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At least 577 records · Page 32

Experimental and computational study of the microporous layer and hydrophobic treatment in the gas diffusion layer of a proton exchange membrane fuel cell

Enabling fuel cell operation at high current density is critical for developing competitive alternative power system to replace internal combustion engine. However, liquid water management continues to be a challenge for high humidity or high current density operation. Water condensation in the porous media hinders efficient oxygen transport to the catalyst layer, which in turns, reduces fuel cell performance. To improve water management capability, the gas diffusion layer is often impregnated with Polytetrafluoroethylene (PTFE) and coated with a thin microporous layer, which have shown to improve fuel cell performance, especially under wet conditions. However, the fundamental mechanism that drives the performance enhancement is still not well understood. In this work, the effects of PTFE impregnation and MPL were studied using both experimental and computational techniques. Both limiting current and polarization tests under dry and wet operating condition are conducted to study the oxygen transport resistance and fuel cell performance. In addition, a 2-D, two-phase, multi-physics PEMFC model is developed to simulate performance and gain a fundamental understanding of local water saturation and oxygen concentration. The combined results show that 5 wt% PTFE impregnation with MPL significantly enhances liquid water management, which enables higher current density operation of a fuel cell.

25 ENERGY STORAGE↗

Influence of hydrocracking and ionic liquid pretreatments on composition and properties of Arabidopsis thaliana wild type and CAD mutant lignins

Lignin is the primary contributor to the high cost of biofuel-production from lignocellulosic biomass. Here, in order to study lignin removal and the release of aromatic monomers, we applied hydrocracking and ionic liquid pretreatments on Arabidopsis thaliana biomass from both wild type (WT) and a mutant (CAD cxd) defective in two cinnamyl alcohol dehydrogenase genes involved in the lignin biosynthetic pathway. For Arabidopsis WT, our results highlight that pretreatments reduce average molecular weight of lignin by about 65% and decrease the content of β-O-4 linkages between lignin monomers. For Arabidopsis CAD mutant, an opposite effect is evidenced. Fewer differences were observed on depolymerization and molecular structure of lignin, which indicates that (8-O-4), (8-5), and (8-8) linkages observed in CAD mutant make lignin more resilient to pretreatment than wild-type lignin. Finally, our study shows the potential of hydrocracking pretreatment technology for extracting valuable aldehyde monomers such as vanillin and syringaldehyde from biomass.

09 BIOMASS FUELS↗

A generalized analytical energy balance model for evaluating agglomeration from a binary collision of wet particles

Agglomeration of wet particles, i.e., particles coated with a thin liquid layer, is a common phenomenon in many processes like fluidized bed combustion of low rank fuels. The availability of an agglomeration model that can evaluate the outcome of a binary collision between wet particles differing in solid particle properties, liquid layer thicknesses, and initial collision (impact) speeds is essential for obtaining a comprehensive understanding on the existing processes experiencing wet particle agglomeration or for a successful development of new processes with high chances of wet particle agglomeration. This study presents a generalized agglomeration model on the basis of energy conservation before and after collision when colliding wet particles may differ in solid particle properties, liquid layer thicknesses, and impact speeds. The model was established based on the approximate values of energy losses that may happen during the collision. It incorporates body forces, solid-solid contacting, liquid capillary, and viscous contributions, as well as the liquid bridge volume effect. Predictions of the new model for collision outcomes of identical wet particles were like those from an analytical energy balance model developed recently by the group for identical wet particles. We also validated the new model by experimental data from literature. The results of a collision direction analysis indicated that the direction often has a minimal effect on the collision outcome in many practical scenarios. The results of Monte Carlo uncertainty analyses with the new model revealed that proper estimations of impact speed, under capillary limiting conditions, and thickness of coating layers and asperity heights, under viscous limiting conditions, are critical for the realistic prediction of collision outcomes at impact speeds close to critical impact speed, i.e., the minimum particle speed required for the particles to rebound.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

High-Fidelity Numerical Investigation on Elucidating Sodium Heat Transfer Characteristics for 37-Pin Wire-Wrapped Fuel Bundle in the PLANDTL Facility

This study involved a Reynolds-averaged Navier-Stokes- (RANS-) based computational fluid dynamics (CFD) analysis of the 37-pin wire-wrapped fuel bundle of the PNC Plant dynamics test loop (PLANDTL) facility. Previously, mainly the hydrodynamic phenomena of the wire-wrapped fuel bundle were analyzed, but the present study additionally included heat transfer analysis through conjugate heat transfer. The main purpose of the study was to benchmark the experimental data of the PLANDTL 37-pin wire-wrapped fuel bundle to investigate the heat transfer phenomena. In addition, the aim was to verify the accuracy of the RANS-based CFD analysis method using the STAR-CCM+ simulation software in comparison with the experimental data. The grid used for verification was an innovative grid system consisting of hexahedra using Fortran-based code. The development of the RANS-based CFD methodology included grid sensitivity analysis, turbulence model sensitivity analysis, and turbulent Prandtl number sensitivity analysis. Information on the temperature, mass flow rate, and area of the CFD results for each subchannel was provided for the top of the heated section and is expected to serve as a reference for future studies aiming to perform the validation and verification of a PLANDTL facility. In addition, the dependence of the peak temperature on the azimuth angle of each pin was analyzed.

97 MATHEMATICS AND COMPUTING↗

Requalification and Declaration of High-alpha Cemented TRU-waste from the Late 70's and Early 80's - 20499

KTE and Siemens launched a project to qualify a series of 391 waste drums and 115 concrete containers produced from 1979 to 1984. The waste is characterized by a high content of long-lived alpha-nuclides (TRU-waste - Pu, U, Am in the range of 1 E+11-1 E+12 Bq/ drum. This high radionuclide inventory requires a special packaging according to the waste acceptance criteria (WAC) of the repository Konrad [1] despite the fact, that the waste was treated manually. The raw waste originally was delivered in nitric acid solution and was solidified using ordinary Portland cement. The existing documentation comprises analyses, transfer documentation for the nuclear fuel nuclides and waste treatment documentation yielding information, which waste was treated together in a batch. Additionally all drums were subjected to non-destructive gamma-spectrometry assay. A small number of 7 drums were opened and sampled to support the existing declaration. For the entire series the waste products were attributed to defined batches using the accountancy data supported by the analyses and the gamma-spectrometry. Each batch was produced from a homogeneous stock of waste solution and can therefore be declared and documented together. The declaration concept is based on the gamma spectrometry results performed on the entire drums, which were corrected and extended for the uncertainty of the analyses. The declaration concept is verified by the regulator and its independent expert. The state of the project is in the documentation of the solidification campaigns followed by the planning of the container configuration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Advanced spectrometric methods for characterizing bio-oils to enable refineries to reduce fuel carbon intensity during co-processing

A promising approach for supplementing petroleum-derived fuels to support reductions in green-house gas emissions is to convert abundant biomass feedstocks into renewable carbon-rich oils using pyrolysis. However, the resultant bio-oils contain various oxygenated compounds that can impart acidity, chemical and thermal instability, and immiscibility with petroleum derived fuels, necessitating further upgrading to derive fuel blendstocks. Co-processing bio-oils and petroleum-derived liquids in existing refineries is a potentially near-term, cost-effective approach for upgrading bio-oils while reducing refinery carbon intensities. However, one cause for hesitation in co-processing bio-oils is limited comprehensive characterization and speciation of the bio-oil components. Advanced analytical techniques are currently under investigation to enable identification of elusive species in bio-oils, enabling researchers to develop strategies to mitigate catalyst deactivation agents and contaminants. This review provides a brief overview of several analytical methods commonly used to analyze bio-oils and their limitations. In addition, advanced techniques currently under development are discussed to further elucidate bio-oil components that may limit its end use. This will help inform the technical and economic feasibility of co-processing bio-oils with petroleum-derived liquids, therefore, improving the overall downstream processes for biofuels blendstock production.

09 BIOMASS FUELS↗

Short Communication: Observation of Initial Burst Release of Fission Gas from High-Burnup UO 2 Nuclear Fuel During Thermal Transient

A system was developed and tested to provide a deeper understanding of the fission gas release kinetics from nuclear fuel during thermal transients. Pressure, temperature, spectral data, and optical images were simultaneously collected during resistive sample heating, and all released gases were collected in liquid nitrogen-cooled charcoal traps. Standup testing was performed with a high-burnup nuclear fuel segment from rod section with an average local burnup of 77 GWd/tU. During heating, the segment released approximately 6 ± 2% of generated fission gas inventory after ramp heating to 460°C. Here, heating ceased when the sample ejected from the holder, as observed by constant imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Selective and Stable Ethanol Synthesis via Electrochemical CO 2 Reduction in a Solid Electrolyte Reactor

Electrochemical CO 2 reduction to ethanol faces challenges such as low selectivity, a product mixture with liquid electrolyte, and poor catalyst/reactor stability. Here, we developed a grain-rich zinc-doped Cu 2 O precatalyst that presented a high ethanol Faradaic efficiency of over 40% under a current density of 350 mA·cm –2 . Our density functional theory (DFT) simulation suggested that Zn atoms inside the structure have a greater carbophilicity than the Cu atoms to help facilitate *CHCHO formation, a key reaction intermediate toward ethanol instead of other C 2 products. Here, a high Faradaic efficiency ratio between ethanol and ethylene (FE EtOH /FE C2H4 ) reached 2.34 in the zinc-doped Cu 2 O precatalyst, representing an over 4-fold improvement compared to bare Cu 2 O precatalyst. By integrating this Cu-based catalyst into a porous solid electrolyte (PSE) reactor with a salt-managing design, we achieved stable ethanol production for over 180 h under a current density of 250 mA·cm –2 while maintaining ethanol selectivity at ~30%.

09 BIOMASS FUELS↗

Computational model development and validation of fuel dispersal phenomena

As the US nuclear industry is renewing efforts to extend the rod average burnup limits in existing pressurized water reactors, it becomes imperative to investigate the safety implications associated with high burnup fuel fragmentation and the potential dispersal of fragmented fuel into the reactor coolant system. Additionally, as there is a growing interest in transitioning to new nuclear fuel designs to accommodate higher fuel burnup levels, it becomes essential to conduct studies on fuel behavior during design basis accident scenarios, particularly during a postulated loss-of-coolant accident, in order to ensure the safe operation of light-water reactors. The present study develops a computational model to simulate the complex three-phase flow of the fuel dispersal phenomena that occur following a breach in the cladding, when the stream of solid fuel particles and carrier fission gas stream may interact with the surrounding liquid medium. Furthermore, the proposed framework considers the solid phase as discrete Lagrangian particles while the interpenetrating gas–liquid continuum is modeled using a Eulerian framework. The developed simulation framework is validated with experimental results performed in a separate-effect test facility by comparing predicted particle settlement locations with the experimental observations reported in the literature. Key highlights of this study include the interpretation of the high-pressure boundary conditions and the three-phase flow coupling strategy. Also discussed are the gas–liquid hydrodynamics and the behavior of particles as they are transported through the gas and liquid phases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Update on Full-Scale Retrieval and Transfer System Testing for Pneumatic Conveyance of High-Level Calcine Waste - 20241

The Idaho National Laboratory processed spent nuclear fuel from 1953 to 1994. The process used solvent extraction to separate uranium, resulting in a liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive solid powder, referred to as calcine. The calcine is stored in stainless steel bins contained within concrete vaults called the Calcined Solids Storage Facilities. There are a total of seven CSSFs. The Calcine Retrieval Project is tasked with retrieving the calcine from the bins of CSSF 1 and pneumatically transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. Once the calcine is removed from CSSF1, that facility may be closed, reducing the footprint of waste storage at the INL. The same process for removing Calcine from CSSF1 may then be utilized to remove the calcine from the remaining CSSFs for final treatment and transport to a long-term storage facility. To overcome these inherent challenges with accessing CSSF 1 and extracting the calcine in a timely manner, a full-scale integrated mockup has been constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. All essential elements of the transport design are represented in the Mockup, including motive and makeup air supply (air compressor), a replica of the CSSF6 cyclone, pre-filter, HEPA filter, vacuum pump, and retrieval control devices. Construction on the Mockup was completed in 2018 and testing began in February 2019. It has already proven very beneficial to the CRP. Experience in running the Mockup have aided in the development of retrieval and transport methodologies and the selection of process parameter values. It has also uncovered some unforeseen technological challenges and will continue to be valuable to the testing team as these hurdles are overcome and the process is proven effective. Operations personnel have also benefitted from its use as a training tool and will continue to use the Mockup as a simulator to reproduce problems encountered during extraction and transfer of Calcine from CSSF 1. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Upgrades to the Thomson Scattering Diagnostic on the Lithium Tokamak eXperiment-Beta (LTX-$β$)

We describe the upgrades to the Thomson scattering (TS) diagnostic on the Lithium Tokamak eXperiment-beta (LTX- β ) to provide complete radial coverage of the plasma. LTX- β is a spherical tokamak with plasma limited at a major radius of R = 14 cm and R = 66 cm. Flat temperature profiles with peaked density were observed in LTX- β using an existing multipoint TS diagnostic with radial coverage of the plasma core. There is a need for more comprehensive core and edge TS measurements to thoroughly examine beam fueling, beam heating, and transport under low recycling conditions with liquid lithium walls and a lower collisionality scrape-off layer (SOL). The core TS set up provides plasma profiles from R = 40–60 cm. To extend the TS diagnostic coverage: 1) we added a TS viewing optics to cover the SOL region at five radial points from R = 57–64 cm, and 2) another TS viewing optics in the HFS region from R = 15–45 cm to obtain TS data from ten radial positions while making use of the original laser beam. The main focus of this article is on the new HFS TS setup that is equipped with better-matched collection optics located in the horizontal midplane where the laser polarization lies and views from a scattering angle of 130°–150° with respect to the laser path. A set of unequal-length Y-branched fiber bundles transports the TS light from two radial points to each polychromator. The new HFS view is better matched in spot size, scattering angle, and throughput to the laser and polychromators and has new V-grooved style viewing dumps. The new HFS TS view, along with the core and SOL TS views, will provide a full radial measurement coverage from R = 15–64 cm to explore the flat temperature profiles.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Corrosion Behavior Characterization of Tantalum in Flowing LBE [Slides]

Tantalum is a refractory metal, it has low coefficient thermal expansion and high creep resistance. Its BCC structure makes Ta less prone to radiation damage. It is 2nd most ductile among refractories. Its oxide (Ta 2 O 5 ) layer is thin and stable at high temperatures and low oxygen concentrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Nuclear Waste Attributes of SMRs Scheduled for Near-Term Deployment

The purpose of this study is to evaluate the nuclear waste attributes of Small Modular Reactors (SMRs) scheduled for deployment within this decade using available data and established nuclear waste metrics, with the results compared to a reference large Pressurized Water Reactor (PWR). The current fleet of commercial nuclear reactors in the U.S. is composed of 92 large Light Water Reactors (LWR) with an average electricity generating capacity of over 1,000 MWe each. These large LWRs built on-site in massive construction projects have been the mainstay of the industry for the last 50 years. However, new construction soon is expected to include several designs of smaller reactors primarily fabricated in factories and installed in the field in modules. Some of these SMRs will also be LWRs, while some will use other coolants such as liquid metals, molten salts or gases, and different types of fuels. The technologies and economics of SMRs have been the focus of many studies, but there has been only minimal information published on the amount of nuclear waste different types of SMRs are expected to generate and no reports focused on near-term-deployable designs. In this study, the nuclear waste attributes of three small reactors scheduled for near-term-deployment, VOYGR TM (from NuScale Power), Natrium TM ab (from TerraPower), and Xe-100 (from X-energy), were assessed by comparing nuclear waste metrics with those of a reference large Pressurized Water Reactor (PWR).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

MSTEC: Molten Salt Thermophysical Examination Capability

Description – shielded modular hotcell with an inert argon atmosphere, housing characterization equipment for determining thermophysical and thermochemical properties of high temperature liquids not limited to but focusing on TRU and irradiated fuel salts

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Status of Multiple Channel Fuel Performance Capabilities Within the SAS4A/SASSYS-1 Safety Analysis Software

SAS4A/SASSYS-1 (SAS) is a fast-running simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. It is a critical element of safety analysis capabilities for the U.S. Department of Energy and is utilized within industry to perform the transient safety analyses required to support the licensing of Liquid Metal-cooled Fast Reactors (LMFRs). Although SAS is exceptionally fast for most transient scenarios, fuel performance calculations, along with the associated pre-transient characterization of the fuel pin, may be required for transient scenarios where fuel pin failure is hypothesized. Both the pre-transient characterization and the transient fuel performance calculation are necessary to properly quantify margins to potential fuel failure and assess the time spent potentially exceeding such margins during events. While safety analysis calculations with fuel performance models provide a more detailed characterization of the reactor during a transient, the pre-transient characterization can be time-consuming and computationally expensive. Often, large numbers of fuel pins have been exposed to similar pre-transient irradiation conditions. Similarly, the same pre-transient fuel characterization may be applicable to numerous transient conditions. This provides an opportunity to optimize the SAS computational framework such that pre-transient fuel characterization can be shared across multiple channels (fuel pins) and across multiple simulations, thus dramatically reducing overall computational costs. This report summarizes progress toward enhancing the SAS computational framework to support shared, multiple channel fuel performance characterizations intended to significantly reduce computational costs. Preliminary testing has shown that the computational time saved by using the pre-transient sharing capability is approximately equal to the time it takes to perform the pre-transient characterization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Nuclear Safety [Vol. 33, No. 4, October-December 1992]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 499 Technical Note: Nuclear Regulatory Commission’s Examination Question Bank System, C. L Jensen and N. L. Skinner; 506 Report on the Specialists’ Meeting on Passive and Active Safety Features of Liquid-Metal Fast Breeder Reactors Organized by the International Atomic Energy Agency at Oarai Engineering Centre of Power Reactor and Nuclear Development Corporation, Japan, November 5-7,1991, S. R. Paranjpe; ACCIDENT ANALYSIS: 514 A Review of Hydrogen Production During Melt/Water Interaction in LWRs, D. F. Fletcher, B. D. Turland, and S. P. A. Lawrence; 535 A Review of the U 3 0 8 -Al Reaction as a Potential Heat Source in Research and Test Reactor Accidents, J. L. Snelgrove; CONTROL AND INSTRUMENTATION: 543 Multilevel Interfaces for Power Plant Control Rooms II: A Preliminary Design Space, K. J. Vicente; DESIGN FEATURES: 549 Twenty-Second DOE/NRC Nuclear Air Cleaning and Treatment Conference, R. R. Bellamy, D. W. Moeller, and M. W. First; 562 Book Review: Safety Assessment of Unit 5 (WWER-440/W-213) of the Greifswald Nuclear Power Station, M. H. Fontana; WASTE AND SPENT FUEL MANAGEMENT: 576 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 586 Analysis of Corrective Actions Applied to Nuclear Power Plant Operations, A. E. Cross; 593 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 596 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; 609 Non-Commercial Power Operating Experience; 614 World Nuclear Performance, Compiled by M. D. Muhlheim; RECENT DEVELOPMENTS: 623 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 635 Reports, Standards, and Safety Guides, D. S. Queener; 639 Proposed Rule Changes as of June 30, 1992; ANNOUNCEMENTS: 513 Fourth International Conference on Nuclear Fuel Reprocessing and Waste Management RECOD '94; 585 International Topical Meeting on Advanced Reactors Safety; 622 International Workshop on Implementation of ALARA at Nuclear Power Plants; 648 Short Course on Nuclear Criticality Safety; 648 8th Topical Meeting on Problems in Reactor Physics MEPhi-VOLGA-93; 643 The Authors; 646 Reviewers of Nuclear Safety, Vol. 33

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Structure of the polysaccharide sheath from the B race of the green microalga Botryococcus braunii

The green microalga Botryococcus braunii forms a colony of cells held together by an intricate extracellular matrix. This extracellular matrix also acts as a storage location for liquid hydrocarbons initially produced inside the cells. These hydrocarbons can be used as feedstocks to produce transportation fuels. In the B race of B. braunii, the extracellular matrix is composed of several elements: a cross-linked hydrocarbon network filled with the liquid hydrocarbons, a retaining wall to hold the hydrocarbons in the extracellular matrix, an arabinogalactan polysaccharide fibrillar sheath that extends outward from the retaining wall, and a single protein known as polysaccharide associated protein is found at the base of the polysaccharide fibers where they connect to the retaining wall. During growth, colonies of the B race of B. braunii shed fragments of the retaining wall with polysaccharide fibrils attached into the media. Here we describe the isolation and structural analysis of these polysaccharides using a series of chemical degradations in combination with glycosyl residue composition and linkage analysis, as well as 2-dimensional nuclear magnetic resonance spectroscopy. The results demonstrate that the polysaccharide is highly branched, consisting of a galactan backbone with short branches comprised of Dgalactopyranose, L-arabinofuranose, 4-O-methyl-D-glucopyranuronic acid, and 6-deoxy-D-altropyranose. To our knowledge, this is the first description of 6-deoxyaltrose in algae or plants.

59 BASIC BIOLOGICAL SCIENCES↗