Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear Reactor Fuel”

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
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Hydrolysis Reaction Pathways of Thorium Oxide Nanoclusters

Density functional theory benchmarked by correlated molecular orbital theory is used to develop a fundamental and predictive understanding of the interaction of thorium oxide nanoclusters with gas phase water to provide insight into nuclear-waste storage, production of thorium nuclear reactor fuels, and reprocessing of spent fuel. The structures of Th n O 2n (n = 3 – 6) clusters and their interactions with water have been studied at the B3LYP, MP2, and CCSD(T) levels. Hydrolysis is initiated by the formation of Lewis acid-base adducts, with relative H 2 O binding energies (physisorption) ranging from −15 kcal/mol to −22 kcal/mol. The initial H 2 O physisorption energy is ca. −21 kcal/mol regardless of the cluster size and is consistent with the experimentally obtained initial adsorption energy on a thorium dioxide surface. The physisorption enthalpies for additional water molecules can be affected by the presence of terminal groups OH groups generated by proton transfer to a Th-O near the site of adsorption. The hydrolysis products (chemisorption) form either bridging or terminal hydroxides. More exothermic hydrolysis steps were predicted for the formation of terminal hydroxides as compared to the formation of bridging hydroxides. Here, the calculated transition state barriers for transfer of protons from bound water complexes to form the chemisorption products are very low. Overall, water readily reacts with thorium oxide clusters preferring hydroxide products over hydrated complexes. First and second order fits were predicted for the combined physisorption and chemisorption energies for the hydrolysis of thorium oxide clusters. Finally, ionization energies and electron affinities were calculated as were HOMO-LUMO gaps to provide additional insights into the properties of the thorium oxide and hydroxide clusters.

Adsorption↗
NASA NTRS1977

Heat pipe reactors for space power applications

A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

Koenig, D. R.↗
NASA NTRS1982

Overview of high-temperature materials for high-energy space power systems

The current state of technology and some of the more pressing research needs and challenges associated with the possible use of high temperature materials in future high energy space power systems are discussed. Particularly, emphasis is on the need to improve and quantify the fundamental understanding of the effects of the following: (1) fast neutron radiation on the properties and behavior of nuclear reactor fuels and claddings; and (2) long term, high temperature, space (vacuum) exposure on the properties of refractory metals considered for use as structural materials in various power conversion systems.

Saunders, N. T.↗
NASA NTRS2019

Near-Term Options for a Nuclear Thermal Propulsion Flight Demonstrator

The Appropriations Bill passed by the US Congress in February 2019 instructed NASA to direct “not less than $100,000,000 for the development of nuclear thermal propulsion, of which not less than $70,000,000 shall be for the design of a flight demonstration by 2024 for which a multi-year plan is required by both the House and the Senate within 180 days of enactment of this agreement." As part of NASA’s response to this direction, the Advanced Concepts Office (ACO) at the Marshall Space Flight Center (MSFC) was tasked with leading a study to develop a nuclear thermal propulsion (NTP) flight demonstration (FD) concept and evaluate its feasibility with respect to the near-term schedule goal. During formulation for the NTP FD study, two perspectives emerged with regards to FD concept design. The first seeks to strictly observe the immediate near-term schedule goal, embracing a completely “off-the-shelf,” high-TRL approach to subsystem design and component selection. The downside to this approach is that the propulsion performance to be expected from such a design is significantly lower than what NTP promises for operational systems, and the value of the flight demo is potentially reduced due to a lack of traceability. The second approach advocates for an FD concept that shows increased traceability to the projected designs of operational systems, providing risk reduction for future NTP-enabled missions. This option comes at the cost of schedule and development risks, as it requires some new investments in nuclear reactor fuels and design. In order to understand the implications and differences between these two approaches, the ACO team elected to perform a concept design of each type, labeling the immediate near-term concept Flight Demo 1 (FD1), and the higher traceability concept Flight Demo 2 (FD2). This paper will present a summary of the mission profiles and system designs for both FD1 and FD2, identifying key drivers and challenges for each design.

Edwards, S. J.↗
DOE OSTI2022

Fast modular reactor nuclear design parameters of fuel cycle and power distributions

This paper describes conceptual design of a 112 MW thermal (50 MW electric) Fast Modular Reactor (FMR) system operating at 7 MPa with inlet/outlet temperatures of 509/800 Celsius degrees. The reactor system includes the reactor core, fuel assemblies, fuel rods, reactor internals, reflector, neutron control system, flow control, and structural components. The nuclear design and analysis were conducted to search for a baseline core with a cycle length greater than 8-year and the power peaking factor less than 1.5 by adjusting the fuel assembly, reactor core, and reflector configurations. The neutron transport calculations of the baseline core showed that a refueling interval of 9-year is achieved with a total peaking factor of 1.47 and a fuel rod average linear power of 3.6 kW/m. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Reactor and fuel cycle performance of light water reactor fuel with 235 U enrichments above 5%

Recent advances in nuclear fuel materials research, particularly on the topic of accident-tolerant fuels, have brought up potential opportunities for expanding the operating envelope of existing light water reactors. As many of the performance improvements offered by these technologies may be most fully realized by increasing fuel enrichment beyond the standard 5% limit, this paper examines the potential reactor performance and fuel cycle performance of low-enriched uranium oxide fueled light water reactors by generically considering pressurized water reactors with 235 U enrichment from 5 to 7%. Advanced cladding, including accident-tolerant cladding, has the potential to increase fuel burnup limits related to hydrogen in the cladding that coincide with those limits associated with end-of-life reactivity. Therefore, higher enrichment will be necessary in order to realize the higher fuel burnups. This work includes evaluation of the fuel cycle length, discharge burnup, reactivity coefficients, and fuel cycle performance, including radioactive waste and environmental impact metrics per unit energy generated. The analysis was performed using the evaluation metrics from the US Department of Energy Office of Nuclear Energy Fuel Cycle Evaluation and Screening Study. The reactor performance and safety analysis show that enrichments between 5 and 7% would have similar fuel temperature and moderator temperature coefficients. However, the soluble boron coefficient would decrease in magnitude, requiring more corrosive boric acid in the coolant or other methods of reactivity control during the fuel cycle. At these higher enrichments the maximum burnup at the rim of the fuel pellet would increase by almost a factor of two, which is expected to impact the formation of high-burnup structure in the fuel and the corresponding thermo-mechanical fuel properties. The fuel cycle performance assessment shows that increasing enrichment reduces the quantity of high-level waste disposed per unit energy generated, but it increases the natural resource requirements normalized to a gigawatt-electricity-per-year basis. Another impact is the slightly higher discharge burnup, resulting in somewhat different activity levels of the spent nuclear fuel and high-level waste radioactivity at 100 and 100,000 years after fuel discharge. The environmental impacts—including land use, water use, carbon emission, and radiological exposure—are of the same magnitude per unit energy generated. Finally, however, the impacts are distributed differently. Less than 5% enrichment has marginally more impact on the back-end of the fuel cycle, and greater than 5% enrichment has marginally more impact on the front-end of the fuel cycle. Ultimately, no neutronic or reactor safety hindrances to employing light water reactor fuel with enrichments greater than 5% are identified; given the achievable reactor performance benefits with advanced fuels, further practical exploration of increased enrichment fuel is recommended.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2022

November 2022 NS&T Highlights

These are the highlights for the Nuclear Science and Technology organization for the month of November.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A Framework to Assess Advanced Reactor Spent Fuel Management Facility Deployment

Advanced nuclear reactors offer various operational advantages over existing light water reactors but could produce types of spent nuclear fuel (SNF) with a wide variety of forms and characteristics depending on how many different concepts are deployed. Each advanced reactor SNF type potentially poses unique management challenges. New planning efforts will be necessary to anticipate how the management requirements of advanced reactor SNF will affect the deployment of an integrated waste management system. This paper applies a framework of high-level facility deployment milestones to a generic SNF management system, reviewing them together with the advanced reactor SNF characteristics and management requirements. This allows for the investigation of factors that influence facility and system deployment, and ultimately, the identification of challenges facing the deployment of different kinds of SNF management facilities. Here, the back end of the once-through fuel cycle is examined for four advanced reactor system technology types: sodium-cooled fast reactors, high-temperature gas-cooled reactors, liquid-fuel molten salt reactors, and lead-cooled fast reactors. It is observed that milestones earlier in the facility deployment process (e.g., siting and facility design) are more impacted by the uniqueness of advanced reactor SNF characteristics than others (e.g., construction and testing). Ultimately, none of the differences are seen as fundamentally disqualifying in a technical sense; however, they should be considered early, potentially as part of reactor design, to avoid issues in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Nguyen, Kien-Cuong↗
DOE OSTI2021

Metallic Fuel Performance Code Requirements for the Versatile Test Reactor Project

Metallic nuclear fuels have been proposed for use in several emerging nuclear reactor designs, and a number of codes have been developed to model these fuels and assess their performance. Qualification of these metallic nuclear fuels will ultimately require monitored irradiation of lead test assemblies, but the use of fuel performance codes can reduce the uncertainty associated with these efforts by quantifying uncertainties and estimating margins to failure ahead of time. In this work, metallic fuel performance code requirements are defined for the Versatile Test Reactor (VTR) project using input from the Experimental Breeder Reactor II functional requirements, operational requirements, and design criteria. This work focuses on the thermomechanical responses and irradiation behaviors of metallic fuel elements that are representative of the proposed VTR driver fuel design concepts. As such, many of the code requirements and physics/modeling discussions in this work are expected to be broadly applicable to metallic fuel applications outside of the VTR project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

TRISO Fuel for High-Temperature Passively-Safe Nuclear Reactors

The poster entitled "TRISO Fuel for High-Temperature Passively-Safe Nuclear Reactors" presents reasons why TRISO fuel is a safe nuclear fuel. The poster contains four blocks: Multiple barriers to fission product release, Thermal stability, Chemical compatibility and Safety and Performance. Each block present information on the TRISO fuel in terms of the mechanical, chemical and thermal properties. In summary the safety and performance of the TRISO fuel based on the experimental results is shown.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Safeguarding Reactors and Spent Nuclear Fuel [Slides]

Highly enriched fuel contains large quantities of ²³⁵U; Burned fuel may contain large quantities of ²³⁹Pu; Reprocessing is the major threat (discussed at length later), but we need to ensure the amount of plutonium going to the reprocessing plant is what we believe it to be; All nuclear material needs to be accounted for (material accountancy); There may be a loss of continuity of knowledge in which case fuel needs to be measured to determine how much plutonium and uranium it contains.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Impact of Irradiation-induced Lattice Defects and Dopants on Thermal Transport in Oxide Nuclear Fuels

Advanced nuclear reactor concepts aim to use fuels that must withstand unprecedented temperature and radiation extremes. In these fuels, thermal energy transport under irradiation is directly related to fuel longevity, reactor safety, and is arguably one of the most important performance metrics. Here we provide a comprehensive, first-principles-informed treatment of phonon mediated thermal transport in a defect-bearing actinide oxide with direct comparison to experimental measurements. Pristine and proton irradiated thorium dioxide was chosen as a model system to treat the complexity of thermal transport in the presence of lattice defects. A thermal transport model is implemented using the linearized Boltzmann transport equation (LBTE) with input from first principles calculations and defect evolution models. The output of the LBTE is compared directly to mesoscopic measurements of thermal conductivity on length scales commensurate with defect accumulation. Parametric measurements of conductivity with irradiation dose and temperature reveal a monotonic decrease in conductivity with irradiation dose, in the range of 0.001 dpa to 0.1 dpa. Besides radiation-induced defects, doping the thorium dioxide crystal with small quantities of uranium atoms also results in a drastic reduction in thermal conductivity. Temperature-dependent measurements of thermal conductivity in uranium-doped thorium dioxide single crystals suggests a resonant scattering mechanism is responsible for the observed reduction. This comprehensive, atomistic- to meso-scale treatment provides the necessary basis to investigate thermal transport under irradiation in more complex systems that exhibit strong electron correlation. This is a poster presentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗