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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 1,657 records · Page 92
All sourcesNASA NTRSDOE OSTI
NASA NTRSDate not supplied

Entry Guidance Design and Post-Flight Performance of the Mars 2020 Mission

Like its predecessor Mars Science Laboratory, the Mars 2020 mission successfully utilized a derivative of the Apollo Entry Terminal Point Controller guidance algorithm, whereby bank angle controls range flown along a trajectory. The flight performance of this algorithm in conjunction with a range-trigger for parachute deploy delivered the Perseverance rover to 1.7 km from the expected touchdown location within an ellipse of 7.5 x 5.2 km. This miss distance is largely attributed to atmospheric and aerodynamic modeling uncertainties between the design and as-flown trajectory. This algorithm for guided entry continues to provide a solid basis for Mars missions to improve upon.

Guidance↗
NASA NTRSDate not supplied

Post-Flight EDL Entry Guidance for the Mars 2020 Mission

Like its predecessor Mars Science Laboratory, the Mars 2020 mission successfully utilized a derivative of the Apollo Entry Terminal Point Controller guidance algorithm, whereby bank angle controls range flown along a trajectory. The flight performance of this algorithm in conjunction with a range-trigger for parachute deploy delivered the Perseverance rover to 1.7 km from the expected touchdown location within an ellipse of 7.5 x 5.2 km. This miss distance is largely attributed to atmospheric and aerodynamic modeling uncertainties between the design and as-flown trajectory. This algorithm for guided entry continues to provide a solid basis for Mars missions to improve upon.

Guidance↗
DOE OSTI2015

Post-Irradiation Examination of 237Np Targets for 238Pu Production

Oak Ridge National Laboratory is recovering the US 238 Pu production capability and the first step in the process has been to evaluate the performance of a 237 Np target cermet pellet encased in an aluminum clad. The process proceeded in 3 steps; the first step was to irradiate capsules of single pellets composed of NpO 2 and aluminum power to examine their shrinkage and gas release. These pellets were formed by compressing sintered NpO 2 and aluminum powder in a die at high pressure followed by sintering in a vacuum furnace. Three temperatures were chosen for sintering the solution precipitated NpO 2 power used for pellet fabrication. The second step was to irradiate partial targets composed of 8 pellets in a semi-prototypical arrangement at the two best performing sintering temperatures to determine which temperature gave a pellet that performed the best under the actual planned irradiation conditions. The third step was to irradiate ~50 pellets in an actual target configuration at design irradiation conditions to assess pellet shrinkage and gas release, target heat transfer, and dimensional stability. The higher sintering temperature appeared to offer the best performance after one cycle of irradiation by having the least shrinkage, thus keeping the heat transfer gap between the pellets and clad small minimizing the pellet operating temperature. The final result of the testing was a target that can meet the initial production goals, satisfy the reactor safety requirements, and can be fabricated in production quantities. The current focus of the program is to verify that the target can be remotely dissembled, the pellets dissolved, and the 238 Pu recovered. Tests are being conducted to examine these concerns and to compare results to code predictions. Once the performance of the full length targets has been quantified, the pellet 237 Np loading will be revisited to determine if it can be increased to increase 238 Pu production.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Amine-incorporated porous polymer networks (PPNs) as sorbents for post-combustion CO2 capture

The Zhou group has long been interested in the development of CO2 capture sorbents, having investigated both metal-organic frameworks (MOFs) and porous polymer networks (PPNs). In this work, we have begun an investigation of a new generation of PPN based sorbents. These PPN sorbents are low cost alternatives to expensive MOF sorbents, while achieving the low heats of adsorption necessary for economic CO2 capture.

post-combustion carbon capture, porous polymer net↗
DOE OSTI2020

AGR-5/6/7 Post-Irradiation Examination Plan

This plan describes the PIE activities to be carried out on AGR-5/6/7 fuel and the non-fuel components of the AGR-5/6/7 irradiation test train. PIE of AGR-5/6/7 fuel and test train components will provide data on fuel performance over a range of irradiation conditions (e.g., burnup, neutron fluence, and temperature), test fuel under postulated accident conditions, and support development of fuel performance and fission product transport models. Examination of the non-fuel test train components will give information on TRISO fuel and compact performance and information on releases of radioactive fission products from the fuel compacts. PIE of the fuel compacts is used for many purposes including, but not limited to: updating the irradiation thermal calculations, measuring burnup, determining retention/release of fission products from the compacts and the TRISO particles, determining TRISO layer failure rates, determining fuel performance at elevated accident temperatures, determining fuel performance under oxidizing atmospheres, determining the distribution of fission products within the compact and fuel particles, and observing compact and particle morphology to explain mechanisms of TRISO layer failure and/or fission product release/retention.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Post-irradiation examination to assess performance and safety of nuclear fuel

This work presents some results from testing campaigns performed at JRC-ITU on irradiated fuels and analogues aimed at characterizing fuel alterations during and after irradiation. The characterization of local structural and property changes is the first step towards the description of the overall behaviour of the fuel, including safety- and performance-relevant properties. High burnup light water reactor (LWR) UO{sub 2} fuel (= 60 GWd/tHM) was used for morphology and mechanical property studies. Radial profiles of porosity obtained from scanning electron microscopy image analysis were compared with corresponding sub-surface data obtained using acoustic microscopy. An increase of porosity was observed at the rim of the pellet in correspondence with the high burnup structure (HBS), and also at intermediate radius locations in correspondence with the occurrence of the so-called dark zone. This zone is characterized by several features similar to the HBS: in addition to increased local porosity, grain subdivision features were observed. The examination of grain free surface morphology along the pellet radius revealed the onset of surface alteration which resulted in almost complete loss of the original grain shape in conjunction with the formation of the dark zone. Additionally, possible ageing effects due to accumulation of microstructural decay damage during storage were investigated and compared to the results obtained under accelerated ageing conditions using alpha-doped analogues. Similar hardening trends were observed by comparing the hardness evolution as a function of accumulated damaged in unirradiated analogues and spent fuel. No significant hardness increase was observed on the same high burnup/high dose spent fuel after >10 years of additional storage in hot cell, confirming that hardening saturates at a damage level near 0.1 dpa. These experimental campaigns contribute establishing a basis of data for further analysis, in combination with suitable modelling tools. The final goal is to achieve the full scientific understanding of mechanisms and processes affecting fuel properties and behaviour, as a necessary step to establish the basis for developing advanced fuel systems with optimized performance and safety. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Plan for Post-Irradiation Examination of Zion Reactor Pressure Vessel Beltline Materials

The reactor pressure vessel (RPV) is a potentially life-limiting component in light-water reactors (LWR) because replacement of the RPV is not considered a viable option. Researchers studying the effects of radiation on RPV materials have long been interested in evaluating service-irradiated materials to assess currently available models for predicting radiation embrittlement of RPV steels. The decommissioning of the Zion Unit 1 NPP in Zion, Illinois presents a unique opportunity for characterization of an actual reactor pressure vessel (RPV) material after inservice degradation. The main interests in this case are the belt-line weld and base metals. Through-RPV wall thickness attenuation and chemical composition distributions are the obvious objectives of this characterization program. However, it also provides an opportunity to examine the inservice degradation and its comparison with currently available models for prediction of radiation embrittlement of RPV steels. All of these efforts will provide a better understanding of materials degradation and other issues associated with extending the lifetime of existing NPPs beyond 60 years of service. The primary foci are the Linde 80 flux, wire heat 72105 (WF-70) circumferential belt-line weld and the A533B-1 base metal from the intermediate shell harvested from a region of peak fluence (0.7 x 10{sup 19} n/cm{sup 2}, E > 1.0 MeV) on the internal surface of the Zion Unit 1 vessel. Following determination of the through-thickness chemical composition, Charpy impact, fracture toughness, tensile, and hardness testing will be being performed to characterize the through-thickness mechanical properties of the base metal and belt-line-weld materials. In addition to mechanical properties, microstructural characterizations will be performed using various techniques, including Atom Probe Tomography, Small Angle Neutron Scattering, and Positron Annihilation Spectroscopy.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Post-Irradiation Examination for Conventional and Advanced Nuclear Fuel

During irradiation in a nuclear reactor, the fuel pellet experiences non-uniform fission rate and strong thermomechanical loads, which result in alterations and restructuring occurring non-homogeneously in the fuel volume. The irradiated fuel is characterized by a highly heterogeneous structure, with a correspondingly heterogeneous distribution of phases and properties. It is a big technical challenge for a hot cell facility to be able to cover effectively a broad variety of fuel types and compounds, characterized by different compositions, physico-chemical properties, geometries and configurations. The first challenge is to provide testing environments and procedures which do not alter or degrade the original irradiated fuel properties and are technically suitable for hot cell or shielded facility operation. For instance, non-oxide fuel compounds require protective atmosphere to avoid sample oxidation during storage, preparation and testing. Moreover, suitable adaptations of tools and measuring devices must be available to characterize fuel configurations which differ from the conventional cylindrical pellet-cladding geometries (e.g. spherical fuel elements for pebble bed HTR and TRISO coated fuel particles). Finally, adequate scientific methods of investigation must be available to measure properties and study the behaviour of highly heterogeneous irradiated fuel. In addition to basic techniques for non-destructive and destructive examination of nuclear fuel systems (e.g. covering fuel and cladding or fuel and coatings, or other configurations), 'in-depth' methods are applied to determine physical, thermomechanical and micro-analytical properties of specific phases/features present in irradiated fuel. This paper presents relevant trends and examples from fuel safety studies performed at JRC-Karlsruhe and in other laboratories which highlight the 'added value' obtainable by applying combined techniques, high resolution tools or 3D methods of analysis for the experimental 'multiscale' characterization of irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Mathematical modeling of a moving bed reactor for post-combustion CO 2 capture

A mathematical model for a moving bed reactor with embedded heat exchanger has been developed for application to solid sorbent-based capture of carbon dioxide from flue gas emitted by coal-fired power plants. The reactor model is one-dimensional, non-isothermal, and pressure-driven. The two-phase (gas and solids) model includes rigorous kinetics and heat and mass transfer between the two phases. Flow characteristics of the gas and solids in the moving bed are obtained by analogy with correlations for fixed and fluidized bed systems. From the steady-state perspective, this work presents the impact of key design variables that can be used for optimization. From the dynamic perspective, the article shows transient profiles of key outputs that should be taken into account while designing an effective control system. Additionally, the article also presents performance of a model predictive controller for the moving bed regenerator under process constraints.

42 ENGINEERING↗
DOE OSTI2015

Post‐Assembly Stabilization of Rationally Designed DNA Crystals

Abstract This manuscript reports an effort to stabilize self‐assembled DNA crystals. Owing to their weak inter‐unit cohesion, self‐assembled DNA crystals are fragile, which limits the potential applications of such crystals. To overcome this problem, another molecule was introduced, which binds to the cohesive sites and stabilizes the inter‐unit interactions. The extra interactions greatly improve the stability of the DNA crystals. The original DNA crystals are only stable in solutions of high ionic strength (e.g., ≥1.2 M (NH 4 ) 2 SO 4 ); in contrast, the stabilized crystals can be stable at ionic strengths as low as that of a 0.02 M solution of (NH 4 ) 2 SO 4 . The current strategy is expected to represent a general approach for increasing the stability of self‐assembled DNA nanostructures for potential applications, for example, as structural scaffolds and molecular sieves.

Zhao, Jiemin↗
DOE OSTI2015

Post‐Assembly Stabilization of Rationally Designed DNA Crystals

Abstract This manuscript reports an effort to stabilize self‐assembled DNA crystals. Owing to their weak inter‐unit cohesion, self‐assembled DNA crystals are fragile, which limits the potential applications of such crystals. To overcome this problem, another molecule was introduced, which binds to the cohesive sites and stabilizes the inter‐unit interactions. The extra interactions greatly improve the stability of the DNA crystals. The original DNA crystals are only stable in solutions of high ionic strength (e.g., ≥1.2 M (NH 4 ) 2 SO 4 ); in contrast, the stabilized crystals can be stable at ionic strengths as low as that of a 0.02 M solution of (NH 4 ) 2 SO 4 . The current strategy is expected to represent a general approach for increasing the stability of self‐assembled DNA nanostructures for potential applications, for example, as structural scaffolds and molecular sieves.

Zhao, Jiemin↗