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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Comparison of FAST Experiments to BISON Simulations

The Fission Accelerated Steady-State Test (FAST) experiments utilize the effects of scaling to accelerate the burnup experienced by the fuel. This experimental technique offers a unique opportunity to evaluate the effects of scaling on fuel and cladding material modelling. Currently, models used in the BISON fuel performance code for metallic fuel performance are calibrated for EBR-II conditions and geometries. Comparing BISON simulations to FAST experimental data will allow for the identification of shortcomings in the models for conditions outside the EBR-II paradigm. FAST experiments have been performed on U-Zr and U-10Mo fuel types with HT9 cladding. Initial comparisons of simulated cladding strains to experimental results show that the BISON simulations consistently overpredict cladding strain in the fuel region but underpredict the cladding strain in the plenum region. This indicates potential issues with the simulation of swelling and gas pressure at smaller geometries. Further comparison with fission gas release measurements is planned as data becomes available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2001

Alleviation of Facility/Engine Interactions in an Open-Jet Scramjet Test Facility

Results of a series of shakedown tests to eliminate facility/engine interactions in an open-jet scramjet test facility are presented. The tests were conducted with the NASA DFX (Dual-Fuel eXperimental scramjet) engine in the NASA Langley Combustion Heated Scramjet Test Facility (CHSTF) in support of the Hyper-X program, The majority of the tests were conducted at a total enthalpy and pressure corresponding to Mach 5 flight at a dynamic pressure of 734 psf. The DFX is the largest engine ever tested in the CHSTF. Blockage, in terms of the projected engine area relative to the nozzle exit area, is 81% with the engine forebody leading edge aligned with the upper edge of the facility nozzle such that it ingests the nozzle boundary layer. The blockage increases to 95% with the engine forebody leading edge positioned 2 in. down in the core flow. Previous engines successfully tested in the CHSTF have had blockages of no more than 51%. Oil flow studies along with facility and engine pressure measurements were used to define flow behavior. These results guided modifications to existing aeroappliances and the design of new aeroappliances. These changes allowed fueled tests to be conducted without facility interaction effects in the data with the engine forebody leading edge positioned to ingest the facility nozzle boundary layer. Interaction effects were also reduced for tests with the engine forebody leading edge positioned 2 in. into the core flow, however some interaction effects were still evident in the engine data. A new shroud and diffuser have been designed with the goal of allowing fueled tests to be conducted with the engine forebody leading edge positioned in the core without facility interaction effects in the data. Evaluation tests of the new shroud and diffuser will be conducted once ongoing fueled engine tests have been completed.

Albertson, Cindy W.↗
DOE OSTI2026

FGMS-poster

Idaho National Laboratory (INL) performs post irradiation examination (PIE) of tri-structural isotropic (TRISO)-coated particle fuel to help qualify it for high temperature gas cooled reactors. TRISO fuel compacts are re-irradiated in the Neutron Radiography Reactor (NRAD) to generate the short lived fission products needed for fission product release testing. The Fuel Accident Condition Simulator (FACS) furnace and newly added Screen Neutron Irradiated Fuel for Failure (SNIFF) furnace heat the compacts in helium to temperatures of up to 2,000°C, prompting fission product release—predominantly gaseous xenon and krypton isotopes and condensable products such as cesium—from failed particles. These released isotopes are transported to a fission gas monitoring system (FGMS 1 or FGMS 3), where they accumulate in cryogenic cold traps and are quantified using high-purity germanium (HPGe) detectors. The addition of SNIFF and FGMS 3 increases throughput by enabling simultaneous testing of multiple compacts. Furthermore, automated INL developed software provides continuous, near-real time monitoring of fission product inventories and manages the liquid nitrogen cooling of the traps. These system enhancements improve the efficiency, data quality, and testing capacity of TRISO fuel performance evaluations.

07 - ISOTOPES AND RADIATION SOURCES↗
DOE OSTI2018

Postirradiation examination on metallic fuel in the AFC-2 irradiation test series

In this paper, irradiation experiments on metallic fuel alloys containing rare earth additions to simulate potential carry-over from pyro-metallurgical reprocessing and minor actinide additions were performed in the Idaho National Laboratory Advanced Test Reactor. These tests were designated AFC-2A and AFC-2B. Additionally irradiation experiments were also performed on metallic fuel alloys containing minor actinide additions that were fabricated by different techniques. These tests were designated AFC-2E. These pins were removed from the reactor after achieving burnups between 6 and 11% fission per initial heave metal atom. Examination of these pins revealed that they were unintentionally operated beyond their desired power and temperature conditions. Highlights from the postirradiation examination of these experiments is discussed. A variety of issue contributed to the poor performance of these irradiation tests including capsule design, reactor power variations and fabrication issues. While all the issues contributed, the major issue was determined to be unanticipated variations in reactor power. Through comparisons between these experiments and historical off-normal experiments, it was possible to conclude bound the operating temperatures of these experiments. There likely was some in-pile melting in the rare earth addition experiments (AFC-2A, AFC-2B), but it is not likely that there was in-pile melting in the fabrication technique experiments (AFC-2E).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Summary of Thermocouple Performance in the Advanced Gas Reactor Experiment AGR-5/6/7 During Irradiation in the Advanced Test Reactor

The Advanced Gas Reactor -5/6/7 (AGR-5/6/7) experiment was the fourth and final experiment in the AGR experiment series and completed irradiation in July 2020. It serves as the formal fuel qualification test for the TRISO fuels under development by the US Department of Energy. This experiment was designed to irradiate fuel particles at temperatures ranging from 800°C ? 1500°C. The high end of the range created unique challenges for thermocouple-based temperature measurements. Commercially available high-temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C) suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. A special low neutron cross-section thermocouple system based on molybdenum/niobium thermoelements called High Temperature Irradiation Resistant thermocouple (HTIR-TC) has been under development at INL since circa 2004. Several examples of this thermocouple type were incorporated into regions of the test operating above 1200°C. For regions in the test operating in the 1000°C ? 1200°C range, high-performance versions of Type N thermocouples recently developed at Cambridge University were installed. Standard Type N thermocouples were used in regions of the test operating below 1000°C. A total of 54 thermocouples were incorporated into the test. A report of the performance of this large heterogenous set of thermocouples over the first third of the irradiation was previously provided (ANIMMA 2019). This follow-on paper discusses results from the balance of the experiment (Feb 2019 - July 2020). Special attention is focused on potential drift of thermocouples operating in the higher temperature regions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Application of Fuel Depletion Chain Simplification to Experiment Analysis in the Advanced Test Reactor

An irradiation experiment analysis can be informed by high-fidelity reactor engineering depletion results, but this comes at a computational cost. Applying depletion chain simplification to the advanced test reactor driver fuel before performing experiment depletions permits their programmatic parameters to be calculated faster, with a small penalty to accuracy. Here, this work contrasts the results of two irradiation experiments with different neutronic characteristics. Overall, the simplified nuclide library produced using a simple one-group microscopic cross-section library for a pressurized water reactor in the depletion chain simplification process performed comparably in terms of accuracy and runtime to the simplified nuclide library produced using a three-group microscopic cross-section library generated specifically for the advanced test reactor experiments being modeled. This is attributed to the additional nuclides and transmutation pathways preserved in the one-group cross-section library, which has data for 297 nuclides, compared to the three-group cross-section library, which has data for 217 nuclides. This indicates that a cross-section library with more nuclides is better than a cross-section library with fewer nuclides for the depletion chain simplification process, even if the cross-section library with fewer nuclides better represents the flux spectrum of the system being considered.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2004

Solid oxidized fuel cells seals leakage setup and testing

As the world s reserves of fossil fuels are depleted, the U.S. Government, as well as other countries and private industries, is researching solutions for obtaining power, answers that would be more efficient and environmentally friendly. For a long time engineers have been trying to obtain the benefits of clean electric power without heavy batteries or pollution-producing engines. While some of the inventions proved to be effective (i.e. solar panels or windmills) their applications are limited due to dependency on the energy source (i.e. sun or wind). Currently, as energy concerns increase, research is being carried out on the development of a Solid Oxide Fuel Cell (SOFC). The United States government is taking a proactive role in expanding the technology through the Solid State Energy Conversion Alliance (SECA) Program, which is coordinated by the Department of Energy. into an electrical energy. This occurs by the means of natural tendency of oxygen and hydrogen to chemically react. While controlling the process, it is possible to harvest the energy given off by the reaction. SOFCs use currently available fossil fuels and convert a variety of those fuels with very high efficiency (about 40% more efficient than modem thermal power plants). At the same time they are almost entirely nonpolluting and due to their size they can be placed in remote areas. The main fields where the application of the fuel cells appears to be the most useful for are stationary energy sources, transportation, and military applications. structure and materials must be resolved. All the components must be operational in harsh environments including temperatures reaching 800 C and cyclic thermal- mechanical loading. Under these conditions, the main concern is the requirement for hermetic seals to: (1) prevent mixing of the fuel and oxidant within the stack, (2) prevent parasitic leakage of the fuel from the stack, (3) prevent contamination of the anode by air leaking into the stack, (4) electrically isolate the individual cells within the stack, and (5) mechanically bond the cell components. The sealing challenges are aggravated by the need to maintain hermetic boundaries between the different flow paths within the fuel cell throughout cycled operation. Within the timeframe of my tenure, the main objective is to assist in building a state-of-art test facility.

Bastrzyk, Marta B.↗
DOE OSTI2025

Preliminary Assessment of Alternate Chlorination Process

Recent advancements towards low-temperature chlorination with disulfur dichloride and thionyl chloride of Al-clad used nuclear fuel (UNF) have been summarized for work control technical expertise development at the Savannah River National Laboratory (SRNL). This technique aims to provide a more effective method for separating U from cladding or alloying elements. A preliminary assessment determined needs for unit operation developments in regard to reaction kinetics and reagent quality. Furthermore, an analysis of an Uruguay fuel plate (U-Al x fuel meat), with low-enriched uranium (LEU) and located at SRNL, was a potential candidate for future experiments with lightly irradiated (0.08% burnup) fuel to test for chlorination. Future proposed work would involve kinetic studies, reagent quality assessments, and the machining of the fuel plate down to an appropriate bench-scale size and testing the effectiveness of low-temperature chlorination on the fuel plate.

Chlorination↗
DOE OSTI2017

Fuel Motion Monitoring System at TREAT – FY2017 Instrumentation Summary

This document is a report summarizing FY2017 efforts related to the Transient Reactor Test Facility (TREAT) Fuel Motion Monitoring System, "hodoscope," refurbishment campaign. TREAT is an air-cooled, graphite-moderated, heterogeneous test facility designed to evaluate reactor fuels and structural materials under conditions that simulate various types of transient overpower and under-cooling situations in a nuclear reactor. Fuel meltdowns, metal-water reactions, thermal interaction between overheated fuel and coolant, and the transient behavior of ceramic fuel for high-temperature systems can be studied. A key instrument that monitors fuel motion as these events take place is the fast neutron hodoscope. The hodoscope is designed to allow for pre-, during-, and post- transient imaging of fuel within a test loop in the center of the reactor core. The preferred technique for imaging is detection of fast neutrons produced by fission in the test fuel, travelling unmoderated through an empty row of core fuel assemblies and then through a multi-slot collimator to a 360-detector array. Due to the age of the instrument, a refurbishment of the hodoscope’s fast neutron detection capability is necessary to support real-time, fuel-position experimental data collection. One particular challenge in this regard is related to the refurbishment of the fast-neutron-sensing detector elements in the detector array. This report summarizes major activities and accomplishments related to the FMMS refurbishment in FY2017. Of particular note was work related to identifying options for new proton recoil scintillator wafer and initial examination of the TREAT proton recoil proportional counter detectors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS2012

Peak Seeking Control for Reduced Fuel Consumption with Preliminary Flight Test Results

The Environmentally Responsible Aviation project seeks to accomplish the simultaneous reduction of fuel burn, noise, and emissions. A project at NASA Dryden Flight Research Center is contributing to ERAs goals by exploring the practical application of real-time trim configuration optimization for enhanced performance and reduced fuel consumption. This peak-seeking control approach is based on Newton-Raphson algorithm using a time-varying Kalman filter to estimate the gradient of the performance function. In real-time operation, deflection of symmetric ailerons, trailing-edge flaps, and leading-edge flaps of a modified F-18 are directly optimized, and the horizontal stabilators and angle of attack are indirectly optimized. Preliminary results from three research flights are presented herein. The optimization system found a trim configuration that required approximately 3.5% less fuel flow than the baseline trim at the given flight condition. The algorithm consistently rediscovered the solution from several initial conditions. These preliminary results show the algorithm has good performance and is expected to show similar results at other flight conditions and aircraft configurations.

Brown, Nelson↗
DOE OSTI2016

Effect of Reactor Radiation on the Thermal Conductivity of Graphite-based Dispersion Fuel

The Fukushima Daiichi nuclear accident in 2011 stimulated materials development efforts to improve the current fuel design for Light Water Reactors (LWRs). From fuel element to fuel cladding, new concepts in advanced nuclear fuel development have been proposed, studied, and tested. The new fuel systems are expected to have better resistance to damage, melting, and chemical reaction during accident scenarios. To test the performance and accident tolerance of the advanced fuel, it is highly preferable to simulate off-normal transients in addition to standard in-core fuel performance tests (available at High Flux Isotope Reactor at Oak Ridge National Laboratory and Advanced Test Reactor at Idaho National Laboratory). The Transient Reactor Test Facility (TREAT) is now being prepared to restart its operation, performing transient testing to assist advanced fuel development. TREAT was operated from 1959 to 1994, and was extensively used to generate thousands of reactor transients for different programs. The reactor has not operated since 1994. The original fuel used at TREAT was graphite-based dispersion fuel with highly-enriched uranium (HEU, 93.1% enriched uranium). For nonproliferation purposes, a fuel conversion program, supported by U.S. Department of Energy, National Nuclear Safety Administration (NNSA), Office of Material Management and Minimization (NA-23) Reactor Conversion Program, is underway to develop a replacement core in which the UO{sub 2} contains only low-enriched uranium (LEU). The preliminary LEU core design has shown a much lower carbon/uranium atomic ratio and enrichment than in the HEU core design. To achieve performance similar to the HEU core, the prospective LEU core will contain a much higher density of UO{sub 2} particles. As a result, a higher level of radiation damage, particularly from fission fragments emerging from the UO{sub 2} particles, can be expected dependent on the UO{sub 2} particle size within the fuel. To support LEU fuel design, the present study investigates the effect of reactor radiations on the thermal conductivity of graphite-based dispersion fuel. Both neutron- and fission-radiation induced degradations of the fuel are modeled, and validated based on the available experimental data. In this study, the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel was analyzed. A semi-empirical method to model the relative thermal conductivity with reactor radiations was proposed and validated. The proposed method can be further applied to predict the changes in the thermal conductivity of the prospective LEU TREAT fuel during long-term operation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Initial Testing of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application

Present and new nuclear fuels that will be tested at the Transient Reactor Test (TREAT) facility will be analyzed with the MAMMOTH reactor physics application, currently under development, at Idaho National Laboratory. MAMMOTH natively couples the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. This system allows the irradiation of fuel from beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing within the same analysis mesh and, thus, retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. The Chebyshev Rational Approximation Method (CRAM) is the current state-of-the-art in the field, as was chosen to be the solver for the decay and transmutation equations. This report shows that the implementation of the CRAM solver within MAMMOTH is correct with various analytic benchmarks for decay and transmutation of nuclides. The results indicate that the solutions with CRAM order 16 achieve the level of precision of the benchmark. The CRAM solutions show little sensitivity to the time step size and consistently produce a high level of accuracy for isotopic decay for time steps of 1x10^11 years. Comparisons to DRAGON5 with 297 isotopes yield comparable results, but some differences need to be further analyzed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Specifications of Legacy U(Pu)Zr Metallography Data

The DOE Nuclear Energy Advanced Reactor Technologies (ART) Program has supported the creation of several databases with information describing the safety performance of fast reactors, components, and fuels. This growing collection of legacy experimental data, operating data, and analysis is available online to registered users. Metallography data is one of the most important types of PIE data being collected, organized and stored in several ART Fast Reactor Databases (https://frdb.ne.anl.gov), including the Fuels Irradiation & Physics Database (FIPD), Out-of-Pile Transient Database (OPTD), and TREAT (the Transient Reactor Test Facility) Experimental Relational Database (TREXR). These databases contain three main sets of metallography data. The first set is the metallography data from Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) irradiated fuel pins measured in the Hot Fuel Examination Facility (HFEF); the second set is the metallography data from EBR-II irradiated fuel pins measured in the Alpha-Gamma Hot Cell Facility (AGHCF); the third set is the metallography data from transient tested fuel pins (the transients tests include the out-of-pile tests and TREAT tests) measured in AGHCF. Since both the second and third sets of data were measured in AGHCF, they are governed by the same specification. The metallography data in the databases are digital images scanned from either positive or negative photos. The quality of the images, including their resolution and contrast, relies on the preserved quality of the pictures and the scanning conditions. To analyze the microstructure of a fuel pin, a series of preparatory steps must be undertaken. These include sectioning, epoxy mounting, mechanical grinding and polishing, and often etching. The resulting samples were then transferred to a secondary hot cell, or glovebox (depending the strength of radiation field) for microscopy examination. The specifications provided herein focus on the metallography examinations. The sample grinding, polishing and etching processes are also discussed. The procedures involving sectioning and epoxy mounting are out of the scope of the current specification; details can be found in the corresponding operation manuals. If more data are collected and added to the ART Fast Reactor Databases, this specification will be updated to accommodate the additional data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Synthesis of Butyl-Exchanged Polyoxymethylene Ethers as Renewable Diesel Blendstocks with Improved Fuel Properties

Methyl-terminated polyoxymethylene ethers (MM-POMEs), having the formula CH3O–(CH2O)n–CH3 (n = 3–5), are a high-cetane, low-sooting group of oxygenates that have recently attracted attention as potential diesel blendstocks. Despite these attractive fuel properties, MM-POMEs have shortcomings due to their low energy density and high water solubility. Guided by a computational fuel property assessment for POMEs with longer end-groups, the most promising improvements in the desired compression ignition fuel properties were observed for butyl-terminated POMEs. Here, an acid-catalyzed transacetalization reaction was developed to exchange the methyl end-groups of MM-POMEs (n = 3–6) with butyl end-groups. The reaction utilizes an ion-exchange resin as the acid catalyst at mild reaction conditions of 60 °C and atmospheric pressure. Approximately 100 mL of butyl-exchanged POMEs in the diesel boiling range were produced, enabling laboratory-scale fuel property testing. The butyl-terminated POME mixture possesses the advantaged fuel properties of the parent MM-POMEs (low-soot, high-cetane) while exhibiting improved energy density (lower heating value (LHV) of 30 MJ/kg) and substantially reduced water solubility (7.3 g/L) compared to the parent MM-POME mixture (LHV of 19 MJ/kg , water solubility of 258 g/L).

09 BIOMASS FUELS↗
DOE OSTI2018

Studies of Lanthanide Transport in Metallic Fuel

Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Creating and Preserving Nanoparticles during Co-Sintering of Solid Oxide Electrodes and Its Impact on Electrocatalytic Activity

A novel processing method that creates and preserves ceramic nanoparticles in solid oxide electrodes during co-sintering at traditional sintering temperatures is introduced. Specifically, carbon templated samarium-doped ceria nanoparticles (nSDC) were successfully integrated with commercial lanthanum strontium cobalt ferrite (LSCF) and commercial SDC powders, producing LSCF-SDC-nSDC cathodes upon processing. The effect of nSDC concentration on cathode electrocatalytic activity was investigated at low operational temperatures, 600 °C–700 °C, with symmetrical cells. Low nSDC loadings, ≤5 wt% nSDC, significantly decreased cell polarization resistance whereas higher loadings increased it. The best electrochemical performance was achieved with 5 wt% nSDC, lowering the polarization resistance by 41% at 600 °C. Fuel cell tests demonstrate that adding 5 wt% nSDC increased the maximum fuel cell power density by 38%. Electrochemical impedance spectra showed substantial improvements in both fuel cell polarization resistance and ohmic resistance, indicating that nSDC increased the electrocatalytically active area of the cathode. This work demonstrates a simple, novel method for effectively increasing electrocatalytic activity of solid oxide electrodes at low operational temperatures.

Muhoza, Sixbert P. (ORCID:0000000299864241)↗
DOE OSTI2019

Separate Effects Testing in TREAT for ATF Fuels

This work performed examinations on high burnup UO2 from commercial power plant with the intention of better understanding fission gas and fragmentation behavior. Fuel fragmentation, relocation, and dispersal (FFRD) is a major factor in fuel lifetime limits imposed on UO2 by regulators, both domestic and foreign, and is a major area of interest by industry groups (e.g., Electric Power Research Institute, [EPRI]). The characterization of the fuel focused on the differences in porosity between the high burnup (rim) zone, the dark (intermediate) zone, and center zones of the fuel as well as the fission gas content of each region. Work involved thermal treatments to analyze fission gas and high resolution microscopy techniques to investigate pore structure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗