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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 343 records · Page 19

Applications of plasma-enhanced metalorganic chemical vapor deposition

The use of plasma enhancement for growth of III-V compound semiconductor materials by metalorganic chemical vapor deposition (MOCVD) is examined, to improve control of microstructure, develop understanding of the underlying growth mechanisms, and expand the range of materials combinations for photovoltaic and solid-state lighting applications. Whereas plasma is commonly used with group-IV materials, such as silicon, few studies have examined the impact of plasma-enhanced MOCVD (PE-MOCVD) for III-V materials growth. PE-MOCVD provides improved decomposition of metalorganic precursors, which is driven by hydrogen plasma to augment hydride reactions and thermal pyrolysis, but plasma generation requires low reactor pressure. We have demonstrated elemental Al films grown by PE-MOCVD, which show distinct crystallographic texturing, and assume epitaxial microstructure upon post-growth annealing. We also demonstrate PE-MOCVD of GaAs at temperatures as low as 300 °C. With increasing radio-frequency power, the GaAs growth rate shows a transition from power limited to mass-transport limited. PE-MOCVD grown Ga 0.5 In 0.5 P shows no detectable spontaneous atomic ordering, which offers a mechanism to form order/disorder unicompositional heterostructures, without temperature change or impurity incorporation. In its current implementation, PE-MOCVD grown films have shown microstructural degradation, that we attribute to direct plasma exposure during growth. Improvements in the system design and growth sequences are at the focus of on-going efforts.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

EVALUATION OF HOT CHANNEL FACTOR FOR SODIUM-COOLED FAST REACTORS WITH MULTI-PHYSICS TOOLKIT

The evaluation of hot channel factor (HCF) is of great significance to the quantification of safety margins for reactor designs. In this paper, HCFs for a sodium-cooled fast reactor (SFR) are evaluated with the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit, which is developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of DOE for multi-physics reactor performance and safety simulations. The high-fidelity neutronics and thermal hydraulics solvers PROTEUS and Nek5000 in the SHARP toolkit are coupled to perform the multi-physics simulations for HCF evaluation. The HCFs induced by cladding manufacturing tolerance, fissile content mal-distribution, wire orientation and uncertainties on the cladding, coolant, and fuel properties are evaluated for a reference core SFR design (AFR-100). The HCFs calculated with the SHARP toolkit are compared to legacy HCFs for similar reactor types. The comparison demonstrates the reduction or elimination of modeling uncertainties in the calculation of HCFs using high fidelity advanced modeling and simulation tools without the need of expensive experiments. Moreover, the reduction of the uncertainties on HCFs evaluation allows an increase in nominal parameters and safety margin, which in turn improves the economic competitiveness of the SFR.

Hot channel factor (HCF)↗

Pu(IV) quantification via visible–near-infrared absorption spectroscopy: tackling interferences using D-optimal design and partial least squares

Here, this study presents a novel analytical approach for quantifying Pu(IV) in glove box environments using fiber-optic-based visible–near-infrared absorption spectroscopy in combination with partial least squares regression (PLSR) and design of experiments. The method addresses significant challenges posed by overlapping spectral features arising from Nd(III), which is a common fission product impurity, and the speciation variability of Pu(IV) nitrato complexes in HNO 3 concentrations ranging from 2.5 to 11 M. A curated training set consisting of data from 20 samples was developed via D-optimal design to enable robust PLSR model calibration for Pu(IV) using the near-infrared band near 1050 nm. The training set was acquired from samples in cuvettes with a 1-cm path length and was used to build the PLSR model. The robustness of the model was validated with data collected using a dip probe with a 1-cm path length and varying Pu(IV) concentrations. The strong performance of the model indicates good model transfer from cuvette to dip probe and highlights the potential for in situ measurements and online monitoring of reactions in a crystallization reactor vessel. The results demonstrate that this combined spectroscopic and chemometric approach can accurately and simultaneously quantify Pu(IV) and HNO 3 , thereby offering a promising tool for real-time monitoring in process environments.

Actinide↗

An Integrated Flame Spray Process for Low Cost Production of Battery Materials for Lithium Ion Batteries and Beyond

The overall objective of this project was to develop an advanced manufacturing technology for Li-ion battery materials production at low cost and in a green chemical process using glycerol as solvent to replace water. It was aimed at developing a prototype pilot production line to reduce production cost by at least 25% to the baseline. During the course of this project, research was conducted to understand spray drying and combustion of eutectic solvent precursors of glycerol and metal salts (acetates) in the formation of mixed metal oxide powders. Reactor design and optimization were done to facilitate spray drying and combustion in powder formation and processing. Various cathode materials powders were produced, including lithium nickel-cobalt-manganese oxide (NCM) and lithium nickel-cobalt-aluminum oxide (NCA). These cathode materials powders were tested in half coin cells for their performances, with materials performances of 160 mAh/g for NCM111 and 200 mAh/g for NCA. NCA powders were also produced using a slurry spray-drying process and tested in full pouch cells. A pilot scale flame assisted spray drying reactor was constructed and demonstrated for its production rate exceeding the target of production rate of four metric tons per year. The work performed in this project has been documented in 9 peer reviewed journal publications, 8 conference presentations, and two IP disclosures/patent applications.

25 ENERGY STORAGE↗

Characterizing the Performance of Fission Chambers for Local Neutron Flux and Spectrum Measurement

Three fission chambers were tested at The Ohio State University Research Reactor (OSURR) to evaluate fission chambers for neutron flux and spectrum measurements. Two of the fission chambers are commercial Photonis CFUR43 fission chambers—one utilizing highly enriched uranium-235 fissile deposits for thermal neutron detection and the other with highly depleted uranium-238 for fast neutron detection. This experiment was also performed in collaboration with French Alternative Energies and Atomic Energy Commission (CEA) with CEA suppling a uranium-235 loaded fission chamber built from the Photonis CFPR CE8/9 kit. The fission chambers were irradiated in a movable 6.5 inch dry tube at reactor powers of 50 kW, 200 kW, and 450 kW at ambient temperature and at 350°C. While all sensors demonstrated good linear sensitivity to reactor power, a leakage current of 6 nA was measured for the CFUR43 detectors. Additionally, the measured signal ratios of the two CFUR43 detectors did not match the theoretical values and are subject to further investigation on individual signal contributors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

ECAR-6564 MARVEL Project Primary Coolant System ASME BPVC Section III Division 5 Design by Analysis

This report demonstrates that the MARVEL Reactor Primary Coolant Boundary (herein referred to as the Primary Coolant System, or PCS) is designed to meet ASME BPVC Section III Division 5 elevated temperature service design-by-analysis criteria. The analysis approach that delineates division of responsibilities to meet project objectives is discussed herein. In short, Design and Service Level A, B, and C code calculations are detailed in ECAR-6580 for the majority of the PCS with complex geometry, while the Lower Downcomers, Bottom Head, and Reactor Core Barrel are analyzed in ANL-23/56. Service Level D code calculations are detailed in this document.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Droplet Entrainment in Steam Supply System of Water-Cooled Small Modular Reactors: Experiment and Modeling Approaches

Droplet entrainment in steam-flow is a prominent phenomenon that needs adequate safety and risk analysis of postulated transient and accident scenarios—including experimental investigation and representative modeling and simulation (M&S)—for small modular reactor (SMR) system design and demonstration. This study identifies knowledge gaps by evaluating experimental and computational fluid dynamics modeling approaches to support early-stage reactor system design, testing, and model evaluation. Previous studies reported in the literature for steam-flow entrainment primarily focused on gigawatt capacity pressurized water reactor (PWR) systems. However, entrainment phenomena are even more prominent for PWR-type SMRs due to their more compact integrated designs, which need further research and development. To fill the research gaps, this study provides insight by specifying the phenomena of interest by leveraging the lessons learned from past research, adopting advanced M&S techniques and advanced instrumentation and control. The findings and recommendations are applicable for evaluating steam-flow entrainment models and for designing integral effect test and separate effect test facilities for gaining reactor design approvals.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Optimal control of thermally coupled Navier Stokes equations

The optimal boundary temperature control of the stationary thermally coupled incompressible Navier-Stokes equation is considered. Well-posedness and existence of the optimal control and a necessary optimality condition are obtained. Optimization algorithms based on the augmented Lagrangian method with second order update are discussed. A test example motivated by control of transport process in the high pressure vapor transport (HVPT) reactor is presented to demonstrate the applicability of our theoretical results and proposed algorithm.

Ito, Kazufumi↗

Development Status of the Fission Power System Technology Demonstration Unit

This paper summarizes the progress that has been made in the development of the Fission Power System Technology Demonstration Unit (TDU). The reactor simulator core and Annular Linear Induction Pump have been fabricated and assembled into a test loop at the NASA Marshall Space Flight Center. A 12 kWe Power Conversion Unit (PCU) is being developed consisting of two 6 kWe free-piston Stirling engines. The two 6 kWe engines have been fabricated by Sunpower Inc. and are currently being tested separately prior to integration into the PCU. The Facility Cooling System (FCS) used to reject convertor waste heat has been assembled and tested at the NASA Glenn Research Center (GRC). The structural elements, including a Buildup Assembly Platform (BAP) and Upper Truss Structure (UTS) have been fabricated, and will be used to test cold-end components in thermal vacuum prior to TDU testing. Once all components have been fully tested at the subsystem level, they will be assembled into an end-to-end system and tested in thermal vacuum at GRC.

Briggs, Maxwell H.↗

Large-Scale Weibull Analysis of H-451 Nuclear- Grade Graphite Specimen Rupture Data

A Weibull analysis was performed of the strength distribution and size effects for 2000 specimens of H-451 nuclear-grade graphite. The data, generated elsewhere, measured the tensile and four-point-flexure room-temperature rupture strength of specimens excised from a single extruded graphite log. Strength variation was compared with specimen location, size, and orientation relative to the parent body. In our study, data were progressively and extensively pooled into larger data sets to discriminate overall trends from local variations and to investigate the strength distribution. The CARES/Life and WeibPar codes were used to investigate issues regarding the size effect, Weibull parameter consistency, and nonlinear stress-strain response. Overall, the Weibull distribution described the behavior of the pooled data very well. However, the issue regarding the smaller-than-expected size effect remained. This exercise illustrated that a conservative approach using a two-parameter Weibull distribution is best for designing graphite components with low probability of failure for the in-core structures in the proposed Generation IV (Gen IV) high-temperature gas-cooled nuclear reactors. This exercise also demonstrated the continuing need to better understand the mechanisms driving stochastic strength response. Extensive appendixes are provided with this report to show all aspects of the rupture data and analytical results.

Nemeth, Noel N.↗

Development Status of the Fission Power System Technology Demonstration Unit

This paper summarizes the progress that has been made in the development of the Fission Power System Technology Demonstration Unit (TDU). The reactor simulator core and Annular Linear Induction Pump have been fabricated and assembled into a test loop at the NASA Marshall Space Flight Center. A 12 kWe Power Conversion Unit (PCU) is being developed consisting of two 6 kWe free-piston Stirling engines. The two 6 kWe engines have been fabricated by Sunpower Inc. and are currently being tested separately prior to integration into the PCU. The Facility Cooling System (FCS) used to reject convertor waste heat has been assembled and tested at the NASA Glenn Research Center (GRC). The structural elements, including a Buildup Assembly Platform (BAP) and Upper Truss Structure (UTS) have been fabricated, and will be used to test cold-end components in thermal vacuum prior to TDU testing. Once all components have been fully tested at the subsystem level, they will be assembled into an end-to-end system and tested in thermal vacuum at NASA GRC.

Briggs, Maxwell H.↗

Status of ISS Water Management and Recovery

Water management on ISS is responsible for the provision of water to the crew for drinking water, food preparation, and hygiene, to the Oxygen Generation System (OGS) for oxygen production via electrolysis, to the Waste & Hygiene Compartment (WHC) for flush water, and for experiments on ISS. This paper summarizes water management activities on the ISS US Segment as of March 2022 and provides a status of the performance and issues related to the operation of the Water Processor Assembly (WPA) and Urine Processor Assembly (UPA).

Water Recovery System↗

Oklo and BWXT Short Decks

Two abstracts, one for each deck: Oklo is planning on producing fuel using spent EBR-II material and demonstrating an NRC licensed reactor. This presentation gives a high level overview of Oklo's progress in this work. The BWXT Advanced Nuclear Reactor Experiment 1 (BANR-1) is an irradiation experiment at INL's Advanced Test Reactor (ATR) that will irradiate novel Tristructural Isotropic (TRISO) fuel for potential future use by BWXT. This presentation gives a program overview and progress.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

ECAR-6564 Rev 1 MARVEL Project Primary Coolant System ASME BPVC Section III Division 5 Design by Analysis

This report demonstrates that the MARVEL Reactor Primary Coolant Boundary (herein referred to as the Primary Coolant System, or PCS) is designed to meet ASME BPVC Section III Division 5 elevated temperature service design-by-analysis criteria. The analysis approach that delineates division of responsibilities to meet project objectives is discussed herein. In short, Design and Service Level A, B, and C code calculations are detailed in ECAR-6580 [9] for the majority of the PCS with complex geometry, while the Lower Downcomers, Bottom Head, and Reactor Core Barrel are analyzed in ANL-24/36. Service Level D code calculations are detailed in this document.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

Thermoelectric power generation in the core of a nuclear reactor

Thermoelectric energy converters offer a promising solution to generate electrical power using heat in the nuclear reactor core. Despite significant improvements in thermoelectric efficiency of nanostructured materials, the performance of these advanced materials has yet to be demonstrated in the harsh radiation environment of a reactor core. Herein, we demonstrate a thermoelectric generator (TEG) made from nanostructured bulk half-Heusler (HH) materials generating stable electrical power density > 1140 W/m 2 after 30 days in the MIT Nuclear Research Reactor under an unprecedented fast-neutron (>1 MeV) fluence of 1.5 × 10 20 n/cm 2 . Despite an initial degradation due to irradiation damage when operating under relatively low temperatures, our TEG showed a 20-fold increase in power output when operating under high temperature due to in-situ annealing and resulting thermoelectric property recovery. First-principles modeling indicates that a chemically disordered metallic phase was formed under irradiation at lower temperatures, resulting in a drastic degradation in thermoelectric properties, while at sufficiently high temperatures the system returned to the initial chemically ordered HH phase and the thermoelectric properties recovered. Transmission electron microscopy and electron diffraction demonstrated that the chemically disordered phase was formed upon ion irradiation, confirming the prediction from first-principles simulations. Furthermore, the results suggest that with proper control over the TEG operating temperatures, the nanostructured bulk TEGs could produce stable electrical power and operate indefinitely in the core of a nuclear reactor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Integrated Security-Safety Consequence Study of a Heat Pipe Reactor

This report presents a preliminary demonstration of integrated security/safety consequence modeling to support potential physical security exemption justifications under the new Title 10 Code of Federal Regulations Part 53 licensing pathway, which allows exemptions if sabotage-induced radiological consequences do not exceed 25 rem at the site boundary two hours after release without mitigative actions. A hypothetical heat pipe microreactor subjected to direct sabotage by a 150 lbs. TNT-equivalent high-explosive device was analyzed using a simplified radionuclide release estimate based on DOE-HDBK-3010-94, implemented in the MELCOR severe accident analysis code to model release transport and attenuation through reactor/building compartments, with MELCOR outputs coupled to MACCS to compute offsite doses and the distance-to-threshold metric D 25rem . Six scenarios were evaluated to examine sensitivities to release pathway/building configuration and radionuclide form (vapor versus aerosols with different size distributions). Results show a positive correlation between leak path factor and D 25rem and indicate that reactor building compartmentalization can substantially reduce consequences (a single-story configuration reduced D 25rem by roughly an order of magnitude relative to a cavity-only configuration), while vapor versus aerosol assumptions had limited impact for the modeled two-story case on the spatial grid used.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Selective electrified polyethylene upcycling by pore-modulated pyrolysis

Plastic waste is a increasing problem, accumulating in landfills and the environment. Pyrolysis is a promising and industrially relevant approach for transforming plastic waste into value-added chemicals. However, the selectivity and yield of traditional plastic pyrolysis are poor, with products featuring broad molar mass distributions. Here we report a highly selective, energy-efficient and catalyst-free pyrolysis method that can upcycle plastic into value-added chemicals via pore-modulated pyrolysis. Using a Joule-heated carbon column, we demonstrate the pivotal role of the reactor’s graded porous structure in decreasing the polydispersity of the reaction intermediates, enabling high product selectivity and yield. The decreasing pore size of the reactor modulates the mass transport in an apparent gating effect—preventing high-molar-mass species from exiting the reactor before sufficient pyrolysis has occurred. Using polyethylene as a model reactant, we demonstrate a high yield of 65.9 ± 5.2% and up to 80.8% selectivity toward value-added aviation fuel precursor (C8–C18 hydrocarbons) without the use of any catalysts.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Results of the KRUSTY Nuclear System Test

The Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was a prototypic nuclear-powered test of a 5-kW(thermal) Kilopower space reactor. This paper presents results from the KRUSTY nuclear system test, which operated the power system at various temperatures and power levels for 28 consecutive hours. The testing showed that the system operated as expected and that the reactor is highly tolerant of possible failure conditions and transients. The key feature demonstrated was the ability of the reactor to load-follow the demand of the power conversion system. The thermal power of the test ranged from 1.5 to 5.0 kW(thermal), with a fuel temperature up to 880°C. Each 80-W(electric)–rated Stirling converter produced ~90 W(electric) at a component efficiency of ~35% and an overall system efficiency of ~25%.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗