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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 19 records
All sourcesNASA NTRSDOE OSTI
NASA NTRS1983

Investigation of sources, properties and preparation of distillate test fuels

Distillate test fuel blends were generated for prescribed variations in composition and physical properties. Fuels covering a wide range in properties and composition which would provide a matrix of fuels for possible use in future combustion research programs were identified. Except for tetralin the blending components were all from typical refinery streams. Property variation blends span a boiling range within 150 C to 335 C, freezing point -23 C to -43 C, aromatic content 20 to 50 volume percent, hydrogen content 11.8 to 14.2 mass percent, viscosity 4 and 11 cSt (-20 C), and naphthalenes 8 and 16 volume percent. Composition variation blends were made with two base stocks, one paraffinic and the other napthenic. To each base stock was added each of three aromatic type fuels (alkyl benzenes, tetralin, and naphthalenes) for assigned initial boiling point, final boiling point, and hydrogen content. The hydrogen content was 13.5 mass percent for the paraffinic base stock blends and 12.5 mass percent and 11.5 mass percent for the naphthenic base stock blends. Sample 5-gallon quantities of all blends were prepared and analyzed.

Bowden, J. N.↗
DOE OSTI2020

Performance of Vehicle Fuel System Elastomers and Plastics with Test Fuels Representing Gasoline Blended with 10% Ethanol (E10) and 16% Isobutanol (iBu16)

The compatibilities of fuel system elastomers and plastics were evaluated for test fuels containing 16 vol.% isobutanol (iBu16) and 10 vol.% ethanol (E10). Elastomers included two fluorocarbons, four acrylonitrile butadiene rubbers (NBRs), and one type of fluorosilicone, neoprene, and epichlorohydrin/ethylene oxide. Plastic materials included four nylon grades, three polyamides, polytetrafluoroethylene (PTFE), polyvinylidene fluoride (PVDF), ethylene tetrafluoroethylene (ETFE), polyphenylene sulfide (PPS), high-density polyethylene (HDPE), polybutylene terephthalate (PBT), polyoxymethylene (POM), flexible polyvinylchloride (PVC), polyetherimide (PEI), polyetheretherketone (PEEK), and a phenol formaldehyde reinforced with glass fiber (GFPF). For each polymer material, the volume, mass, and hardness were measured before and after drying. Dynamic mechanical analysis (DMA) measurements were also performed on the dried specimens.For the elastomer materials the measured properties were similar for both fuels. The fluorocarbons and fluorosilicone swelled the least (~20%), while more moderate (20-45%) expansion occurred for the two NBR hose grades and (ECO). HNBR, neoprene, and silicone exhibited high swelling and softening, which likely precludes their use in many fuel systems. For the plastic materials, the observed swell was low; Nylon 11 swelled around 15%, but otherwise, their measured swell was <10%. Many of the plastics also showed sensitivity to alcohol type, as the E10 test fuel often imparted appreciably higher swell than iBu16. In general, the plastic materials showed good compatibility with the iBu16 and E10 test fuels. Here, the sole exception was the PVC material, which was structurally degraded from exposure to either fuel type. Compositional analysis showed high fuel retention in Nylon 12 and PVC. PVC also experienced a significant reduction in plasticizer compounds following exposure, which resulted in embrittlement and an increase in the glass-to-rubber transition temperature.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2022

Dynamic System Scaling Application to Accelerated Nuclear Fuel Testing

The development of nuclear fuel and materials requires a continuous effort to investigate the acute and prolonged effects of irradiation, thermal-material stress, chemical change, or other conceivable damage mechanics acquired during normal operation or accident scenarios throughout its lifetime. As in-core fuel property measurement techniques advance to support in the real-time, non-invasive, and enhanced accuracy realm, it is the future of fuel development to pursue to higher degrees of control, predictability of integral test behavior via separate effects test (SET), and shorter test time intervals. The fuel development life cycle from initial concept to commercial licensing is approximated to be 20 years and current literature suggests by optimizing fuel performance codes with SETs, the process could possibly be compressed to 5 to 10 years. Recently, a research group in the Idaho National Laboratory (INL) is testing reduced scale fuel rods and increased power density to accelerate evolution of fuel phenomena in metallic fuels. In support of nuclear fuel rod development, compressing fuel test process, and accelerating fuel phenomena, it was the purpose of this study to investigate nuclear fuel performance phenomena via literature review and effectively scale the initial conditions, boundary conditions, and geometric properties to describe the time-dependent response including to fuel burnup, thermal and mechanical stress, transmutations and inter-diffusion, and other relevant observed phenomena. The study was based on BISON simulations of historic EBR-II metallic fuel experiments and Dynamical System Scaling (DSS) method are utilized to assess effects of scaling activity including fuel phenomena acceleration and calculations of time-dependent distortions. The research successfully scaled metallic fuel phenomena, accelerated fuel testing, and assessed the distortions for each scaled case derived.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Status of LEU U-10Mo Monolithic Fuel Testing and Qualification Efforts to Support US High Performance Research Reactor Conversions

The US High Performance Research Reactor Project has begun a new phase of fuel testing and post irradiation examination (PIE) of plate-type low-enriched uranium (LEU) U-10Mo Monolithic fuel. The testing campaigns are intended to first confirm that the fuel performance behavior of commercially produced fuel matches that of the lab-scale fabricated fuel tested during development, followed by qualification of the U-10Mo monolithic fuel system through the US Nuclear Regulatory Commission. Planned tests include miniature-plate, full-size plate, and fuel element irradiations, during which fuel test specimens fabricated with qualified processes will be irradiated under conditions that bound the NRC-regulated reactors in order to demonstrate that the fuel meets the overarching fuel performance requirements of 1) maintaining mechanical integrity 2) maintaining geometric stability and 3) exhibiting stable and predictable behavior. The MP-1 test, with the objective of confirming fuel performance behavior, is currently undergoing irradiation in the Advanced Test Reactor and PIE in INL hot cells. Status of upcoming qualification and licensing test design activities and highlights of initial as-fabricated characterization and testing results from the MP-1 test will be provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Computational and Experimental Benchmarking for Transient Fuel Testing

In 2018 the Transient Reactor Test (TREAT) Facility was restarted. This now-operational facility brings back world-class transient fuel testing capabilities to the U.S. Among a number of parallel efforts prior to the TREAT Facility’s restart a Department of Energy (DoE) Integrated Research Project (IRP) was funded for the purpose of benchmarking the TREAT with several goals in mind: To produce a comprehensive evaluation of existing TREAT Facility neutronics data using the next generation reactor core neutronics codes in accordance with established guidelines per the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP). To perform a benchmark using ready-use computational fluid dynamics tools against that of a historically collected sodium loop experiment collected in the TREAT. To design, develop and test a new experimental thermal hydraulic system of a representative TREAT Facility water flow loop. And benchmark the data collected against that produced via RELAP5-3D and TRACE. To develop a comprehensive instrumentation plan for the TREAT Facility by benchmarking performing of advanced in-pile instruments in the Massachusetts Institute of Technology Research Reactor (MITR) and the TREAT. The integrated team brought together collaborating members from three universities including Oregon State University (OSU), the University of Michigan (UM), and Massachusetts Institute of Technology (MIT); three national laboratories, Idaho National Laboratory, Argonne National Laboratory (ANL), and Oak Ridge National Laboratory (ORNL); and two industry partners, TerraPower (TP) and Harris Thermal Transfer Products (HTTP). A brief summary of accomplishments resulting from this IRP includes two filed patents, more than 40 conference papers, two published benchmark reports through IRPhEP, a new representative thermal hydraulic test facility alike the in-pile PWR water flow loop for TREAT and the first advanced instruments tested within the TREAT core upon its restart. The papers presented herein were developed and provide a summary of the significant technical accomplishments which align with the four project’s objectives. The collaborators has significant support from countless parties including the operational and technical staff at TREAT to ensure that their outcomes most impactfully aligned with program needs. Lastly the collaborators would like to acknowledge the fiscal support of the DoE’s Office of Nuclear Energy under work package identification number NU-15-OR-OSU- 0701-01. This report contains selected papers from the Computational and Experimental Benchmarking for Transient Fuel Testing Integrated Research Project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Accident tolerant fuel test 2c (ATF-2C) irradiation test report

A fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2C has completed its four cycles of prototypic steady state irradiation. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and burnup histories of each of the 18 fueled test pins has been calculated.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Concept Descriptions for the VTR Rabbit System and Driver Fuel Test Assemblies

Two of the experiment vehicles being developed for the Versatile Test Reactor (VTR) are presented here. The first is a rabbit system that will enable rapid insertion of small test capsules into the high fast flux of the VTR core for relatively short durations. The rabbit concept development includes the construction/demonstration of a near-full-scale system in a deep-water pool to demonstrate functionality, development of a concept of operations and initial procedures, and validation of thermal-hydraulic modeling. In addition, modeling efforts are underway to simulate the thermal and neutronic environment of a rabbit capsule. The second type of experiment vehicle presented here is a driver fuel test assembly for inserting fuel and materials tests into the core by replacing a driver fuel assembly. Here, a novel design for dismountable test assemblies is proposed for the VTR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Multi-modal tomographic imaging system for poolside characterization of nuclear test fuels: Design considerations and studies

Testing and qualification of advanced nuclear fuels involves an iterative process of prototyping, in-pile irradiation testing, and in-situ or ex-situ examination. Fuel restructuring and fission product migration during burnup are among the most important aspects of fuel evolution that affect several important performance characteristics such as heat removal, accident tolerance, and fission product retention. Pool-side non-destructive characterization techniques provide fuel developers with tools to understand fuel evolution at different points of burnup. A design for a compact, submersible, and multi-modal gamma-ray tomography system for imaging irradiated nuclear fuel is presented here. Detector selection, collimator geometry and fabrication, mechanical design, imaging protocol and acquisition speed are discussed. Modeling calculations show that sub-millimeter resolution can be achieved in both transmission computed tomography images as well as in emission computed tomography images in a matter of hours. Several design compromises and fabrication challenges are discussed which should be taken into consideration for future submersible gamma-ray tomography instruments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fast Reactor Fuel Testing Considerations

The slide deck is to be presented in a public (i.e., no content restrictions) panel session titled "Versatile Test Reactor: Current Developments" at the 2021 ANS Virtual Annual Meeting. The slides will be used for opening remarks and to present to the audience the technical considerations that motivate the experiment capabilities being incorporated into the Versatile Test Reactor. The line plot included on slide is copied directly from a journal publication, as cited. The values use to make the bar chart come from an open document (ANL-NSE-1) available on OSTI.gov, and are from the EBR-II Mark-VA U-Pu-Zr core design, which was never used. They will be described simply as an unused but illustrative EBR-II core design, and for the purpose of making the point that a range of irradiation and in-service conditions need to be addressed when testing and qualifying a new fuel design for a fast reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

FAST - A Unique Experiment and Simplified Approach to Accelerated Integral Fuel Testing

Idaho National Laboratory has a long and internationally recognized history of performing irradiation testing of advanced fast reactor fuels in the Experimental Breeder Reactor II (EBR-II) and the Advanced Test Reactor (ATR). Since EBR-II’s decommissioning in the early 1990’s, irradiation testing has been exclusively performed in ATR. The testing of fast reactor fuels in ATR is challenging given its thermal neutron spectrum and coolant temperatures far below those prototypic of fast reactors. Adept researchers have overcome these obstacles through the use of Cadmium shrouds which attenuate all thermalized neutrons and an experiment design that fosters fuel and cladding temperatures prototypic of fast neutron reactors. Although this design has successfully demonstrated temperatures and temperature gradients that closely mimic those seen in fast reactors, irradiation performance is highly sensitive to fabrication tolerances and the time requirements necessary for high fuel burnup, where irradiation performance data of fast fuel is lacking, are significant. This work summarizes two novel experiment designs and their associated physics analyses for testing fast reactor fuels in ATR; an accelerated irradiation experiment and a fast spectrum experiment. The accelerated irradiation experiment does not utilize a Cadmium shroud, has been shown to triple the burnup rate, yet yields temperatures and radial power profiles prototypic of fast reactor fuels. The fast spectrum experiment utilizes a booster fuel element that that hardens the incident neutron spectrum and also yields temperatures and radial power profiles prototypic of fast reactor fuels. These experiment designs efficiently reduce irradiation time requirements, are more economical, and significantly increase ATR throughput. Fast reactor experiments that capitalize on these designs expedite irradiation performance data for fast reactor fuels and provide information invaluable to fuel performance modeling and simulation validation over a wide range of parameters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Nuclide inventory characterization of EBR-II MOX fuel test pins

Characterization of 3 mixed oxide (MOX) fuel pins from the Experimental Breeder Reactor II (EBR-II) SPA-2/-2B experiment program was performed in support of the Advanced Reactor Experiments for Sodium Fast Reactor Fuels (ARES) joint project between Idaho National Laboratory (INL) and the Japan Atomic Energy Agency (JAEA). The characterization efforts were performed to enable transient testing in the Transient Reactor Test (TREAT) facility at INL and will include post irradiation examination (PIE) and other analyses. MOX fuel pins UW02009, 02011, and 06024 were historically irradiated from 1989 to 1994 for a total of 822.42 effective full power days (EFPDs) and peak burnup of approximately 14.3 at.% (atomic percent of heavy metal atoms, ∼134.4 GWd/t). They had since been placed in storage until a future use was identified. Characterization of the fission products and activation of each fuel pin was necessary to enable removal from storage and subsequent irradiation testing or PIE activities. The initial fuel pin geometry and composition were established using historical documentation. Irradiation history details were utilized to compute irradiated and decayed fuel pin masses, atomic densities, and activities. A pellet-wise axial neutron flux was also computed to assist in evaluation of the axial distribution of {sup 235}U and {sup 239}Pu, per pellet, within each of the test fuel pins. The results computed herein supported experiment design for advanced high-burnup MOX fuel designs under slow transient overpower (MOXTOP) conditions, which was Phase I of the ARES project. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Accident Tolerant Fuel Test (ATF-2c) Irradiation Test Report

A new fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2C has completed its first cycle of prototypic steady state irradiation. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and fast neutron flux histories of each of the 24 test pins has been calculated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Accident Tolerant Fuel Test 2B (ATF-2B) Irradiation Test Report

A fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2 has completed 7 cycles of prototypic steady state irradiation irradiating 41 test pins to burnups as high as 30 MWd/kgU. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and fast neutron flux histories of each of the 41 test pins has been calculated using a combined MCNP, and ORIGIN methodology to provide accurate and reliable power histories on a per pin basis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Hydrogen fuel quality analyzer with self-humidifying electrochemical cell and methods of testing fuel quality

A fuel quality analyzer for detecting contaminants in a fuel supply includes an anode flow field plate defining a first fuel flow field channel and a fuel inlet port, a cathode flow field plate defining a second fuel flow field channel and a fuel outlet port, a polymer electrolyte membrane between the anode and cathode flow field plates, a first electrode between the anode flow field plate and the polymer electrolyte membrane, and a second electrode between the cathode flow field plate and the polymer electrolyte membrane. The second electrode has a higher platinum loading than the first electrode. A reservoir volume is defined by the anode and cathode flow field plates. At least a portion of the polymer electrolyte membrane extends into the reservoir volume. The reservoir volume is configured to retain water to humidify the polymer electrolyte membrane.

08 HYDROGEN↗
NASA NTRS1985

Develop and test fuel cell powered on-site integrated total energy systems: Phase 3, full-scale power plant development

Two 25 cell stacks of the 13 inch x 23 inch cell size (about 4kW) remain on test after 4000 hours and 2900 hours, respectively, using simulated reformate fuel. These tests are focusing on the durability of fuel cell stack components developed through the end of 1983. Also, these stacks are serving as forerunners of a 25kW stack that will contain 175 cells of the same size and will employ the same technology base. The stack technology development program has focused on a new, low cost bipolar plate edge seal technique and evaluation of advanced cathode catalysts, an electrolyte replenishment system, and nonmetallic cooling plates in small stacks.

Kaufman, A.↗
DOE OSTI2022

Non-destructive analysis of swelling in the EMPIrE fuel test

The European Mini-Plate Irradiation Experiment (EMPIrE) was designed to support the development and testing of a coated uranium-molybdenum (U-Mo) dispersion fuel for the conversion of select high-performance research reactors (HPRRs) to utilize low-enriched uranium (LEU). To aid in the development of the coated fuel form, the EMPIrE test included several plate designs and irradiated them in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) at a high meat power density (~21 kW/cm 3 ) and to high fuel particle fission densities (~6.4 × 10 21 fissions/cm 3 ). These conditions mimic the bounding conditions of the BR-2 reactor in Belgium, where a concurrent irradiation experiment was performed, and exceed those previously explored in dispersion U-Mo fuel plates. A local fuel swelling analysis, as determined through high-fidelity, post-irradiation mini-plate profilometry, was used along with statistical methods to non-destructively evaluate the overall performance and separate the effects of convoluted fabrication variables. While some effects observed with this non-destructive analysis were subtle, others had more significant, and possibly competing, effects on the fuel swelling behavior. In closing, these observations will be examined further with destructive examinations to more fully assess them as the fuel design is developed and qualified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1981

Develop and test fuel cell powered on-site integrated total energy system. Phase 3: Full-scale power plant development

An integrated 5 kW power system based upon methanol fuel and a phosphoric acid fuel cell operating at about 473 K is described. Description includes test results of advanced fuel cell catalysts, a semiautomatic acid replenishment system and a completed 5 kW methanol/system reformer. The results of a preliminary system test on a reformer/stack/inverter combination are reported. An initial design for a 25 kW stack is presented. Experimental plans are outlined for data acquisition necessary for design of a 50 kW methanol/steam reformer. Activities related to complete mathematical modelling of the integrated power system, including wasteheat utilization, are described.

Kaufman, A.↗