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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
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 OSTI2016

Prototype Demonstration of Gamma- Blind Tensioned Metastable Fluid Neutron/Multiplicity/Alpha Detector – Real Time Methods for Advanced Fuel Cycle Applications

The content of this report summarizes a multi-year effort to develop prototype detection equipment using the Tensioned Metastable Fluid Detector (TMFD) technology developed by Taleyarkhan [1]. The context of this development effort was to create new methods for evaluating and developing advanced methods for safeguarding nuclear materials along with instrumentation in various stages of the fuel cycle, especially in material balance areas (MBAs) and during reprocessing of used nuclear fuel. One of the challenges related to the implementation of any type of MBA and/or reprocessing technology (e.g., PUREX or UREX) is the real-time quantification and control of the transuranic (TRU) isotopes as they move through the process. Monitoring of higher actinides from their neutron emission (including multiplicity) and alpha signatures during transit in MBAs and in aqueous separations is a critical research area. By providing on-line real-time materials accountability, diversion of the materials becomes much more difficult. The Tensioned Metastable Fluid Detector (TMFD) is a transformational technology that is uniquely capable of both alpha and neutron spectroscopy while being “blind” to the intense gamma field that typically accompanies used fuel – simultaneously with the ability to provide multiplicity information as well [1-3]. The TMFD technology was proven (lab-scale) as part of a 2008 NERI-C program [1-7]. The bulk of this report describes the advancements and demonstrations made in TMFD technology. One final point to present before turning to the TMFD demonstrations is the context for discussing real-time monitoring of SNM. It is useful to review the spectrum of isotopes generated within nuclear fuel during reactor operations. Used nuclear fuel (UNF) from a light water reactor (LWR) contains fission products as well as TRU elements formed through neutron absorption/decay chains. The majority of the fission products are gamma and beta emitters and they represent the more significant hazards from a radiation protection standpoint. However, alpha and neutron emitting uranium and TRU elements represent the more significant safeguards and security concerns. Table 1.1 presents a representative PWR inventory of the uranium and actinide isotopes present in a used fuel assembly. The uranium and actinide isotopes (chiefly the Pu, Am and Cm elements) are all emitters of alpha particles and some of them release significant quantities of neutrons through spontaneous fissions

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
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 OSTI2016

Reactor Physics Analysis of Transitioning to a Thorium Fuel Cycle with Molten Salt Reactors

Commercial power reactors operating in the United States are light-water reactors (LWRs) fueled with low-enriched uranium (LEU) in a once-through fuel cycle; the important fissile isotopes are {sup 235}U, {sup 239}Pu, and {sup 241}Pu. In a thorium fuel cycle, the fertile isotope {sup 232}Th is converted into {sup 233}U, a fissile isotope that is not naturally occurring. In transitioning from the once-through fuel cycle to the thorium fuel cycle, the availability of man-made {sup 233}U presents a challenge: without it, the initial fissile material fueling the first reactors built in this transition is not the fissile isotope that the reactor is breeding. The transition requires an alternate fissile material until enough surplus {sup 233}U is available to start up additional reactors. In liquid-fueled molten salt reactors (MSRs), the fuel is dissolved into molten salt, which is continuously circulated through the core, simultaneously undergoing irradiation, chemical treatments and separations, and feeds. MSRs are an attractive technology in which the thorium fuel cycle can be deployed, as the {sup 233}U bred from {sup 232}Th is continuously being recycled back through the core. The bred fissile material is immediately available to fuel the core without processing or fuel fabrication. After start-up, only a fertile {sup 232}Th feed is necessary to sustain criticality in the MSR. But, in single-fluid MSRs, the initial fissile and fertile materials are mixed into the same molten fuel salt, complicating separations of initial and bred fissile materials. In solid-fuel systems, these materials are often separate. This analysis herein builds on previous work modeling thorium-fueled MSR start-up. It is based on unit cell representations of the graphite-moderated, 2250 MWt Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR). This paper discusses the start-up of the first MSR in the transition to a thorium fuel cycle and identifies challenges and limitations of fuel cycle deployment with this reactor technology. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Detailed Study of the Transient Rod Pneumatic System on the Annular Core Research Reactor

The Annular Core Research Reactor Facility (ACRRF) at Sandia National Laboratories (SNL) houses the Annular Core Research Reactor (ACRR), which is a TRIGA design with a few modifications by General Atomics and SNL. Being a TRIGA design, the reactor operates in either steady-state or pulse mode, the latter caused by rapid ejection of three transient rods (TRs) with attached poison rods from the reactor core. The primary mission of the ACRR is to provide a significant flux of high energy neutrons into a dry central cavity for experiments. This research was prompted by intermittent failures in the dashpot rod in TR A. Throughout the 40+ year history of operation, all TRs have failed but TR A has exhibited an increased rate of failure as compared to TR B and C. The research presented here is a foundation for understanding why TR A is failing more often. This paper focuses on the analysis of a pressure distribution and derives from it displacement and timing of the TRs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2015

Economics of Hybrid-Energy Systems and the Impact on Advanced Reactor Design and Operation

The economics of electricity production in the U.S. is changing with the introduction of recent federal regulations limiting carbon dioxide emissions in the electric power industry. This has already resulted in the closing of coal plants, plants which in the past to a degree served to balance production with daily variation in load. The resulting loss of electricity production is being made up with wind and solar renewables and gas-fired units, and with an increasing need for energy storage. It would appear that with the loss of the baseload component of the coal plants, the duty-cycle of the gas-fired units may necessarily increase, encroaching on the traditional nuclear plant domain with its inexpensive fuel. Gas plants are costly to operate on a continuous basis with relatively volatile fuel prices. It appears then that nuclear plants that are capable of flexible operation have the potential on an economic basis to displace gas-fired power units. For flexible operation with renewables advanced reactors have advantages over the current generation of pressurized water reactors (PWR) and the new small modular reactors (SMR). For PWRs and SMRs, pellet-cladding mechanical interaction, xenon poisoning, and thermal stress concentrations in structures place limits on the all-important response time for operation with renewables. These phenomena are, however, not as limiting for the liquid-metal fast reactor (LMR). This may permit greater power ramp rates than in a light water reactor which bestows an economic advantage in energy systems with renewables. An economic model is developed for studying electric grid options including plant mix and market rules for the case of a high renewable fraction. It returns the minimum cost solution from among a mix of plant types including nuclear, gas, and storage (i.e. hybrid energy systems) for a given fraction of electricity production by renewables. This model can be used to assess the economic advantage of nuclear plants over gas plants and advanced reactors over light water reactors in an electric grid with a high fraction of renewables. The model has an explicit treatment of plant dynamic effects, an important feature for analysis of hybrid energy systems with a renewables production component that is unpredictable and uncontrollable.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
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 OSTI2016

Temperature and neutron flux dependence of in-reactor creep for cold-worked Zr-2.5Nb - 2016-0042

Over the past 10 years, internally pressurized capsules made from Zr-2.5Nb tubing have been irradiated in the National Research Universal (NRU) reactor at Chalk River Laboratories at temperatures of 280, 320, and 340 deg. C and dose rates between 3 X10{sup 15} n. m{sup -2}. s{sup -1} and 2 X10{sup 17} n. m{sup -2}. s{sup -1} (E>1 MeV). Periodic gaging has been used to assess the primary and secondary (steady-state) creep behavior. The objective of this detailed and controlled experiment was to determine, for the first time, the creep and microstructure evolution in Zr-2.5Nb tubing over a wide range of irradiation conditions for fast neutron fluxes applicable to a CANDU pressure tube. Similar but 'accelerated' creep experiments have been conducted in the Osiris test reactor at fast neutron fluxes of approximately 1.8 X10{sup 18} n. m{sup -2}. s{sup -1} (E>1 MeV), much greater than the neutron fluxes in the NRU reactor. Although accelerated tests in high-flux reactors such as Osiris provide information on irradiation creep, they do not represent the neutron flux conditions applicable to a power reactor. Tests covering power reactor operating conditions are needed to develop models for in-reactor creep of pressure tubes under the appropriate conditions. The data from the NRU reactor are compared with results from creep capsules with similar starting microstructures but irradiated in the Osiris reactor. The results show that the steady-state diametral and axial creep rates have a complex dependence on stress, temperature, and fast neutron flux. Data from out-reactor creep tests on unirradiated and pre-irradiated creep capsules that show the effect of prior irradiation on creep are also reported. The results are discussed in terms of a combination of creep mechanisms involving dislocation glide and mass transport. (authors)

36 MATERIALS SCIENCE↗
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↗