Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “URANIUM OXIDE”

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 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Nuclear fuels for transient test reactors

Transient test reactors with the ability to test fissile specimens under extreme conditions have been crucial tools in the development of nuclear technologies. Less than 10 unique facility designs have ever been constructed, most of which remain operational today and still use the original nuclear fuel constructed for them more than 40 years ago. Historic fuel systems for transient test reactors vary in significant ways which have marked influences on reactor capabilities. Eventually, new fuel will be needed to support the longevity of transient test reactor missions. This paper reviews precedent transient reactor fuel systems in the context of their unique requirements. A few key conclusions are illustrated by comparing and contrasting these transient test reactors. Fuel composites which are mostly graphite can enable transient reactors with very high neutron fluence capability (>2E16 n/cm 2 ) and are amenable to longer “shaped” transients but cannot achieve pulses <10 ms in duration. Reducing the graphite-to-uranium ratio can yield a very narrow pulse capability but delivers less fluence and requires cores with considerably more fissile material. Designs based on uranium dioxide (UO 2 ) make use of readily available materials to create compact cores with narrow pulse width capabilities but with moderate neutron fluence capabilities (~2E15 n/cm 2 ). Uranium zirconium hydride (U-ZrHx) is a well-established fuel system for pulsing reactors which has been intermittently manufactured throughout the decades. U-ZrHx offers similar capabilities to UO 2 designs in terms of nuclear kinetics, but with about half the fluence capability (~1E15 n/cm 2 ). An evolution of the UO 2 system, termed “ternary ceramic” fuel, shows that dispersing UO 2 in zirconium oxide and calcium oxide can increase fluence capability greatly (~7E15 n/cm 2 ), but is not presently a commonly available fuel form. A unique composite of UO 2 and beryllium oxide (UO 2 -BeO) can be used to create a core with similar kinetics and compact core geometry as U-ZrHx designs, but with significantly higher fluence capability (~6E15 n/cm 2 ). Like ternary ceramic fuel, newly fabricated UO 2 -BeO would require reestablishing its historic manufacturing process which would be further complicated by the health hazards associated with beryllium. In conclusion, like most engineering problems, there is no perfect solution, but this paper outlines the advantages and disadvantages of candidate fuel options to help guide detailed evaluations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Multicomponent pyrochlore solid solutions with uranium incorporation – A new perspective of materials design for nuclear applications

Multicomponent pyrochlore solid solutions with and without uranium incorporation were fabricated and their thermal-mechanical properties were characterized. Multicomponent pyrochlore solid solutions without uranium exhibit comparable thermal conductivity and higher mechanical strength compared to baseline single component rare-earth titanate pyrochlore (A 2 Ti 2 O 7 ). Uranium incorporation reduces hardness as compared with single component compositions. High entropy pyrochlore with uranium displays the highest thermal conductivity within multicomponent pyrochlore solid solutions with significantly better mechanical properties than UO 2 . The measured thermal conductivity correlates well with A-site cation mixing entropy and a modified size disorder parameter, and thus the size disorder and mixing entropy could be good indicators for predicting thermal conductivity of multicomponent pyrochlore solid solutions. This work opens up the possibility of designing multicomponent oxide solid solutions by controlling their chemical disorder/mixing entropy to achieve acceptable thermal-mechanical properties, desired radiation and corrosion performance for potential nuclear waste form and inert matrix fuel applications.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors

Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.

TRISO↗
DOE OSTI2024

Degradation and Failure Phenomena of Advanced Reactor Fuel Concepts: Sodium-Cooled Fast Reactor Metallic Fuel

The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of new fuel types in advanced nuclear power reactors that would be designed and built in the United States. Pacific Northwest National Laboratory (PNNL) is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs that would be deployed in these reactors. This report focuses specifically on the metallic fuel that is being considered for sodium-cooled fast reactors and specifically on mechanisms that would cause damage or failure to the fuel under reactor operating conditions and design basis accident conditions. There is historic experience with both metallic and oxide fuels in experimental sodium-cooled fast reactors, but this report will focus solely on metallic fuel. Currently two U.S.-based reactor designers are engaged with the NRC in the application or pre-application review and considering a sodium-cooled fast reactor. TerraPower is engaged with NRC in pre-application review of its Natrium reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel clad in HT9 stainless steel. ARC Clean Technology is engaged with NRC in pre-application review of its ARC-100 reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel with steel cladding. This report will focus on this fuel system specifically, with broader information given regarding other metallic fuel systems with other stainless steel alloy claddings. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of metallic fuel concepts for sodium-cooled fast reactors, including fuel performance characteristics that may not be addressed within existing regulatory documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1971

Radiation heat transfer calculations for the uranium fuel-containment region of the nuclear light bulb engine.

Calculation results are reviewed of the radiant heat transfer characteristics in the fuel and buffer gas regions of a nuclear light bulb engine based on the transfer of energy by thermal radiation from gaseous uranium fuel in a neon vortex, through an internally cooled transparent wall, to seeded hydrogen propellant. The results indicate that the fraction of UV energy incident on the transparent walls increases with increasing power level. For the reference engine power level of 4600 megw, it is necessary to employ space radiators to reject the UV radiated energy absorbed by the transparent walls. This UV energy can be blocked by employing nitric oxide and oxygen seed gases in the fuel and buffer gas regions. However, this results in increased UV absorption in the buffer gas which also requires space radiators to reject the heat load.

Rodgers, R. J.↗
DOE OSTI2021

Sandwiched Kagomé Lattices in a Coordination Polymer Based on Mixed-Valent Uranium

The metal–organic material U V O(U VI O 2 ) 2 (OH) 5 (Triaz) 2 (Triaz = 1,2,4-triazolate) has been isolated from the reaction of UO 2 (NO 3 ) 2 ·6H 2 O with 1,2,4-triazole in the ionic liquid 1-ethyl-3-methylimidazolium acetate ([C 2 mim][OAc]). The compound’s crystal structure is comprised of planar inorganic layers interconnected by organic linkers into a 3D framework. These layers represent a uranium-based coordination polymer with a Kagomé topology that to the best of our knowledge has never been recognized in f-element coordination chemistry.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Venturing Past Uranium: Synthesis of a Np(IV) Polyoxomolybdate–Alkoxide Sandwich Complex

The synthesis of a Np(IV) polyoxomolybdate– alkoxide sandwich complex, (TBA) 2 [Np{Mo 5 O 13 (OMe) 4 NO} 2 ] (TBA = tetrabutylammonium), is reported. This compound represents a rare example of a neptunium polyoxometalate cluster isolated outside of water, allowing for characterization of its electrochemical properties in nonaqueous solvents. Complexation of An(IV) cations fine-tunes the redox properties of the cluster, with the observed four reversible reductive events varying slightly both in potential and peak separation depending on the actinide present. The new Np(IV) complex also shows an irreversible event assigned to oxidation of Np(IV) to Np(V). New methodology for facile 17 O enrichment of (TBA) 2 [Mo 5 O 13 (OMe) 4 NO][Na(MeOH)] is presented, which provides a simple pathway to 17 O enriched analogues of the sandwich complexes discussed (Zr(IV), Hf(IV), Th(IV), U(IV), U(V), Np(IV)). 17 O NMR spectroscopy subsequently provides insights into both the nature of metal–oxygen bonding, as well as the influence of unpaired f-electrons on the local environment of the oxygen nuclei.

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

Multiple Sources of Riparian Wetland Suspended Solids during Episodic Rain Events: Influence on Uranium Transport

Suspended solids can be the primary vector for transporting contaminants in streams. The objective of this study was to determine whether changes in the properties of suspended solids during rain events impacted contaminant transport. Stream water was collected during five episodic events downstream from a U-contaminated wetland located in South Carolina, USA. The suspended particles were initially composed of Fe-flocs (particles formed in situ prior to the rain event) that had significantly greater Fe, Mn, organic-C, and U content than particles collected later during a sampling rain event. XANES and EXAFS revealed that U in the Fe-flocs was U(VI) and that it was not incorporated in a mineral structure but existed as inner- or outer-sphere adsorbed uranyl species associated with organic matter and Fe-oxides. The uranyl had an extraordinarily high affinity for the suspended solids, with solid to liquid U ratios of >72,000 (μg/kg)/(μg/L). After the initial flush of Fe-flocs, a greater fraction of the suspended solids had lower organic-C, Fe, Mn, and amorphous phases and were composed of more quartz, kaolinite, and gibbsite, resulting in lower U concentrations than those in the solids collected earlier in the rain event. This study highlights the importance of understanding suspended solids as transport vectors and their potential dynamic nature during rain events.

computer simulations↗
DOE OSTI2024

Development of an Attenuated Total Reflectance–Ultraviolet–Visible Probe for the Online Monitoring of Dark Solutions

Optical spectroscopy is a valuable tool for on-line monitoring of a variety of processes. Ultraviolet-visible (UV-vis) spectroscopy in particular, can monitor the concentration of analytes as well as identify speciation and oxidation state. However, it can be difficult to impossible to employ UV-vis based sensors on chemical systems that are very dark (i.e., high optical density) as exceedingly short pathlengths are required (for transmission approaches) or effective means of backscattering are needed (for reflectance approaches). Examples of processes that would benefit significantly from the use of optical sensors and encounter these challenges include used nuclear fuel recycling and molten salts with high concentrations of dissolved uranium. Utilizing an attenuated total reflectance (ATR) UV-vis approach can overcome these challenges and allow for the measurement of solutions orders of magnitude more concentrated than transmission UV-vis. However, determining ideal sensor specifications for varied processes can be time consuming and expensive. Here, in this study, we evaluate the ability for a novel ATR-UV-vis probe to measure very concentrated solutions of Co(II) and Ni(II) nitrate as well as organic dyes (methylene blue, acid red 1, and crystal violet). This sensor design provides a modular method for exploring possible “pathlengths” by altering the exposed ATR fiber length. Also studied were approaches to loading and measuring the sensor cell. These results are compared to a traditional 1 cm cuvette measured by transmission UV-vis. It was found that the ATR-UV-vis probe was capable of measuring solutions 600 times more concentrated than the 1 cm cuvette. Advanced data analysis in the form of multivariate curve resolution (MCR) was used to analyze the speciation of methylene blue over a large concentration range. The application of this novel ATR-UV-vis probe to the interrogation of dark solutions is a promising avenue for use in on-line monitoring of nuclear processes.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

Preliminary MITR Design-Demonstration Element Thermal-Hydraulics and Structural Analyses

The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). Massachusetts Institute of Technology Reactor (MITR) is one of five research reactors selected for this program. The objective of this report is to provide preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the MITR Design Demonstration Element (DDE) under conservative approximations for the plate power distribution. This report provides details on the modeling approach and the simulation results obtained, including pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment for the irradiation cycles in the Belgian Reactor (BR)-2.

42 ENGINEERING↗
DOE OSTI2020

Fabrication and thermophysical properties of UO 2 -UB 2 and UO 2 -UB 4 composites sintered via spark plasma sintering

Uranium dioxide (UO 2 ) composites with uranium diboride (UB 2 ) and uranium tetraboride (UB 4 ) have been proposed as advanced fuel candidates due to their high thermal conductivity, high melting point, high fissile density and their ability to incorporate a built-in burnable poison by tailoring the targeted 10 B/ 11 B ratio. As such, it is important to assess the fabrication, and thermal and micromechanical properties of such composites. In this work, UO 2 -UB 2 and UO 2 -UB 4 samples with boride phase fractions of 5, 15 and 30 wt% were fabricated to high densities (above 95 % theoretical density) via spark plasma sintering (SPS). This enabled sintering at relatively low temperatures and short timescales. SPS also aided in maintaining the target phase fractions of the samples as reactions between the constituent phases were suppressed due to the short timescales and reducing environment during sintering. Here, thermal diffusivity measurements from 299 to 1273 K were conducted through laser flash analysis (LFA). The diffusivity increased as a function of boride weight fraction, and UB 2 additions increased the thermal diffusivity of the composites more than UB 4 additions. Assessment of the LFA results indicated in-situ reactions between the UO 2 and boride phases that suppress the thermal diffusivity occur above 800 K for all samples. Oxidation of the boride phase was proposed as the underlying reaction. This was supported by thermodynamic assessments from the literature, as well as microstructural, crystallographic, and nanoindentation characterization performed on these samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Preliminary NBSR Design-Demonstration Element Thermal-Hydraulics and Structural Analyses

1.1 OBJECTIVE The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). The National Bureau of Standards Reactor (NBSR) is one of five research reactors selected for this program. The objective of this report is to provide preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the NBSR Design Demonstration Element (DDE) under conservative approximations for the plate power distribution. This report provides details on the modeling approach and the simulation results obtained, including pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment for the irradiation cycles in the Belgian Reactor (BR)-2.

42 ENGINEERING↗
DOE OSTI2024

Nuclear Fuel and Pu Redox Studies from The Glenn T. Seaborg Institute at Idaho National Laboratory

The Glenn T. Seaborg Institute at Idaho National Laboratory (INL-GTSI) focuses on advancing fundamental research in the actinide sciences by providing unique opportunities to early career scientists and engineers to gain experience studying the actinide elements and their associated systems. The INL-GTSI is built from three focus areas that are based on the expertise and supporting infrastructure at INL and include solid state chemistry and physics, solution phase chemistry and physics, and forensic and isotope science. INL is the lead Laboratory for nuclear energy research and development in the U. S. and the research on nuclear fuels performed under the INL-GTSI gives good examples of solid state studies. Uranium-Molybdenum (U-Mo) alloys are leading fuel candidates for conversion of high performance research and test reactors to low-enriched fuels. During irradiation, generated fission gas accumulates into bubbles and self-organizes into a gas bubble superlattice (GBS) that effectively stores fission gases and inhibits fuel swelling. A study on the early self- organizing behavior of the GBS shows that not only grain boundaries but the interfaces between the U-Mo matrix and uranium carbide (UC) impurities are important to GBS formation.[1] In solution, understanding the complex redox behavior of plutonium in aqueous environments is critical for establishing optimized nuclear waste reprocessing solvent systems and storage tank environments. INL-GTSI researchers have produced an experimentally validated multi-scale model of the gamma radiation induced behavior of plutonium ions in concentrated aqueous HNO3 solutions.[2] Here, gamma radiation effected only minimal steady state changes in the redox distribution of the plutonium oxidation states. The redox cycling between Pu(IV) and Pu(III) is demonstrated to be mediated by the •OH/NO3• radical oxidation of Pu(III) and the H2O2/HNO3 driven reduction of Pu(IV). The INL-GTSI offers young researchers the unique chance to work directly with actinide bearing materials in a U. S. National Laboratory environment. Further topical areas of interest to the INL-GTSI include, but are not limited to, fundamental actinide properties, structure/property (electronic, magnetic, thermal) relations, actinide quantum criticality, f- electron interactions, electron correlations, computational studies, new phases, defect effects, interface interactions, isotope production and separation, forensic analytical chemistry, structure and dynamic properties of actinides in non-aqueous media, separations chemistry and kinetics for advanced nuclear fuel cycles, radiation effects, and innovative and advanced ligand design for complexation of the actinides.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Hyperspectral X-ray Imaging with TES Detectors for Nanoscale Chemical Speciation Mapping

We are developing an imaging capability (“Hyperspectral X-ray Imaging”) for mapping chemical information (molecular formula, phase, oxidation state, hydration) that is based on ultra-high-resolution X-ray emission spectroscopy with large transition-edge sensor microcalorimeter arrays in the scanning electron microscope. By combining microcalorimeter arrays with hundreds of pixels, high-bandwidth microwave frequency-division multiplexing, and fast digital electronics for near real-time data processing, our goal is to enable measurements using laboratory-scale instrumentation rather than synchrotron beamlines. Our application focus here is on mapping the chemical form of uranium compounds on the nanoscale. Furthermore, we will present our approach to developing the Hyperspectral X-ray Imaging capability, progress toward a 128-pixel microwave multiplexed X-ray fluorescence instrument at LANL, and the path to high-throughput nanoscale chemical mapping.

47 OTHER INSTRUMENTATION↗
DOE OSTI2021

Combined Technologies for In Situ Remediation of Tc-99 and U in Subsurface Sediments

In this study, combinations of chemical remedies were tested in bench-scale batch experiments to evaluate a two-step reduction-sequestration approach to effectively stabilize high concentrations of inorganic contaminant mixtures. Bench tests simulated contaminant and geochemical conditions of a perched aquifer located within the Central Plateau at the Hanford Site, located in southeastern Washington State (USA). Pairwise combinations of a reductant [e.g., zero valent iron, sulfur modified iron (SMI), or calcium polysulfide] and a sequestering agent [e.g., calcite, apatite, or dilute alkaline solution (e.g., NaOH)] were evaluated for immobilization and stabilization of technetium (Tc) (50,000 pCi/L), uranium (U) (150 mg/L), and nitrate (NO 3 ) (200 mg/L) in high ionic strength groundwater. The results of these batch studies demonstrated that reduction by SMI and sequestration in apatite or calcite are the most effective combination for these contaminant mixtures and conditions. Aqueous concentrations of Tc and U decreased by 95.6% ± 2.5% and 101.1% ± 5.2%, respectively, with SMI-apatite and 98.3% ± 0.0% and 101.2% ± 5.2%, respectively, with SMI-calcite. Sequential extractions showed that sequestered contaminants had limited capacity for re-oxidation; in fact, less than 10% of immobilized Tc and U was recovered by selective extraction of mineral phases most susceptible to oxidation. In addition, X-ray absorption near edge structure analysis of the sediment samples treated with SMI-calcite showed the presence of only U(IV), while both U(IV) and U(VI) were present in the SMI apatite combination [ratio of 0.43 U(IV):0.59 U(VI)].This study describes preliminary results that a two-step approach for stabilizing contaminant mixtures of long-lived radionuclides can be effective at reducing contaminant fluxes to groundwater from vadose and perched water zones.

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

Thermodynamics of CeSiO 4 : Implications for Actinide Orthosilicates

Zircon (ZrSiO 4 : I4 1 /amd) can accommodate actinides, such as thorium, uranium, and plutonium. The zircon structure has been determined for several of the end member compositions of other actinides, such as plutonium and neptunium. However, the thermodynamic properties of these actinide zircon structure-types are largely unknown due to the difficulties in synthesizing these materials and handling transuranium actinides. Thus, we have completed a thermodynamic study of cerium orthosilicate, stetindite (CeSiO 4 ), a surrogate of PuSiO 4 . For the first time, the standard enthalpy of formation of CeSiO 4 was obtained by high temperature oxide melt solution calorimetry to be -1971.9 ± 3.6 kJ/mol. Stetindite is energetically metastable with respect to CeO 2 and SiO 2 by 27.5 ± 3.1 kJ/mol. The metastability explains the rarity of the natural occurrence of stetindite and the difficulty of its synthesis. Applying the obtained enthalpy of formation of CeSiO 4 from this work, along with those previously reported for USiO 4 and ThSiO 4 , we developed an empirical energetic relation for actinide orthosilicates. Here, the predicted enthalpies of formation of AnSiO 4 are then made with a discussion of future strategies to efficiently immobilize Pu or minor actinides in the zircon structure.

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

Chemical Modeling of the Capture of Radioiodine Species

Iodine is a common fission product resulting from the transmutation of uranium fuel in nuclear power reactors. The short-lived radioactive isotope, 131 I, with a half-life of around eight days, would be a particular health concern if released into the environment. Capture materials, such as activated carbon, are used routinely in nuclear power plants and medical isotope production facilities to remove this hazard during normal operations to reduce emissions to acceptable levels. Iodine, unlike other common fission products such as noble gases, is challenging because of the wide range of molecular forms and oxidation states (from +VII in IF 7 to -I in iodide) that it displays, which can interconvert both in the gas phase and in the solution phase for example. The work in this project is focused on three computational tasks. In the first, we consider the binding of iodine-containing molecules of varying oxidation states to the major component of activated carbon – pristine graphene-like sheets. The second task focused on molecular binding and possible chemical transformations resulting from the reaction of molecular iodine with chemical functional groups on the edge of the graphene sheets or surface defects in the sheet itself. The final task was associated with radiolytic breakdown of 131 I to 131 Xe in iodine containing molecules and investigated the resulting molecular fragmentation process both in the gas phase and on the graphene-like surface. For each of the tasks, the intent was to learn about the thermodynamics and kinetics of the chemical transformations and use the calculations to provide parameters for input into kinetic models describing these processes for interpretation of experimental data in future work.

07 ISOTOPE AND RADIATION SOURCES↗