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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

The Research of MOX Fuels in Japan

In Japan, uranium and plutonium mixed oxide (MOX) fuels has been developed for sodium-cooled fast reactors. The developing MOX fuels come in variety of O/M ratio, Pu content, minor actinide (MA) content and density. We have studied a science based fuel technology to evaluate fuel behaviors in fabrication process and irradiation condition of such various fuels. The technologies which are constructed based on experimental database can apply to mechanistic evaluation of fuel behaviors. To develop the science based fuel technology, many different varieties of basic properties have been investigated, and experimental database was constructed. And a mechanistic physical property model has been studied. The models contribute to describe various behaviors in fuel fabrication process and irradiation condition. In Japan, the researches on MOX fuel technologies have been carried out for sodium cooled fast reactor development. It is important to develop the fuel technologies on fuel fabrication process and irradiation performance model evaluation. The researches will be carried out by experimental approach and expanded in international collaboration frame works such as the CNWG and The Generation IV International Forum (GIF). (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Computational Design of Advanced Nuclear Fuels

The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Defining the Shape Density of Cerium Oxide and Titanium Dioxide Aerosol

The “shape density” of CeO 2 (cerium oxide) powder (American Elements, Los Angeles CA) is estimated to be 0.82 ± 0.02 g cm -3 and the shape density for TiO 2 (titanium dioxide) powder (American Elements, Los Angeles CA) is estimated as 0.54 ± 0.01 g cm -3 . The “shape density” is defined as the measured loose density (ASTM 2018 and ASTM 2021) of a powder divided by the ICRP (International Commission on Radiological Protection) default aerosol shape factor (Valetin 2002). This information is used for experimental measurements of the RRF (respirable release factor) and other respirable source term parameter (US DOE 1994) experiments at the Los Alamos National Laboratory (LANL). The shape density is the input parameter to the TSI Inc (Shoreview MN) AIM (Aerosol Instrument Manager) software for the TSI model 3321 APS (Aerodynamic Particle Sizer) (Figure 1).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2013

Time cycle analysis and simulation of material flow in MOX process layout

The (U,Pu)O{sub 2} MOX fuel is the driver fuel for the upcoming PFBR (Prototype Fast Breeder Reactor). The fuel has around 30% PuO{sub 2}. The presence of high percentages of reprocessed PuO{sub 2} necessitates the design of optimized fuel fabrication process line which will address both production need as well as meet regulatory norms regarding radiological safety criteria. The powder pellet route has highly unbalanced time cycle. This difficulty can be overcome by optimizing process layout in terms of equipment redundancy and scheduling of input powder batches. Different schemes are tested before implementing in the process line with the help of a software. This software simulates the material movement through the optimized process layout. The different material processing schemes have been devised and validity of the schemes are tested with the software. Schemes in which production batches are meeting at any glove box location are considered invalid. A valid scheme ensures adequate spacing between the production batches and at the same time it meets the production target. This software can be further improved by accurately calculating material movement time through glove box train. One important factor is considering material handling time with automation systems in place.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2015 Report

Methodology development for pore water leaching studies has been continued to support Savannah River Site High Level Waste tank closure efforts. For FY2015, the primary goal of this testing was the achievement of target pH and Eh values for pore water solutions representative of local groundwater in the presence of grout or grout-representative (CaCO 3 or FeS) solids as well as waste surrogate solids representative of residual solids expected to be present in a closed tank. For oxidizing conditions representative of a closed tank after aging, a focus was placed on using solid phases believed to be controlling pH and E h at equilibrium conditions. For three pore water conditions (shown below), the target pH values were achieved to within 0.5 pH units. Tank 18 residual surrogate solids leaching studies were conducted over an E h range of approximately 630 mV. Significantly higher Eh values were achieved for the oxidizing conditions (ORII and ORIII) than were previously observed. For the ORII condition, the target Eh value was nearly achieved (within 50 mV). However, E h values observed for the ORIII condition were approximately 160 mV less positive than the target. E h values observed for the RRII condition were approximately 370 mV less negative than the target. Achievement of more positive and more negative E h values is believed to require the addition of non-representative oxidants and reductants, respectively. Plutonium and uranium concentrations measured during Tank 18 residual surrogate solids leaching studies under these conditions (shown below) followed the general trends predicted for plutonium and uranium oxide phases, assuming equilibrium with dissolved oxygen. The highest plutonium and uranium concentrations were observed for the ORIII condition and the lowest concentrations were observed for the RRII condition. Based on these results, it is recommended that these test methodologies be used to conduct leaching studies with actual Tank 18 residual solids material. Actual waste testing will include leaching evaluations of technetium and neptunium, as well as plutonium and uranium.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers

Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.

42 ENGINEERING↗
DOE OSTI2014

Static electric dipole polarizabilities of An{sup 5+/6+} and AnO{sub 2}{sup +/2+} (An = U, Np, and Pu) ions

The parallel components of static electric dipole polarizabilities have been calculated for the lowest lying spin-orbit states of the penta- and hexavalent oxidation states of the actinides (An) U, Np, and Pu, in both their atomic and molecular diyl ion forms (An{sup 5+/6+} and AnO{sub 2}{sup +/2+}) using the numerical finite-field technique within a four-component relativistic framework. The four-component Dirac-Hartree-Fock method formed the reference for MP2 and CCSD(T) calculations, while multireference Fock space coupled-cluster (FSCC), intermediate Hamiltonian Fock space coupled-cluster (IH-FSCC) and Kramers restricted configuration interaction (KRCI) methods were used to incorporate additional electron correlation. It is observed that electron correlation has significant (∼5 a.u.{sup 3}) impact upon the parallel component of the polarizabilities of the diyls. To the best of our knowledge, these quantities have not been previously reported and they can serve as reference values in the determination of various electronic and response properties (for example intermolecular forces, optical properties, etc.) relevant to the nuclear fuel cycle and material science applications. The highest quality numbers for the parallel components (α{sub zz}) of the polarizability for the lowest Ω levels corresponding to the ground electronic states are (in a.u.{sup 3}) 44.15 and 41.17 for UO{sub 2}{sup +} and UO{sub 2}{sup 2+}, respectively, 45.64 and 41.42 for NpO{sub 2}{sup +} and NpO{sub 2}{sup 2+}, respectively, and 47.15 for the PuO{sub 2}{sup +} ion.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Relationships between experimental signatures and processing history for a variety of PuO 2 materials

The primary oxide of plutonium, PuO 2 , is a complex material due to ongoing change in lattice structure resulting from radioactive alpha decay. In addition, the element plutonium is complex, existing in various allotropes, which can influence formation of oxides due to different nucleation and growth mechanisms. The juxtaposition of these unique properties, along with formation conditions of the initial oxide, makes for PuO 2 materials with a wide variety of local and bulk experimental signatures. This work measures these signatures with X-ray diffraction and Raman spectroscopy from five different oxide materials, and relates these signatures to specific PuO 2 production and aging routes. Furthermore this information is of potential use in nuclear fuels processing and production, nuclear non-proliferation, forensics analyses, and interplanetary power sources.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Milestone report: The simulation of radiation driven gas diffusion in UO 2 at low temperature

Below 1000 K it is thought that fission gas diffusion in nuclear fuel during irradiation occurs through atomic mixing due to radiation damage. This is an important process for nuclear reactor performance as it affects fission gas release, particularly from the periphery of the pellet where such temperatures are normal. Here we present a molecular dynamics study of Xe and Kr diffusion due to irradiation. Thermal spikes and cascades have been used to study the electronic stopping and ballistic phases of damage, respectively. Our results predict that O and Kr exhibit the greatest diffusivity and U the least, while Xe lies in between. It is concluded that the ballistic phase does not sufficiently account for the experimentally observed diffusion. Preliminary thermal spike calculations indicate that the electronic stopping phase generates greater fission gas displacement than the ballistic phase, although further calculation must be carried out to confirm this. A good description of the system by the empirical potentials is important over the very wide temperatures induced during thermal spike and damage cascade simulations. This has motivated the development of a parameter set for gas-actinide and gas-oxygen interactions that is complementary for use with a recent many-body potential set. A comprehensive set of density functional theory (DFT) calculations were used to study Xe and Kr incorporation at a number of sites in CeO 2 , ThO 2 , UO 2 and PuO 2 . These structures were used to fit a potential, which was used to generate molecular dynamics (MD) configurations incorporating Xe and Kr at 300 K, 1500 K, 3000 K and 5000 K. Subsequent matching to the forces predicted by DFT for these MD configurations was used to refine the potential set. This fitting approach ensured weighted fitting to configurations that are thermodynamically significant over a broad temperature range, while avoiding computationally expensive DFT-MD calculations. The resultant gas potentials were validated against DFT binding energies and are suitable for simulating combinations of Xe and Kr in solid solutions of CeO 2 , ThO 2 , UO 2 and PuO 2 , providing a powerful tool for the atomistic simulation of conventional nuclear reactor fuel UO 2 as well as advanced MOX fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Accident tolerant, non-oxidizable MOX fuel fabrication

Thorium dioxide (ThO{sub 2}) has a set of physical-chemical properties that make it well suited to serve as a nuclear fuel matrix: it has a very high melting point, it cannot be further oxidized, it has low leachability in water, reasonable thermal conductivity (better than UO{sub 2}), and of course it is fertile. Collectively, these properties allow for a fuel ceramic with increased in / out of core stability. As such, Thor Energy and a consortium of entities that include Westinghouse, KAERI, Fortum, NNL, ITU and IFE have recently fabricated (Th,Pu)O{sub 2} pellet specimens in Norway. The Th-MOX production methodology was developed over the course of 2 years of ceramic production trials involving Ce-Th oxide mixtures. It was understood that parameters developed for CeO{sub 2} material might need to have been modified if trial results with PuO{sub 2} led to sub-optimal pellets. Among observations made in this work, the most important production parameter was the preparation of constituent / mixed powders. Correct powder morphology was key to achieving high fuel densities. The process parameters developed for (Th,Ce)O{sub 2} pellets were adopted with the expectation that these would serve as a sound basis for obtaining dense, homogeneous (Th,Pu)O{sub 2} fuel ceramic. The Th-MOX fuel manufacture campaign was successful in producing enough (Th,Pu)O{sub 2} pellets to assemble two 30 cm long fuel pins for testing in the Halden test reactor. These fuel rods will be fitted with various instruments to enable temperature, fission gas release and dimensional changes to be measured while operating at power in conditions similar to those of a BWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Study of the U{sub 3}O{sub 8} effect on MOX fuel sintering mechanisms

The manufacturing of uranium-plutonium mixed oxides fuel (also known as MOX) for light water reactor (LWR), starting generally from UO{sub 2} and PuO{sub 2} raw powders and resulting in (U,Pu)O{sub 2} pellets, involves several steps: i) preparation of a UO{sub 2}+PuO{sub 2} master blend with a Pu content of about 25-30 at% by milling and sieving, ii) dilution by mixing the master blend with non-milled UO{sub 2} to reach a Pu content of about 10 at%, iii) pressing and iv) sintering under Ar / 4 % H{sub 2} + H{sub 2}O gas mixture. In this paper, triuranium octoxide (U{sub 3}O{sub 8}) is used as an additional raw powder in the LWR MOX fuel manufacturing. One of the objectives of this work is to understand the sintering behavior of U{sub 3}O{sub 8} under Ar / 4 % H{sub 2} + H{sub 2}O gas mixture and its effect when added to a UO{sub 2}-PuO{sub 2} mixture. A special attention is given on U{sub 3}O{sub 8} reduction, phase transitions, redox reactions and oxygen partial pressure (pO{sub 2}) evolution during shrinkage: the impact of these phenomena on the densification kinetics is discussed, especially the de-densification phenomenon which could occur at high temperature. For that purpose a specific dilatometer, coupled with a zirconia probe implemented at furnace inlet and outlet has been used which allows following continuously and simultaneously the fuel shrinkage and the oxygen partial pressure variation as a function of temperature. The experiments have shown that, contrary to UO{sub 2}, the sintering of milled PuO{sub 2} is accompanied by a de-densification at high temperature. This de-densification phenomenon already occurs with co-milled 71 at% UO{sub 2} + 29 at% PuO{sub 2}, but with a smaller magnitude. When pure milled U{sub 3}O{sub 8} is sintered, no de-densification phenomenon is observed but the low density of the sintered pellet could hide such a phenomenon. The incorporation of U{sub 3}O{sub 8} into a UO{sub 2}-PuO{sub 2} master blend containing 29 at% of plutonium leads to an exacerbation of the de-densification phenomenon with a higher loss on ignition suggesting that U{sub 3}O{sub 8} contributes also to the de-densification phenomenon. This could be due to high temperature reactions involving the initial carbon content of the raw U{sub 3}O{sub 8} powder. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Performance evaluation of two-stage fuel cycle from SFR to PWR

One potential fuel cycle option being considered is a two-stage fuel cycle system involving the continuous recycle of transuranics in a fast reactor and the use of bred plutonium in a thermal reactor. The first stage is a Sodium-cooled Fast Reactor (SFR) fuel cycle with metallic U-TRU-Zr fuel. The SFRs need to have a breeding ratio greater than 1.0 in order to produce fissile material for use in the second stage. The second stage is a PWR fuel cycle with uranium and plutonium mixed oxide fuel based on the design and performance of the current state-of-the-art commercial PWRs with an average discharge burnup of 50 MWd/kgHM. This paper evaluates the possibility of this fuel cycle option and discusses its fuel cycle performance characteristics. The study focuses on an equilibrium stage of the fuel cycle. Results indicate that, in order to avoid a positive coolant void reactivity feedback in the stage-2 PWR, the reactor requires high quality of plutonium from the first stage and minor actinides in the discharge fuel of the PWR needs to be separated and sent back to the stage-1 SFR. The electricity-sharing ratio between the 2 stages is 87.0% (SFR) to 13.0% (PWR) for a TRU inventory ratio (the mass of TRU in the discharge fuel divided by the mass of TRU in the fresh fuel) of 1.06. A sensitivity study indicated that by increasing the TRU inventory ratio to 1.13, The electricity generation fraction of stage-2 PWR is increased to 28.9%. The two-stage fuel cycle system considered in this study was found to provide a high uranium utilization (>80%). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Simultaneous Thermal Analysis of Remediated Nitrate Salt Surrogates

The actinide engineering and science group (MET-1) have completed simultaneous thermal analysis and offgas analysis by mass spectrometry (STA-MS) of remediated nitrate salt (RNS) surrogates formulated by the high explosives science and technology group (M-7). The 1.0 to 1.5g surrogate samples were first analyzed as received, then a new set was analyzed with 100-200mL 10M HNO 3 +0.3 MHF added, and a third set was analyzed after 200 mL of a concentrated Pu-AM spike (in 10M HNO 3 +0.3 MHF) was added. The acid and spike solutions were formulated by the actinide analytical chemistry group (C-AAC) using reagent-grade HNO 3 and HF, which was also used to dissolve a small quantity of mixed, high-fired PuO 2 / AmO 2 oxide.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Point defects and impurities in fluorite PuO 2

The native surface oxide of plutonium plays a critical role in ensuring the stability and safe storage of the underlying metal; consequently, understanding the role of defects and impurities in determining the properties of the oxide layer is critical. Here, in this study, we use hybrid density-functional theory calculations to evaluate the electronic structure and defect chemistry of PuO 2 , the most stable of the native oxide phases, including both native and extrinsic defects. We find that oxygen vacancies (𝑉 O ) form readily in PuO 2 , as do polarons. Electron polarons (𝜂 − ) are the lowest-energy acceptor species in PuO 2 , while the charge compensating donor species will shift from 𝑉 O under O-poor conditions to hole polarons (𝜂 + ) under O-rich conditions. Nitrogen and fluorine can substitute readily for oxygen atoms under O-poor conditions, while fluorine can also incorporate in an interstitial configuration (F$^−_i$) under more O-rich conditions. Carbon and chlorine incorporation in PuO 2 will be very limited. We also evaluate the kinetic barriers for oxygen-related defects, which we find to diffuse readily when present. Our results provide valuable insights into the critical role and variable chemistry of point defects and impurities in PuO 2 , which in turn have important implications for the safe storage of the underlying metal layer. In short, exposure of freshly prepared plutonium to reactive nitrogen- and fluorine-containing contaminants should be avoided, while carbon- or chlorine-containing contaminants are less likely to incorporate readily into the oxide.

Materials science↗
DOE OSTI2015

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Oxygen exchange reaction kinetics for cerium(IV) oxide at 1000 °C

Bulk oxygen exchange rate kinetics on CeO{sub 2} at 1000 °C were observed to have a first order dependence on the fraction of reaction remaining and to be independent of oxygen partial pressure, total pressure, particle size, and specific surface area. This suggests that the exchange reaction is dominated by an internal chemical reaction that is occurring throughout the bulk of the material, and not at the material surface. Oxygen exchange rates were limited by this internal chemical reaction for all CeO{sub 2} powders studied (15 nm to −325 mesh), and had a rate constant of 1.19×10{sup −2} s{sup −1} with a time to completion of 617 s. These results are similar to the exchange rates observed previously on PuO{sub 2}, suggesting that oxygen exchange on PuO{sub 2} may also be dominated by an internal chemical reaction under similar conditions. This work will help guide future experiments on {sup 238}PuO{sub 2} oxygen exchange reactions. - Graphical abstract: Oxygen exchange kinetics on CeO{sub 2} at 1000 °C are independent of a wide range of experimental conditions and exhibit first-order chemical reaction kinetics. - Highlights: • Stable oxygen exchange rates obtained on a variety of CeO{sub 2} powders at 1000 °C. • Exchange rates are independent of atmospheric composition and specific surface area. • Exchange rates are limited by an internal chemical reaction, not a surface reaction. • CeO{sub 2} exchange rates appear similar to the rates observed on PuO{sub 2} at 1000 °C.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗