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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Neutrino-detector design for safeguarding small modular reactors

Nuclear reactors have long been a favored source for antineutrino measurements for estimates of power and burnup. With appropriate detector parameters and background rejection, an estimate of the reactor power can be derived from the measured antineutrino event rate. Antineutrino detectors are potentially attractive as a safeguards technology that can monitor reactor operations and thermal power from a distance. Advanced reactors have diverse features that may present challenges for current safeguards methods. By comparison, neutrino detectors offer complementary features, including a remote, continuous, unattended, and near-real-time monitoring capability, that may make them useful for safeguarding certain classes of advanced reactors. This study investigates the minimum depth and size of an antineutrino detector for a small modular reactor to meet safeguards needs for advanced reactors. Furthermore, extrapolating performance from several prior reactor antineutrino experiments, this study uses an analytical approach to develop a possible design for a remote antineutrino-based monitoring device.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Model Development and Analysis of a High-Fidelity Neutron Transport Sensor: The Quadrupole Detector Concept for Measurement of the Neutron Flux Gradient

Accurate reconstruction of the neutron flux distribution within a reactor core is essential for safe and efficient reactor operation. Traditional power shape synthesis in Light Water Reactors relies on hundreds of in-core detectors. However, this approach becomes impractical for Advanced Reactors and Microreactors due to limited space and harsh environments. To address this challenge, we propose a data-driven methodology that combines high-fidelity modeling with real-time ex-core sensor measurements, enabling the reconstruction of core power distribution while minimizing the reliance on intrusive in-core instrumentation. This project began in FY24 and achieved two initial milestones: (1) the definition of a three-year development plan for a Digital Twin framework and (2) the development of high-fidelity neutronics models of the Purdue University Reactor One (PUR-1) using both MCNP6 and OpenMC. The PUR-1 reactor, a zero-power facility, was selected due to its suitability for neutronics-focused modeling and the availability of experimental data for validation. Both models were benchmarked using neutron flux measurements obtained from irradiated gold foils, which were strategically placed within the core during a dedicated campaign in July 2024. This report marks the continuation and completion of those foundational tasks. The OpenMC model has been refined (improved geometric accuracy, expanded cross-section libraries, and refined sampling) and validated using additional experimental data. An updated sensor design—based on quadrupole configuration—was designed to measure both ex-core flux and its spatial gradient. These measurements will serve as inputs to a neural network-based reconstruction algorithm. Finally, the methodology was demonstrated on a two-dimensional test case representative of the heterogeneous material composition of the PUR-1 reactor core. A neural network implementation of the Kirchhoff-Helmholtz integral equation was employed to solve the boundary value problem using peripheral sensor measurements. The preliminary results confirm the strong potential of the proposed approach for accurate and minimally invasive neutron flux reconstruction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

A Percolating Path to Green Iron

About 1.9 gigatonnes of steel is produced every year emitting 7% (2.7 gigatonnes) of global CO 2 in the process. More than 50% of the CO 2 emissions come from a single step of steelmaking, known as ironmaking. Hydrogen based direct reduction (HyDR) of iron oxide to iron has emerged as an emissions free ironmaking alternative. However multi-scale phenomena ranging from nanometers to meters inside HyDR reactors exhibit detrimental microstructure evolution which resists gaseous transport of H 2 /H 2 O, slows reaction rates and disrupts continuous reactor operation. To resolve the conundrum between atomic and reactor scales, we devise a percolation-theory model to reconcile nanoscale porosity with macroscopic properties relevant to reactor design models. Using synchrotron nano X-ray computed-tomography, we quantify the evolution of pores in iron oxide pellets, and demonstrate how nano-scale pore networks influence micro and macro-scale flow properties such as permeability, diffusivity and tortuosity. Our new modeling framework bridges the gap between scales and offers the criteria to accelerate HyDR by at least 5x via feedstock-reactor synergies based on percolation.

Paul, Subhechchha↗
NASA NTRS2009

Component and System Sensitivity Considerations for Design of a Lunar ISRU Oxygen Production Plant

Component and system sensitivities of some design parameters of ISRU system components are analyzed. The differences between terrestrial and lunar excavation are discussed, and a qualitative comparison of large and small excavators is started. The effect of excavator size on the size of the ISRU plant's regolith hoppers is presented. Optimum operating conditions of both hydrogen and carbothermal reduction reactors are explored using recently developed analytical models. Design parameters such as batch size, conversion fraction, and maximum particle size are considered for a hydrogen reduction reactor while batch size, conversion fraction, number of melt zones, and methane flow rate are considered for a carbothermal reduction reactor. For both reactor types the effect of reactor operation on system energy and regolith delivery requirements is presented.

Linne, Diane L.↗
DOE OSTI2020

Self-welding of Inconel 617 under high-pressure-high-temperature conditions for nuclear reactors

In response to continuously declining availability of fossil fuels, nuclear power has become an important alternative source of power generation. Growing application of nuclear power has relieved the shortage of electricity. However, it is also accompanied with new drawbacks, such as radioactive waste, and potential catastrophic consequences in case of major accidents. For nuclear power plants, severe accidents mostly originate from the reactor, making the reliability of the reactor of paramount importance to the overall safety of a nuclear power plant. Massive heat generation through nuclear fission/fusion and heat exchange render the whole reactor operating at extremely high temperatures (i.e., 900–1000 °C), which, in conjunction with large load from subsystems/ components made of special heavy alloys, renders the reactor under the risk of self-welding, a phenomenon that could lead to strong bonding between components and paralyze the operability of the nuclear reactor. This paper investigates the self-welding behavior of Inconel 617, a primary candidate nickel alloy for key components of the next generation gas cooled nuclear plant, inside a controlled atmospheric furnace with preset static load, simulating the high-pressure high-temperature conditions inside nuclear reactors. Therefore, test results show that Inconel 617 experiences self-welding through common oxide zone at the interface, and it has remarkable bonding strength, requiring 4114 N (57.15 MPa) and 4338 N (60.25 MPa) to break apart a mated sample pair, under an apparent contact pressure of 0.341 MPa after 50-h aging, in helium and air atmospheres, respectively. The higher oxygen presence in air results in slightly stronger oxide bond at the interface, compared to helium, and the bond strength increases with dwell time.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Data recovery via covert cognizance for unattended operational resilience

One of the important premises of unattended operation, a highly promoted characteristic of fission batteries and advanced microreactors, is the ability to automate the analysis of sensors data used in support of operational monitoring and control. Here, to meet this vision, this work proposes a new monitoring and data recovery paradigm to ensure resilience against data corruption which may be the result of malicious intrusion into the reactor operational network. This is paramount to ensure 100% availability under contingency scenarios such as cyberattacks. In support of this vision, earlier work has presented the concept of covert cognizance and demonstrated its mathematical ability to identify and embed cognizance parameters under the noise-dominated null space of the sensors data. This work extends this concept and applies it in real-time to demonstrate three key characteristics: zero-impact, zero-observability, and data recovery, where the first characteristic is to ensure no impact on operation, the second is immunity to discovery by pattern recognition techniques, and the third is to allow recovery of corrupt or falsified data. Recognizing that fission batteries are designed to operate under steady state most of the time, we elect to employ a small modular reactor model under transient operational conditions to demonstrate the operational resilience enabled by the covert cognizance paradigm. Specifically, the PI controller is augmented with the covert cognizance modules to develop self-awareness and enable automatic data recovery. The developed modules are expected to be equally applicable to a wide range of advanced reactor technologies relying on full or partial unattended control.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Advanced Reactor Control and Operations (ARCO): A University Research Facility for Developing Optimized Digital Control Rooms

The Advanced Reactor Control and Operations (ARCO) facility was constructed in January 2018 to serve as a test bed for advanced reactor control rooms and operator support systems. Since then, it has supported human-machine interface user experience research, fault detection and mitigation technology development, control room concept of operations development, and remote operations research. ARCO serves as the control room for the Compact Integral Effects Test (CIET) facility, which replicates the primary-side flow paths and thermal-hydraulic behavior of a fluoride-salt-cooled high-temperature reactor (FHR) using simulant fluids and scaling principles. New reactor designs feature different operating conditions and scenarios than those in existing reactors. ARCO supports the research and development of digital tools for operator communications, intuitive real-time data analysis, online health monitoring and prognostics, and control room cybersecurity. By integrating these different technologies, ARCO acts as a prototypical control system to iteratively develop methods and tools of operation in advanced small modular nuclear reactors. This paper describes the features of and challenges to operating advanced small modular reactors underlying the design basis for ARCO and its operator support systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

A Critical Review of the Experimentally Known Properties of U-Pu-Zr Alloys. Part 2: Thermal and Mechanical Properties

Although U-Pu-Zr alloys have been investigated for more than 60 years, relatively little experimental information is available, and many of the original values are in government reports that appeared more than 40 years ago. Information about the technologically important alloy U-20Pu-10Zr (wt%) is even more limited. U-Pu-Zr alloys are difficult materials to study experimentally, and it is therefore important to understand what results have already been obtained, how reliable they are, and where they were reported. This paper provides a summary and critical assessment of the available experimental measurements of thermal and mechanical properties of U-Pu-Zr alloys. Knowledge of these properties is crucial for understanding and modeling fuel constituent redistribution, fuel swelling and creep, fission gas release under normal reactor operations, and melting or formation of liquid phases under reactor transient scenarios. This paper builds on a previous paper that assessed experimental data about phases and phase diagrams in U-Pu-Zr alloys. Both papers are intended as resources for fuel designers and modelers and as guides for prioritizing future experimental work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

NASA Nuclear Thermal Propulsion: Thermodynamics of Zircaloy-4 Hydride

Nuclear thermal propulsion (NTP) is explored as a candidate technology by National Aeronautics and Space Administration (NASA) for Mars and other deep space missions. NTP reactors operate by passing its propellent, hydrogen, through a nuclear reactor. The nuclear reactor transfers energy to the hydrogen before hydrogen expands to generate thrust.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Numerical investigation of the AP1000 response following loss-of-coolant accident using PCTRAN and CFD to support R&D of SMRs

Small Modular Reactors (SMRs) present a promising solution for the future of sustainable energy, offering advantages such as reduced waste generation, advanced passive safety features, and potential cost efficiencies. Ongoing research focuses on the design and development of SMRs, addressing challenges through numerical simulations and experimental test data. Given that many next-generation reactors are first-of-a-kind (FOK), current operational reactors serve as valuable benchmarks for understanding various thermal-hydraulic phenomena during postulated design basis accidents (DBAs), including loss-of-coolant accidents (LOCA), main steam line breaks (MSLB), and steam generator tube ruptures (SGTR). This study employs the AP1000 model Personal Computer Transient Analyzer (PCTRAN) to simulate reactor responses to small-break LOCA scenarios, specifically analyzing break sizes of 2 and 10 inches. Additionally, ANSYS FLUENT software is utilized to assess containment responses to large-break LOCAs, focusing on the quantification of decay heat removal via natural convection. Verification of the PCTRAN results is achieved using Westinghouse data, with CFD results corroborating the findings. The outcomes demonstrate strong agreement with Westinghouse data, confirming the accuracy of the simulations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

An Approach to Automate tools for the Risk Assessment of Digital Instrumentation and Control Systems

Reliable digital instrumentation and control systems (DI&C) are integral for sustaining the continued operation of nuclear power plants. These systems ensure that nuclear reactors operate safely, efficiently, and within regulatory requirements. Yet, the cost of designing and licensing new nuclear DI&C can be prohibitively expensive. Under the U.S. Department of Energy Light Water Reactor Sustainability Program, Idaho National Laboratory has developed a framework for supporting the risk-informed design of DI&C systems by offering methods to support the identification, quantification, and evaluation of risks for various DI&C design architectures. The framework indicates potential software failure modes and provides pathways for quantifying the potential for these software failures, including common cause failures. Using the framework’s systematic approach, challenges for assessing risks within new and existing nuclear DI&C systems can be reduced. Nevertheless, the current framework can be further improved using the convenience of automation. This paper introduces the development of Software for the Hazard Identification and Evaluation of Digital Systems (SHIELDS). SHIELDS is an engineering software package that enables the identification, elimination, and mitigation of potential risks and reduces the burden of deploying reliable DI&C systems. This work introduces plans and techniques to digitize and improve the manual risk assessment modules of the framework. These improvements will save time and increase the repeatability and usability of the framework, making it more accessible to a wider range of users. Ultimately, this introduces SHIELDS and how its modules support efficient development of safe and reliable DI&C systems.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2024

UO 2 -Liquid Metal Suspension Fuel Concept: Alloy Selection and Depletion Calculations

A novel nuclear fuel was recently proposed that consists of UO 2 particles suspended in a liquid metal mixture. Two liquid metal candidates were identified: BiPbSn and PbSn. Scoping simulations were performed on an over-simplified reactor core that showed a potential limit on achievable burnup. Those findings led to two questions that needed to be addressed: What is the corrosion behavior of the liquid metals toward Zircaloy-4 (Zr4)? And what level of burnup can the fuel be expected to achieve? This report documents efforts to answer both of those questions. First, corrosion experiments were performed between the liquid metal candidates and Zr4 coupons. Second, the previous scoping simulations were expanded to include depletion analyses so that maximum burnups could be estimated. At 400°C, neither PbSn nor BiPbSn caused significant corrosion to the Zr4 coupons. At 600°C, BiPbSn did not cause significant corrosion, but PbSn did. In addition, pure bismuth and pure tin were both highly corrosive, whereas pure lead was non-corrosive. The depletion simulations found that low-enriched fuel could maintain light-water reactor operating temperatures but achieve a burnup of only about one-third of existing reactors. Higher enrichments could achieve burnups equal to or higher than those of existing reactors but at the expense of higher operating temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Design and characterization of an electromagnetic-resonant cavity microwave plasma reactor for atmospheric pressure carbon dioxide decomposition

Plasma processes are ideally suited for the conversion of renewable electricity into gas-phase reactivity, such as for the decomposition of carbon dioxide (CO 2 ). The design, development, and characterization of a microwave plasma reactor for atmospheric pressure undiluted carbon dioxide decomposition are presented. The reactor operates as an electromagnetic-resonant cavity in which the generated plasma forms a bulb attached to a converging-diverging nozzle and stabilized by streams of tangentially-injected processing gas. Electromagnetic wave confinement, residence time, and critical gas vorticity constitute fundamental reactor sizing and operation parameters. Experimental findings demonstrate that flow rate plays a dual role in plasma stabilization and process performance, whereas deposited power has a minor role in CO 2 decomposition.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Overview of simulation and control tools for extended operability of nuclear reactors - 183

The increasing penetration of Renewable Energy Sources (RES) requires the current power grid to be suitably improved. Because of the unsteady nature of RES, as long as energy storage facilities are not deployed and commercialized, an increased effort for the frequency regulation in order to ensure high reliability and performance is necessary. In such a scenario, nuclear units (traditionally deployed as baseload generating units) will be required to cooperate with intermittent RES-based power plants. Therefore, new reactors will need larger operational flexibility with respect to the capabilities of the currently operated units. Over the last few years, a comprehensive research program oriented at developing a suite of simulation tools and control system development capabilities was pursued at Argonne National Laboratory. In addition to the study of advanced reactor concepts (e.g. metal-fueled Sodium-cooled Fast Reactors) and technological innovations (such as S-CO{sub 2} Energy Conversion Cycles), new control approaches were developed to improve the operability of nuclear power plants. In particular, Model-based Predictive Control algorithm for Multiple Input Multiple Output control strategies and set-point definition algorithm based on Reference Governor were studied and applied to specific scenarios, providing with a range of novel capabilities able to meet the expected greater sets of requirements mentioned above. In this work, an overview of such tools and case studies is presented with particular focus on their impact on operability and economic competitiveness of nuclear energy. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Volume 1: Introduction to the HFIR Futures - Enhanced Capabilities Series

Since it began operating at full power in 1966, the High Flux Isotope Reactor (HFIR) has contributed unparalleled neutron science capabilities to research on neutron scattering, isotope production, materials and fuels irradiation, and neutron activation analysis. HFIR is a high-performance, multi-mission research reactor operated on behalf of the US Department of Energy (DOE) at the Oak Ridge National Laboratory (ORNL). In 2020, a DOE Basic Energy Sciences Advisory Committee Subcommittee (BESAC) published a report recommending that DOE make significant investments in HFIR to enable continued operations beyond the year 2100 while also enabling new research and isotope production capabilities. ORNL organized an initiative to investigate how to specifically address the report’s key recommendations, including enabling long-term operation, brainstorming future scientific research needs, and outlining the infrastructure required to realize these future research capabilities. A multivolume series of reports has been developed to document the nonscattering enhancements. This volume, the introductory report, provides an overview of HFIR and the initiative.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Advanced Moderation Module for High-Temperature Micro-Reactor Applications

Development and deployment of micro-reactors that provide competitive efficiency, prevailing compactness, and inherent safety is an immediate focus of the U.S. nuclear industry. For thermal neutron micro-reactors operating at elevated temperature for optimized efficiency, such as micro molten-salt reactors (MSRs), heat-pipe reactors, and very-high temperature reactors (VHTRs), high-performance moderator based on metal hydride can enhance the neutron economy and therefore achieve reduced weight and enhanced portability. As unclad metal hydride inevitably decomposes at elevated temperature, an enclosure is required for hydride moderator to deliver desired performance at elevated temperatures. Conventional moderator enclosure solutions based on high-temperature alloys introduce extraneous neutron penalty into the micro-reactor, affecting the neutronic benefits provided by the hydride moderator. Additionally, the compatibility between the high-temperature alloy enclosure and the micro-reactor matrix is also a potential issue. In this report, we disclose an Advanced Moderator Module (AMM) concept enabled by an innovative enclosure solution combining the advantages of refractory metals, ceramic matrix composites (CMCs), and advanced coating technology to serve as hydrogen permeation barrier up to very-high temperatures. A schematic description of the AMM structure is illustrated in Figure A1. The AMM contains a moderating material core made of metal hydride with high thermal stability, such as YH 2-x . The hydride core is enclosed by an H 2 barrier layer coated on a ductile refractory metal liner to minimize hydrogen loss during high-temperature operation. A ceramic matrix composite (CMC) cladding is adopted to provide further structural strength, especially during power transients. Between the CMC cladding and metal liner, an extra diffusion barrier coating is inserted to suppress the chemical interaction at elevated temperatures. Hence, based on a series of innovative material solutions, the AMM is capable of containing the metal hydride core at elevated temperature (>900oC) inside coated and lined CMC envelop with negligible hydrogen loss throughout the microreactor lifetime. The benefit of the AMM technology was assessed based on a comprehensive multi-stage reactor-physics analysis. Advanced moderation based on hydride metals (such as YH2-x) enables reaching optimum moderation with higher fuel content than traditional VHTR technology, which is required to design compact micro-reactor cores targeting long-life operation. The AMM enclosure technology provides lower thermal neutron absorption rates than traditional solutions (based on stainless steel or Mo-based alloy such as TZM), which reduces the fissile enrichment requirements by 6-8% on a TRISO-fueled design based on the EMPIRE core. Finally, combining the hydride moderator with neutron transparent enclosure solutions provides significant potential to boost neutronics performance of micro-reactors in terms of increased core lifetime or reduced size and weight by 30-50% on a micro-reactor based on the Holos Quad technology. The progress and plans for on-going development and demonstration efforts are also discussed in this report. The current demonstration of AMM is focused on the hydrogen diffusion barrier demonstration under thermal cycling, while radiation tolerance demonstration is planned. The next step of the demonstration plan will be the assembly of miniature AMM for high-temperature testing, which can be used as a prototype for future pilot scale demonstration.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

SCALE Analyses of Scenarios in the Molten Salt Reactor Fuel Cycle

In support of the US Nuclear Regulatory Commission non-light-water reactor fuel cycle demonstration project, the capabilities of SCALE 6.3.1 for radionuclide characterization, criticality, and shielding were demonstrated through several scenarios in a molten salt reactor (MSR) nuclear fuel cycle. Three scenarios were selected for this study, with the molten salt breeder reactor (MSBR) serving as the reference design. The original thorium-based fuel salt used in MSBR was replaced with a 235 U-enriched fuel salt to better reflect anticipated future MSR concepts. The first scenario focused on the fuel salt preparation stage. Criticality analyses, using SCALE/CSAS-Shift and SCALE/SAMPLER, examined potential critical configurations in the fresh fuel salt container based on variables such as 235 U enrichment, UF 4 molar fraction, temperature, and container geometry. The second scenario explored the release of fission products during reactor operation in which the tritium buildup in the primary fuel salt was quantified through depletion calculations using SCALE/TRITON with continuous fuel makeup and fractional fission product removal. This scenario also assessed radioactivity levels in the off-gas system (OGS), quantifying radiotoxic nuclides that could be released into the environment. SCALE/SAMPLER was employed to perform uncertainty quantification to identify key operating parameters influencing the buildup of radiotoxic nuclides in the OGS. Additionally, radiation dose rates were calculated using SCALE/MAVRIC to estimate the maximum exposure levels within the reactor cell complex during operation. The final scenario evaluated the decay heat and radioactivity of the primary fuel salt collected in the drain tank, as well as the radiation dose rate outside the building from the presence of the drain tank. Overall, the application of SCALE’s various capabilities for radionuclide inventory generation, criticality analysis, and shielding in MSR fuel cycles was successfully demonstrated across these scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Low cost modular plasma system for reforming of natural gas

The initial aim of this project was to develop a low-cost compact, modularly scalable, point of use, on demand production system for the reforming of natural gas (NG) into products such as acetylene, hydrogen and other related hydrocarbons. Based on a pulsed dielectric barrier discharge (DBD), the fast rising voltage pulses produced by the system causes the dielectric breakdown of the gas to lag behind the rapid voltage increase so that a voltage greater than the static breakdown voltage of the gas, called an overvoltage, is briefly attained. The century long commercial use of the DBD discharge to push the highly endothermic oxygen to ozone reaction served as an additional motivation. The project meets the recommendations of the National Research Council 2007 report on plasma science for a greater investment in low temperature plasma science and engineering. An important finding is that the method investigated has a high specific energy requirement (SER) of the order of 1,000 to 5,000 kWh per kg of hydrogen, and a corresponding low efficiency of only ~0.1 to 0.5% of converting electrical into chemical (ΔH, ΔG) energy. This compares unfavorably with 52 kWh/kgH 2 required for NG thermal plasma reforming, or a pulsed spark discharge, Lotfalipour et. al, and Gao et. al. (For comparison electrolysis of water into hydrogen and oxygen gas requires only 45 to 80kWh/kgH 2 ). This project garnered data with different reactor configurations, materials, gases and mixtures including NG, methane, hydrogen, propane, argon, nitrogen and air, gap sizes ranging from 0.5 to 4.1 mm, electrode areas from 5 to 460 cm, voltages to 27 kV, pulse frequencies to 20 kHz and peak currents into the hundreds of amperes, temperatures to 625 °C and pressures from 15 mbar to 2 bar. An SRS 300 residual gas analyzer (RGA) yielded quadrupole mass spectrometry (QMS) data on the reformate. To analyze the data we extended the public domain software OpenChrom®, Wenig, to QMS data, yielding accurate identification and quantitative data on reformate products. This extension of OpenChrom® is applicable beyond this project and is a significant contribution to the general community of researchers. The documentation is available on the internet, see Whitlock, and a copy has been appended to this report. The main product of methane and natural gas plasma reforming is hydrogen, constituting ~90% of the products, followed by 1% to 10% ethane, propane, ethylene, acetylene and isobutene. Neopentane, propene, butane and pentane constitute less than 1 % of product. Adding a nickel catalyst increases the selectivity for hydrogen. Carbon nanosheets (graphene), a product of commercial interest, were synthesized by plasma reforming of acetylene at low pressure. The reformate analysis shows hat saturated and branched hydrocarbons are favored compared with unsaturated and linear hydrocarbons. For example, ethane, propane, iso-butane, 2-methyl butane, and 2-methyl pentane are produced in higher amounts compared with unsaturated and linear components. At low pressure acetylene and ethylene production rate is higher and propane production lower. We developed a recirculating system in which hydrogen is removed through a permeable membrane as fast as it is being produced by the reactor and the remaining products are recirculated at a high flow rate of several liters per minute. This keeps the concentration of H 2 in the reactor low, while achieving a total output concentration of 79%, high enough for pressure swing adsorption to be used for further purification. The recirculating system could be used with other gas reforming technologies. We report on four different reactors,1) a spark discharge reactor, 2) a standard DBD cylindrical reactor, 3) a small DBD reactor where catalyst powder can be quickly loaded onto a carrier to investigate the effect of the plasma on the catalyst and the effectiveness of the of the catalyst in promoting gas reformation and 4) a configurable DBD reactor to study the performance effects of dielectric selection, electrode gap distance, and electrode/dielectric interface on plasma reactor operation. This project achieved a commercializable 1kW plasma system including reactors a plasma pulse generator and a gas supply module.

03 NATURAL GAS↗