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At least 1,531 records · Page 85

Fluidized-Bed Gasification of Coal-Biomass-Plastics for Hydrogen Production

Coal is one of the most abundant fossil energy resources in the United States and in the world. The recoverable reserves in the United States are estimated to be about 252 billion tons – more than 350 years of supply at current rates of usage. However, the share of coal in total primary energy consumption in the US has been declining over the years. The decline of coal is mainly attributed to cheap natural gas and precipitous declines in the cost of electricity production from renewable technologies such as wind and solar. Coal can potentially be used if it is coupled with carbon-neutral feedstock such as biomass-agricultural residues, forest biomass, and forest residues. Co-gasification of coal and biomass can become a negative carbon emission technology if the carbon dioxide (CO 2 ) is captured and sequestered. Although the biomass gasification process has a lot of similarities to coal gasification, the large-scale adaptation of power production sourced from biomass has not come to fruition. The main reason is that the power production from biomass is still expensive when compared with natural gas or coal power technologies. To address the feedstock cost, one approach is to use low-cost feedstocks, such as municipal solid wastes (MSW) or plastics, for gasification. Gasification involves the partial oxidation of coal and/or biomass feedstocks to produce a combustible fuel called synthesis gas (syngas) which is composed of carbon monoxide (CO), hydrogen (H 2 ), CO 2 , methane (CH 4 ), nitrogen (N 2 ), water (H 2 O), and other compounds that might be considered as contaminants. Raw syngas from gasification must go through multiple steps to produce high-purity hydrogen. The specific steps depend upon the quality (gas composition, contaminants, and their concentration) and condition (pressure and temperature) of syngas. The long-term goal of the project was to utilize coal and plastics together with biomass to produce energy and fuels using a gasification platform while reducing greenhouse gas emissions. The main objective of this research was to examine gasification performance in a laboratory-scale fluidized-bed gasifier for hydrogen production. The specific objectives of the research were to: (i) study coal-plastic-biomass mixture flowability for consistent feeding in the gasifier; (ii) understand the gasification behavior of the mixture in steam and oxygen environments; (iii) perform thermal property characterization of ash and slag from the mixture feedstock and refractory-ash interface of the mixture under gasification conditions; and (iv) develop process models to determine the technology needed for syngas cleanup and contaminants. The study found that there was no apparent segregation when biomass, coal, and waste plastics were mixed together during feeding. Although there were differences in hydrogen production when individual feedstock were fed, the hydrogen concentration remained almost constant with various blends. Therefore, blending waste plastics with biomass and coal, which are all abundant, is a better approach for energy production. Results of the techno-economic analysis suggested that integration of advanced gasification (GTI’s R-GAS™) and syngas cleanup and conditioning technologies (RTI’s WDP and AFWGS) for clean hydrogen production resulted in substantial benefits, including significant capital cost and operating cost reductions. Advanced technologies resulted in 16% reduction in the hydrogen production cost (COH) from 2.94 $\$$/kg to 2.47 $\$$/g, with further scope for optimization and cost reduction. These advanced technologies also result in lower emissions, and improved energy efficiency.

01 COAL, LIGNITE, AND PEAT↗

MeV-scale Physics in MicroBooNE

The scope of this public note is to present preliminary measurements of MeV energy signatures and relevant backgrounds for beam neutrino interactions using a dedicated reconstruction technique, expanding upon the original study by ArgoNeuT. We also use this technique to identify and quantitatively analyze the low energy activity emanating from the field cage support ribs in the MicroBooNE detector and discuss how the usage of similar construction materials could be a source of background for low energy physics analyses in future neutrino detectors such as the Deep Underground Neutrino Experiment (DUNE). We also highlight the application of the low energy reconstruction technique to studies of neutrino interactions from core-collapse supernovae and muons decaying at rest ($\mu$DAR).

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

NEAMS Burnup Extension Accomplishments and Remaining Modeling Gaps

The economic viability of light-water reactors (LWRs) in the United States is declining in heavily subsidized markets, and as a result, the nuclear industry is looking for opportunities to enhance the economic competitiveness of nuclear power. This is not a foreign concept to the nuclear industry: in the mid-2000s, the nuclear industry set out to achieve zero fuel failures by 2010. The goal in this effort was to drive down the cost of reactor shut down by replacing a pin or bundle in response to fuel rod failure. 2010 brought about the initiative to deliver the nuclear promise to reduce operating cost by 30% to improve nuclear energy’s economic competitiveness before 2020. The emergence of accident-tolerant fuel also offers the nuclear industry an opportunity to build on these past successes and deliver affordable, clean energy. Accident-tolerant fuel has been shown to provide superior performance compared to traditional Zircaloy/UO2 fuel concepts, offering the unique ability to remove operational limitations that inhibit the economic viability of nuclear power. This has led the industry to begin building a technical case to extend the peak rod average burnup beyond 62 GWd/tU to extend pressurized water reactor cycle lengths to 24 months and to develop more efficient boiling water reactor core designs. The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program mission is to develop advanced modeling and simulation tools and capabilities to accelerate the deployment of advanced nuclear energy technologies. The primary safety concern inhibiting the nuclear industry from extending burnup is related to high-burnup fuel fragmentation, relocation, and dispersal. Therefore, the NEAMS program developed a targeted 5-year plan to support the industry’s efforts to extend burnup. This milestone report summarizes the 5-year plan that was enacted in FY20, followed by a discussion of the ongoing activates required to fulfill the 5-year plan, as well as the approach to address the current modeling gaps. Additionally, an LWR stakeholder meeting was held to communicate work performed in the NEAMS program over the past three years, to assess the LWR community’s perspective on the impact of the program, and to identify remaining significant gaps in the NEAMS suite of capabilities. This engagement will be documented by the Electric Power Research Institute and used by NEAMS to redirect current LWR scope as needed and to develop the next phase for LWR research and development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Adaptable Technology Integration for Enhanced Worker Safety and Efficiency within the SPD Gloveboxes [Slides]

Surplus Plutonium Disposition (SPD) Overview: (1) Application Background – Location: Savannah River Site, Aiken, South Carolina – High backlog of plutonium material to process – Parallel lines of gloveboxes – Highly manual operations requiring full PPE and reader/worker steps; (2) Development Scope – Development of automated processes and opportunities to provide robust engineering design alternatives for baseline design that would meet process improvement goals.

61 RADIATION PROTECTION AND DOSIMETRY↗

High Flux Isotope Reactor Uncertainty Factors

This report provides a brief summary of the uncertainty factors used for High Flux Isotope Reactor (HFIR) steady-state heat transfer analyses. These factors are mainly used in the HFIR Steady-State Heat Transfer Code (HSSHTC) to perform core thermal margin evaluations that determine safe reactor operation. The attempt to classify these factors stems from the assumption that the current approach is characterized by an excess of conservatism, thereby restricting reactor performance. The work documented herein was of a limited scope and pertained mainly to factors’ description and initial grouping based on their functional use. Suggestions are provided for further evaluation for the low-enriched uranium (LEU) to the uranium silicide dispersion fuel (U 3 Si 2 -Al).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Post-irradiation examination of legacy high burnup fuel to support safety testing

Safety/transient testing to evaluate performance under off-normal conditions is an essential pillar for both the development of Accident Tolerant Fuels (ATF) and the optimization of fuel operation economics beyond current discharge burnups. Among other factors, the successful interpretation of the transient testing results relies upon the knowledge of the initial conditions of the test, including the characteristics of the fuel system under scrutiny. When testing pre-irradiated material, the assumptions that the fuel and the cladding still have the same properties as in the pre-irradiation stage is obviously wrong and could affect the results of the test. This is particularly true the more burnup accumulates in the fuel rod and irradiation progresses. The knowledge of the initial microstructure of both fuel and cladding allows a clearer interpretation of the subsequent transient testing results, provides validation of the physical phenomena underlying the model predictions and eliminates the uncertainties related to the limited knowledge of the sample status before the test. One example is the phenomenon of fine fragmentation that occurs in Light Water reactor (LWR) fuel. During a Loss of Coolant Accident (LOCA) or Reactivity Initiated Transient (RIA) the fuel can severely fragment. During LOCA, high burnup fuel tend to finely fragment, which has raised safety concerns due to the increased likelihood of dispersal of such small particles once the cladding has burst and due to the increased fission gas release. Therefore, efforts have been devoted to the assessment of a pulverization threshold that could determine the conditions under which fine fragmentation is predominant. However, the lack of information regarding the initial conditions of the fuel, and the connections between those conditions and the pre-transient irradiation history, have hindered the development of a fully mechanistic fragmentation and pulverization criterion. The empirical relationships rely on conservative estimations, due to the lack of information on critical material properties and characteristics. More generally, experimental evidence of the irradiation-induced modifications at microstructural scale are necessary to determine the behavior of the material at the macroscopic scale, with the latter being the one of technological interest. Significant progress has been made in the last two decades in the developments of analytical materials science techniques that can be applied to highly radioactive materials, such as high burnup fuels. The availability of new techniques and the improvement of existing ones has enabled investigations previously not possible that can deepen the understanding of the fuel characteristics and properties at high burnup. The better knowledge of material behavior and irradiation-induced phenomena could help the prediction of its performance. In this context, the scope of the present work is to apply a wide portfolio of advanced characterization techniques to determine properties that are relevant for safety and performance. The results are interpreted in the context of engineering scale post-irradiation examinations and available information on the irradiation conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Neutron Dosimetry for the University of Central Florida (UCF3) Irradiation in ATR

PNNL project 74242 involves the analysis of neutron fluence monitors irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory in accordance with MPO 00236287 and Statement of Work No. 17370, Rev. 0, PNNL Analysis of NSUF Flux and Melt Wire Capsules. This report is for the University of Central Florida (UCF)-3 third stage experiment which was conducted in position B8 of the ATR. Three other irradiations included in the scope of work are reported separately. The neutron fluence monitors were prepared by PNNL and loaded into the UCF-3 assemblies at INL prior to irradiation. Following irradiation, the capsules were returned to PNNL for analysis. The neutron dosimetry capsules were opened, the flux wires were removed for gamma analysis, and the measured activities were used to determine the activation rates for various activation products. Following suitable corrections, the measured activation rates were used to adjust calculated neutron spectra at 8 fluence monitor locations. The adjusted neutron spectra were then used to determine displacement per atom (dpa) and gas production for irradiated materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Neutron Dosimetry for the Colorado School of Mines (CSM 16-10584) Irradiation in ATR

PNNL project 74242 involves the analysis of neutron fluence monitors irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory in accordance with MPO 00236287 and Statement of Work No. 17370, Rev. 0, PNNL Analysis of NSUF Flux and Melt Wire Capsules. This report is for the Colorado School of Mines (CSM 16-10584) experiment which was conducted in position B5 of the ATR. Three other irradiations included in the scope of work are reported separately. The neutron fluence monitors were prepared by PNNL and loaded into the CSM assemblies at INL prior to irradiation. Following irradiation, the capsules were returned to PNNL for analysis. The neutron dosimetry capsules were opened, the flux wires were removed for gamma analysis, and the measured activities were used to determine the activation rates for various activation products. Following suitable corrections, the measured activation rates were used to adjust calculated neutron spectra at 12 fluence monitor locations. The adjusted neutron spectra were then used to determine displacement per atom (dpa) and gas production for irradiated materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of Automation Technologies for Glovebox Processing, Part 2

Topics • Surplus Plutonium Disposition (SPD) Overview • Robotic Arm Development for Glovebox Operations • Automatic Sphincter Loading System (ASLS) Update Surplus Plutonium Disposition (SPD) Overview • Application Background • Development Scope • Improvement Goals • Limitations

Krementz, Daniel↗

A Conceptual Design of the Reactor Cavity Cooling System for the Horizontal Compact High Temperature Gas Reactor (HC-HTGR)

Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multi-disciplinary team of nuclear, mechanical, and structural engineers under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed HC-HTGR in 3 years and support its commercialization as a safe, low-cost HTGR. Argonne National Laboratory (Argonne) is responsible for the design and analysis of the reactor cavity cooling system (RCCS) as a safety system for passive decay heat removal of the reactor concept. This report documents the design study to derive a conceptual design study of the RCCS for the HC-HTGR. It includes the identification of the functions and requirements of the HC-HTGR RCCS, design analyses including high-level design consideration and the calculations for optimizing design space of the system with supporting component-level analysis to inform the material selection and performance of the water panel, the description of the conceptual design of the HC-HTGR RCCS derived based on the analyses results, and performance evaluation of the conceptual RCCS for the HC-HTGR. A detailed concept of the RCCS has been identified and high-level system requirements has been developed for the HC-HTGR. Design space focusing on the natural circulation loop portion of the RCCS has been investigated to optimize the system performance. The initial baseline dimensions were firstly derived based on the scoping calculations. A component level design analysis was conducted for the water panel to inform the material selection and to assess its conduction performance. A preliminary system-level performance analysis was performed for the 1/8th of the compartment of the initial baseline design of the RCCS using RELAP5-3D. To improve the system thermal performance, the RCCS design has been updated by exploring various design options by design parametric analyses. Based on the results, the conceptual design of the RCCS for the HC-HTGR has been derived, which satisfies the target performance of ~1 MWt at the elevated vessel wall temperature conditions. Transient simulations were conducted for the conceptual RCCS design for the HC-HTGR under various operation modes and heat load conditions using RELAP5-3D. The system dynamics in different operating states was investigated and the system performance under transients of interest was evaluated. The results demonstrated the overall system feasibility that the RCCS design maintains structures temperatures lower than maximum allowable temperature with sufficient system inventory without any active heat removal in the design process with certain transients addressed. The HC-HTGR RCCS will have additional design updates of subsystems or optimization of the system components during the preliminary and final design phases. Since the entire plant has not been integrated yet, this delivered conceptual design is subject to changes for integration, that require additional conceptual design activities and Quality and Assurance implementation (Q&A). The performance assessment of the RCCS for the HC-HTGR will be then revisited and optimized to finalize the system design, and the RCCS integrated primary system analysis will be utilized to simulate selective accident scenarios of interest where efforts are currently undergoing in the project.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Accelerated Irradiation and Qualification of Ceramic Nuclear Fuels

Accelerated irradiation efforts are an important topic in accelerated qualification of new nuclear fuel concepts for light water reactor (LWRs), microreactors, and other special purpose reactors. Accelerated fuel qualification combines both experimental and modeling work to expedite the total qualification time from approximately 20-25 years to 5-10 years timeframe. The scope covered under the current work package focuses on the experimental contributions for the neutron irradiations. Accelerated irradiation aims to reduce the total time needed for neutron irradiation to achieve targeted burnup, which is timely using conventional irradiation profiles. The purpose of this report is to detail the samples preparation for such irradiations, which contain standard, medium, and large grain undoped UO 2 , as well as large grain, Cr 2 O 3 -doped UO 2 . The data that will be generated from the irradiation of these samples will yield important information regarding fission gas retention as a function of grain size and the effect that Cr dopants have on diffusivity of fission gases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Better Climate Challenge: Greenhouse Gas Reporting Guidance

Better Climate Challenge Greenhouse Gas Reporting Guidance document can help you define the scope of your commitment, set your goal and base year, and annually report progress toward your goal for Better Climate Challenge Program. You will share GHG emissions and energy data across your building/plant portfolio with DOE, including owned and operated fleets. This document provides instructional guidance on how to quantify GHG emissions across your organization. Establishing an accurate GHG inventory is a critical first step to developing a plan to reduce emissions across your organization.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

Methane Pyrolysis for CO2-free H2 and Carbon Nanomaterials - CRADA 576 (Final Report)

In this CRADA project we continued to develop the Pacific Northwest National Laboratory (PNNL) patent-pending Regenerable Catalytic Pyrolysis (ReCaP) process technology for producing CO 2 -free hydrogen (H 2 ) from inexpensive and domestically-abundant natural gas (NG), while simultaneously reducing H 2 ’s net production cost to $\$$1.0/kg through the sale of valuable crystalline solid carbon co-product. Producing clean hydrogen at this price is a DOE Hydrogen Energy Earthshot goal. This effort builds on our prior catalyst and processing advances made thermocatalytic decomposition of methane (TCD). The additional scope performed here accelerated the commercial deployment of TCD for CO 2 -free H 2 and valuable solid carbon nanotubes (CNT) co-product, by i) scaling up the production of CNT co-product using a fluidized bed reactor (25 g catalyst scale versus the 1 g catalyst scale demonstrated prior), ii) producing approximately 1 kg of CNT byproduct, produced via multiple cycles of TCD, carbon-catalyst separation, and catalyst re-synthesis, to enable the production of sufficient quantities of solid carbon so as to explore its market potential, iii) understanding the quality of the co-product CNTs, produced at larger scale, using advanced characterization, and iv) beginning to explore multiple promising high volume carbon product applications (e.g., aluminum and polymer composites, steel additive, and cement reinforcement applications). We are in discussions with Department of Energy and potential commercial partners to continue funding of this effort with the goal to facilitate eventual commercial deployment.

08 HYDROGEN↗

Reliability of Materials and Components for Solid Oxide Fuel Cells

Planar stack solid-oxide fuel cells (SOFCs) require seals that must operate reliably under demanding conditions for lifetimes of 40000 hours. This includes temperature fluctuations between 800°C and RT during on and off cycles, thermal stresses, oxidizing environments and chemical degradation to name a few. This comprehensive report provides results from long term testing of two commercially available multicomponent barium alkali silicate glasses: SCN and G6, chosen as sealing candidates. In this scope, the glass seals were deposited on YSZ and Al 2 O 3 substrates simulating electrolytes (Zrbased) and coatings (both zirconia and Al 2 O 3 ). The seal-substrate couples were subjected to 800°C under air and steam+H 2 +N 2 environments up to 40000 hours to test their integrity under real operating conditions. Extensive studies on the effects of exposure have been conducted over the span of testing at various time intervals. Within the context of characterization, mechanical properties such as density, roughness, thermal expansion and glass transition, viscosity and wettability behavior; and microstructural properties such as glass chemistries, defect formation (cracks and pores), phase transformations (devitrification) and glass-interface reactions are investigated. Results and discussions are provided with a focus on the degradation of the properties over long term interrupted testing.

30 DIRECT ENERGY CONVERSION↗

Qualitative Conditions for Reasonable Estimates of Kerma Using Total and Kerma Cross Section Attenuation of Photons

Performing a full adjoint simulation in the Integrated Tiger Series (ITS) can be time consuming to obtain statistically significant results. The ray-trace capability in ITS allows for rapid scoping calculations of the uncollided kerma from photon sources without needing to run a full adjoint simulation. However, the uncollided estimate will always underestimate the full-physics kerma since it neglects scattered radiation, and under certain conditions, the result from the capability may not provide a sufficiently accurate estimate of the full-physics kerma. To exemplify the conditions in which the capability provides reasonable estimates, two problem geometries with different materials are simulated using the full adjoint capability as well as the ray-trace capability with a total cross section treatment and a new kerma-attenuation cross-section treatment. The results are then compared to show under which conditions the estimates are accurate. Reasonable estimates are provided from the ray-trace feature when there is minimal scattering into a region of interest occurring, such as with low-Z material and when the photons travel through small amounts of material.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Safeguards Technology Development Program: FY 2024 Mid-Year Report

This project has 2 major thrusts and 2 minor tasks – One of the major ones is a scoping study for improving uranium isotopics by analyzing higher energy, lower yield U-235 gamma emissions. The second major one benefits past projects such as auto-enrichment, and peak-based analysis routines previously added to GADRAS-DRF [1]. The two minor tasks involve an in-person or remote GADRAS-DRF training for the IAEA, and characterizing detectors of interest.

07 ISOTOPE AND RADIATION SOURCES↗

Condenser Enclosure Vibration Isolation Mount Selection

A mini split cooling system will be used to maintain temperature requirements in a mobile secure transport system. While the split cooling system was designed to be used in static residential or commercial applications, it was selected for this transportation application due to a unique set of security requirements. However, the system’s ability maintain reliability and survive prolonged long-term shock and vibration is a significant concern. The mitigation strategy is to select vibration isolation mounts and perform lifetime shock and vibration testing to demonstrate survivability. The goal of this study is to generate a finite element model of the system and perform modal analysis to inform selection of vibration mounts to minimize the amount of vibrational energy transferred to the split cooling system. The scope of this report is limited to study of the condensing unit only, and geometric variation of the assembly will not be allowed.

42 ENGINEERING↗

Sandia Mechanics Challenge 2023 Information Packet: Structure with a Threaded Fastener Joint

The Sandia Mechanics Challenge (SMC) provides the solid-mechanics community a forum for assessing its ability to predict mechanical behavior in structures and materials through a blind, round-robin format. Computationalists are asked to predict the behavior of an unfamiliar geometry given experimental calibration data, their predictions are compared to experimental measurements of the SMC scenario, and then the participants assess and compare their approaches, documenting their findings. The SMC broadens the scope of Sandia-hosted benchmarking problems that previously focused on ductile failure through the Sandia Fracture Challenges, enabling an enduring, community-wide self-assessment of predictive capabilities for a variety of mechanics topics. The SMC is part of the Structural Reliability Partnership, which offers other benchmarking challenges hosted by several participating institutions.

42 ENGINEERING↗