (i.e., removed using a solvent) during the wash steps; they can be purified and recycled into new uranium targets.

07 ISOTOPE AND RADIATION SOURCES↗

Modeling the behavior of concentrated aqueous HNO 3 using machine learning interatomic potentials

We develop two multi-defect machine learning interatomic potentials (MLIPs) trained at the BLYP-D2 and PBE-D3 density functional theories using the DeepMD-kit, allowing for the investigation of structural and thermodynamic properties of nitric acid over a wide range of concentrations via molecular dynamics (MD) simulations. We directly compute the degree of dissociation, α, and pK a from MD simulations, revealing that HNO 3 behaves as a weaker acid at higher concentrations, noting that our standard-state pK a value is in excellent agreement with the experimental one. In general, good agreement is observed with experimental results such as α and density outside the training dataset, with only modest deviations at low-to-medium concentrations. We benchmark our custom multi-defect DeepMD MLIPs against foundational models MACE-MP0 and MACE-OFF23. The foundation models capture some aspects of HNO 3 /NO 3 − solvation in concentrated nitric acid but show noticeable density errors and miss subtle structural features relevant to spectroscopy, whereas the bespoke DeepMD MLIPs yield more compact solvation shells, reproduce density-concentration trends, and run ∼12–15× faster than MACE-MP0. Although classical FFs are still more efficient and match experimental densities better, they lack chemical reactivity and thus cannot predict α or pK a , underscoring the need for system-specific reactive MLIPs beyond universal MLIPs.

Dinpajooh, Mohammadhasan [Pacific Northwest Nation↗

Neptunium extraction by N,N-dialkylamides

Separation of neptunium by solvent extraction has been based on tributylphosphate (TBP) for decades, but TBP is not fully incinerable, which adds to the burden of long-lived radioactive waste. Alternatives to TBP for uranium and plutonium extraction, such as the N , N -diakylamides, previously have been explored in the hopes of transitioning to an extractant that is incinerable. Four N , N -diakylamides, N , N -dihexylhexanamide (DHHA), N , N -dihexyloctanamide (DHOA), N , N -di(2-ethylhexyl)butanamide (DEHBA), and N , N -di(2-ethylhexyl)-iso-butanamide (DEHiBA) were considered in this work for their potential to extract millimolar concentrations of Np(IV), Np(V), and Np(VI) from nitric acid solutions into organic solutions containing 1 M extractant in Exxsol D60. Furthermore, under these conditions the branching of the alkyl substituents affects the extractability of Np(VI) and Np(IV), causing three of the dialkylamides, DHHA, DHOA and DEHBA, to extract neptunium in the expected order Np(VI) > Np(IV) > > Np(V). In contrast, branched DEHiBA is so poor an extractant for Np(IV) that the extraction order becomes Np(VI) > > Np(V) > Np(IV) between 0.1 and 5.6 M HNO 3 due to partial oxidation of the Np(V) in nitric acid.

07 ISOTOPE AND RADIATION SOURCES↗

Model bias in simulating major chemical components of PM 2.5 in China

High concentrations of PM 2.5 (particulate matter with an aerodynamic diameter less than 2.5 µm) in China have caused severe visibility degradation. Accurate simulations of PM 2.5 and its chemical components are essential for evaluating the effectiveness of pollution control strategies and the health and climate impacts of air pollution. In this study, we compared the GEOS-Chem model simulations with comprehensive datasets for organic aerosol (OA), sulfate, nitrate, and ammonium in China. Model results are evaluated spatially and temporally against observations. The new OA scheme with a simplified secondary organic aerosol (SOA) parameterization significantly improves the OA simulations in polluted urban areas, highlighting the important contributions of anthropogenic SOA from semivolatile and intermediate-volatility organic compounds. The model underestimates sulfate and overestimates nitrate for most of the sites throughout the year. More significant underestimation of sulfate occurs in winter, while the overestimation of nitrate is extremely large in summer. The model is unable to capture some of the main features in the diurnal pattern of the PM 2.5 chemical components, suggesting inaccuracies in the presented processes. Potential model adjustments that may lead to a better representation of the boundary layer height, the precursor emissions, hydroxyl radical concentrations, the heterogeneous formation of sulfate and nitrate, and the wet deposition of nitric acid and nitrate have been tested in the sensitivity analysis. The results show that uncertainties in chemistry perhaps dominate the model biases. The proper implementation of heterogeneous sulfate formation and the good estimates of the concentrations of sulfur dioxide, hydroxyl radical, and aerosol liquid water are essential for the improvement of the sulfate simulation. The update of the heterogeneous uptake coefficient of nitrogen dioxide significantly reduces the modeled concentrations of nitrate. However, the large overestimation of nitrate concentrations remains in summer for all tested cases. The possible bias in the chemical production and the wet deposition of nitrate cannot fully explain the model overestimation of nitrate, suggesting issues related to the atmospheric removal of nitric acid and nitrate. A better understanding of the atmospheric nitrogen budget, in particular, the role of the photolysis of particulate nitrate, is needed for future model developments. Moreover, the results suggest that the remaining underestimation of OA in the model is associated with the underrepresented production of SOA.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Materials Data on Co2H36N13Cl5O3 by Materials Project

(Co(NH3)6)4(NO3)2(Cl2)5 crystallizes in the monoclinic C2/m space group. The structure is zero-dimensional and consists of eight azane;cobalt molecules, twenty hydrochloric acid molecules, and four nitric acid molecules.

36 MATERIALS SCIENCE↗

Materials Data on H20SN8O13 by Materials Project

(NH4)4NH4SO4(NO3)3 crystallizes in the monoclinic P2_1 space group. The structure is zero-dimensional and consists of eight ammonium molecules, two azanium;sulfuric acid molecules, and six nitric acid molecules.

36 MATERIALS SCIENCE↗

Materials Data on H20SN8O13 by Materials Project

(NH4)4NH4SO4(NO3)3 crystallizes in the monoclinic P2_1 space group. The structure is zero-dimensional and consists of eight ammonium molecules, two azanium;sulfuric acid molecules, and six nitric acid molecules.

36 MATERIALS SCIENCE↗

Project 18-15263 X-ray Studies of Interfacial Molecular Complexes in ALSEP Back-Extraction (Research Performance Final Report)

The objective of the project was to use synchrotron X-rays to measure the molecular-scale organization of extractants, complexants, buffers, and metal ions at the organic-aqueous (liquid-liquid) interface during solvent back-extraction under conditions relevant to ALSEP (Actinide-Lanthanide Separation Process). The capability of X-ray fluorescence near total reflection (XFNTR) was extended to explore the distribution of Eu(III) between the liquid-liquid interface and bulk organic and aqueous phases which contained extractants HDEHP (bis(2-ethylhexyl) phosphoric acid) and HEHEHP (2-ethylhexyl phosphonic acid mono-2-ethylhexyl ester), the complexant DTPA (diethylene triamine pentaacetic acid), and citric and nitric acids. Although an Am-compatible sample cell was developed, extending x-ray measurements of flat liquid-liquid interfaces to the actinide Am(III) did not prove to be possible within constraints imposed by the Advanced Photon Source. However, preliminary measurements with a drop cell may eventually provide a path forward to investigate Am(III) at a liquid interface. It was also demonstrated that competitive binding that occurs during back-extraction of lanthanide ions to either a phosphoric acid extractant (DHDP, di-hexadecyl phosphate) at the liquid interface or to complexant DTPA in the bulk aqueous phase could be reversibly tuned by X-ray exposure. Separately, X-ray measurements from the liquid-vapor interface explored the role of the aqueous solubility of HDEHP and HEHEHP and their interactions with a range of lanthanides and transition metal ions. Finally, the capability of X-ray absorption spectroscopy (EXAFS) was extended to probe metal ion coordination in lanthanide ion-extractant complexes at the liquid-vapor interface. Complementary to the specific interfacial systems studied during this project, the development of x-ray techniques expands our ability to understand ion ordering at liquid interfaces.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗