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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 1,387 records · Page 77
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to the qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from measurements is of the order of 2.5 W/g at half of the possible MARIA power - 15 MW. The approach and the detailed program for experimental verification of calculations will be presented. The following points will be discussed: - Development of a gamma heating model of MARIA reactor with TRIPOLI 4 (coupled neutron-photon mode) and APOLLO2 model taking into account the key parameters like: configuration of the core, experimental loading, control rod location, reactor power, fuel depletion); - Design of specific measurement tools for MARIA experiments including for instance a new single-cell calorimeter called KAROLINA calorimeter; - MARIA experimental program description and a preliminary analysis of results; - Comparison of calculations for JHR and MARIA cores with experimental verification analysis, calculation behavior and n-γ 'environments'. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2019

Evaluation Studies of a 800W Solid Oxide-Based Fuel Cells Stack for Electrical Power in Aviation

As both NASA and the aeronautics industry recognize the need for higher fuel efficiency and lower carbon emissions in both commercial airline and private aviation applications, development of all-electric or hybrid electric aircraft have garnered renewed interest in the aviation community. For the particular example of the hybrid-electric option, the solid oxide fuel cell (SOFC) is an attractive option for the power source, due to its potential to utilize aviation fuels thereby having minimal impact to aviation infrastructure. SOFC stack performance depends upon many factors, one of the most important is the way the oxidant and fuel gases are delivered to the fuel cells. System modeling of various aircraft configurations for FUELEAP (Fostering Ultra-Efficient, Low-Emitting Aviation Power) point to the need to operate SOFC stacks at high current densities. This creates challenges in the thermal profile of the stacks with potential to create large thermal gradients and hot spots. This study investigates two types of commercial solid oxide fuel cell stacks, the cross flow and co-flow gas designs, both convectively cooled with cathode air. High fuel utilization factors were also employed under varying electrical loads expected from the demands of flight. In addition, performance, range of operation and endurance were investigated under conditions of high current loads and thermal cycling. Evaluations include the study of gas kinetic using electrochemical spectroscopy. Testing took place at the facilities of NASA Glenn using a commercial test system (FuelCon AG, Magdeburg Germany). These studies are crucial to the Glenn Research Center's ability to conduct research, evaluation and development of the next-generation SOFC based stacks for cutting-edge energy technologies for aerospace applications. This study supports NASA's Convergent Aeronautics Solutions' (CAS) FUELEAP project.

Goldsby, Jon C.↗
DOE OSTI2021

Isobutanol Octane Blending Model with Gasoline: Cooperative Research and Development Final Report, CRADA Number CRD-17-00689

The purpose of this agreement between the National Renewable Energy Laboratory (NREL) and Gevo, under the DOE Small Business Vouchers Pilot, is to develop a predictive octane blending model for isobutanol and blendstocks for oxygenated blending (BOBs). Validated models of ethanol octane blending effects are currently used in refinery blending models. While it is known that isobutanol increases octane when blended into BOBs, the effect is non-linear, and dependent on hydrocarbon blendstock properties. Because no predictive model currently exists for isobutanol and BOBs, blenders must perform expensive and time-consuming tests to determine the octane effect on each batch of finished fuel. NREL will make basic measurements of research octane number (RON) and motor octane number (MON) of finished gasoline-isobutanol blends prepared in a broad range of BOBs. The key properties and detailed composition of BOBs utilized will be measured by NREL. Using these measured fuel properties, NREL will develop a predictive octane blending model for isobutanol in BOBs that can be used by terminals to further reduce barriers to market penetration of isobutanol.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Biochemical Conversion of Lignocellulosic Biomass to Hydrocarbon Fuels and Products (2021 State of Technology and Future Research)

The annual State of Technology (SOT) assessment is an essential activity for biochemical platform research. It allows the impact of research progress to be quantified in terms of economic improvements in the overall cellulosic biofuel production process for a particular conversion pathway. As such, initial benchmarks can be established for currently demonstrated performance and progress can be tracked towards out-year goals to ultimately demonstrate cost-competitive cellulosic biofuel technology. The purpose of this report is to benchmark the latest experimental developments across a number of potential bioconversion pathways as quantified by modeled minimum fuel selling prices (MFSPs), as a measure of current status relative to those final targets. For this state of technology, TEA models were run for two separate biological conversion pathways to fuels, based on available data for integrated biomass deconstruction and hydrolysate processing; namely carboxylic acids (primarily butyric acid) and diols (2,3-butanediol [BDO]), reflecting NREL's recently-published 2018 biochemical design report focused on those two pathways. The models were run across three scenarios for lignin utilization, namely combustion, conversion to coproducts based on "base case" performance with biomass hydrolysate, and conversion to coproducts based on "high" performance demonstrated with model lignin monomer components. A key improvement reflected in the 2021 SOT is centered around making use of the latest lignin conversion data, which over the past year focused primarily on production of ß-ketoadipate (BKA) as a more optimal molecule compared to the closely-related adipic acid coproduct of prior recent focus, both in terms of superior product properties and biology, as well as reduced processing complexity (reducing two steps for sequential production of muconate followed by hydrogenation to adipic acid down to a single step for direct production of BKA). This update translated to a roughly 17% increase in mass yield of final coproduct output at a nearly four-fold increase in fermentation productivity on lignin monomers relative to prior 2020 SOT benchmarks for muconic/adipic acid production.

09 BIOMASS FUELS↗
DOE OSTI2025

Direct numerical simulations for hybrid rocket boundary layers: Performance modeling and scaling

This paper presents a comprehensive performance and scaling analysis of direct numerical simulations for reacting boundary layers, focusing on slab burner configurations. Using a PETSc-based finite volume CFD framework, the study evaluates the scalability and computational cost of flow, chemistry, and radiation evaluations across 2D and 3D simulations. Polymethyl methacrylate (PMMA) is the fuel with pure O 2 as the oxidizer, modeled using a detailed chemical kinetics mechanism with 113 species and 660 reactions. A ray-tracing-based radiation solver, designed for distributed memory applications, is implemented to model radiation heat transfer. Parallel scalability is analyzed for the coupled flow, chemistry, and radiation heat transfer processes. Weak and strong scaling studies are conducted on up to 15,000 computational ranks, revealing robust performance when flow cells exceed 200 per rank. Chemistry evaluations dominate the computational cost in large 3D simulations, accounting for approximately 40% of the total runtime, while flow processes contribute around 35%, and radiation solver contributions remain below 10% due to reduced evaluation frequencies. GPU accelerated chemistry evaluation, implemented with Zero-RK, demonstrates significant promise, achieving up to a 4x speedup for workloads exceeding 30,000 cells per GPU. However, diminishing returns are observed for smaller workloads due to CPU-GPU communication overhead. This study identifies key challenges, including memory bottlenecks and the effects of domain partitioning on flow scalability, while highlighting the potential of GPU-accelerated chemistry to reduce computational costs. In conclusion, these findings provide realizable run configurations for 2D, 3D, and GPU-accelerated cases, offering insights for optimizing reactive flow solvers.

CFD Scalability↗
NASA NTRS1995

Adaptive Optimization of Aircraft Engine Performance Using Neural Networks

Preliminary results are presented on the development of an adaptive neural network based control algorithm to enhance aircraft engine performance. This work builds upon a previous National Aeronautics and Space Administration (NASA) effort known as Performance Seeking Control (PSC). PSC is an adaptive control algorithm which contains a model of the aircraft's propulsion system which is updated on-line to match the operation of the aircraft's actual propulsion system. Information from the on-line model is used to adapt the control system during flight to allow optimal operation of the aircraft's propulsion system (inlet, engine, and nozzle) to improve aircraft engine performance without compromising reliability or operability. Performance Seeking Control has been shown to yield reductions in fuel flow, increases in thrust, and reductions in engine fan turbine inlet temperature. The neural network based adaptive control, like PSC, will contain a model of the propulsion system which will be used to calculate optimal control commands on-line. Hopes are that it will be able to provide some additional benefits above and beyond those of PSC. The PSC algorithm is computationally intensive, it is valid only at near steady-state flight conditions, and it has no way to adapt or learn on-line. These issues are being addressed in the development of the optimal neural controller. Specialized neural network processing hardware is being developed to run the software, the algorithm will be valid at steady-state and transient conditions, and will take advantage of the on-line learning capability of neural networks. Future plans include testing the neural network software and hardware prototype against an aircraft engine simulation. In this paper, the proposed neural network software and hardware is described and preliminary neural network training results are presented.

Simon, Donald L.↗
DOE OSTI2025

Initial steady-state core simulation capability or thermal and pool-type molten salt reactors, coupling reactor physics, thermal-hydraulics, and evolving chemistry

This report presents the development and validation of an initial steady-state multiphysics capability for molten salt reactors (MSRs) under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in Fiscal Year 2025. The framework integrates neutronics, thermal-hydraulics, species transport, and thermochemistry to capture the coupled dynamics of liquid-fueled systems. Implementation and testing were performed on two representative designs: the Molten Salt Reactor Experiment (MSRE), a thermal-spectrum, channeled-flow reactor, and the Lotus Molten Salt Reactor (L-MSR), a fast-spectrum, pool-type reactor. The modeling suite employs Griffin for reactor physics and depletion, Pronghorn and SAM for thermal-hydraulics, Thermochimica for chemistry, and Saline for thermophysical properties, with benchmarking and validation carried out against historical MSRE data, experimental flow-loop measurements, and reference depletion calculations from Monte Carlo codes. The framework demonstrated the ability to reproduce key reactor behaviors including temperature feedback, reactivity losses, delayed neutron precursor transport, xenon poisoning, and redox potential evolution. The results confirm the feasibility and accuracy of the coupled models in predicting steady-state and selected transient MSR behaviors. This latter ones are used in this report as a proxy indicating that the steady-state models from which the transient starts are accurate. For MSRE, validation showed good agreement with pump start-up and natural circulation tests, while for the L-MSR, benchmarking confirmed hydraulic calibration and consistency of neutronics–thermal coupling. The tools also provided new insights into species transport, noble metal deposition, and salt solidification dynamics. On the Xenon transport front, the code is validated against the steady state Xenon poisoining measurement and showed good agreement with the experimental value. Identified areas for future work include advanced void transport modeling, three-dimensional simulations, improved alloy corrosion models, and tighter integration with high-fidelity Monte Carlo codes. These developments provide a foundation for high-fidelity MSR simulations that can support reactor design optimization, safety assessments, and long-term operational strategies.

42 - ENGINEERING↗
DOE OSTI2016

Phased Development of Accident Tolerant Fue

The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modeling predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology readiness levels (TRL) defined for nuclear fuel development. TRL 1–3 corresponds to the “proof-of-concept” stage (Phase 1), TRL 4–6 to “proof-of-principle” (Phase 2), and TRL 7–9 to “proof-of-performance” (Phase 3). This paper will provide an overview of the anticipated activities within each phase of development and will provide an update on the current ATF development status.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Phased development of accident tolerant fuel

The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modeling predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology readiness levels (TRL) defined for nuclear fuel development. TRL 1-3 corresponds to the 'proof-of-concept' stage (Phase 1), TRL 4-6 to 'proof-of-principle' (Phase 2), and TRL 7-9 to 'proof-of-performance' (Phase 3). This paper will provide an overview of the anticipated activities within each phase of development and will provide an update on the current ATF development status. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Time-Dependent Sensitivity Analysis of OECD Benchmark using BISON and RAVEN

Developments in uncertainty quantification for nuclear simulations have decreased the computational cost required to perform accurate sensitivity analysis. This work demonstrates implementation of these methods in the RAVEN framework, allowing for time-dependent sensitivity analysis. By demonstration we consider an OECD benchmark case. We propagate uncertainties using RAVEN operating on the BISON fuels performance code. We then consider the time-evolution of output response sensitivity to uncertain inputs. We perform sensitivity analysis using time-based stochastic collocation for generalized polynomial chaos (SCgPC) and high-dimension model reduction (HDMR). While uncertainty quantification applied to end-of-cycle statistics is useful, we demonstrate that dominant sources of uncertainty for figures of merit may fluctuate throughout the simulation. Time-dependent sensitivity analysis allows more insight on key design parameters for fuel. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Application of SPH Factors to PHWR Lattice Homogenization

Pressurized Heavy-Water Reactors (PHWRs) are heavy-water moderated and cooled. They consist of a horizontal, cylindrical, non-pressurized calandria vessel which contains the heavy-water moderator and is penetrated axially by fuel channels. Fuel channels consist of two concentric tubes separated by a gas gap: an inner tube called the pressure tube, and an outer tube called the calandria tube. The pressure tubes hold the fuel bundles, which are cooled by the flow of coolant at high temperature and pressure. The coolant flows in opposite directions in adjacent channels. The gap between the pressure and calandria tubes ensures that the temperature of the moderator remains relatively low, below 80 deg. C. For a CANDU 6 reactor, there are 380 fuel channels, each holding 12 37-element fuel bundles, approximately 10-cm in diameter and 50-cm long each. The fuel pins are arranged in three concentric rings (consisting of 6, 12 and 18 pins respectively), plus a central pin. The distance between channels (lattice pitch) is approximately 28.6 cm. Production core-level neutronics calculations are usually performed in a Cartesian geometry using few group diffusion and a simplified core model, whereby neutronics properties are homogenized over large regions called nodes. For Light Water reactors (LWRs), the two-dimensional axial (x-y) projection of a node corresponds to one fuel assembly. For PHWRs, a node usually consists of one fuel bundle and its associated moderator. For LWRs, advanced core-level calculations based on advanced node-level homogenization methods, such as Generalized Equivalence Theory (GET) or Super-homogenisation (SPH), improve the accuracy of full-core calculations compared to - standard homogenization. For rectangular fuel-pin arrangements, SPH can be used to perform pin-cell level homogenization which allows subsequent core-level diffusion calculations to be performed using fine Cartesian meshes. For PHWRs, the distribution of fuel pins in concentric rings makes the use of Cartesian mesh refinement impractical. However, SPH can still be applied, albeit using a coarser node subdivision, by dividing the heterogeneous node into 3x3 sub-regions in such a way that the central sub-region includes the entire fuel channel. This work investigates the use of SPH homogenization using a 3x3 subdivision of the PHWR node. Preliminary results show that the use of SPH factors does not yield more accurate results than standard homogenization when applied to PHWR nodes using a 3x3 subdivision. The explanation appears to lie in the fact that SPH factors are dependent on node-boundary leakage. Future investigations to confirm the leakage-dependence of SPH factors are recommended. Accuracy improvements may also be achievable by using smaller SPH regions at the periphery of the PHWR node, in an attempt to reduce the SPH factors' dependence on node-boundary leakage. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Full Integration of the Fission Product Release Module STACY into the HTR Code Package (HCP) - 18661

The HTR Code Package is under development since 2010 at Forschungszentrum Juelich and at Aachen University. The HCP combines the capabilities of individual legacy codes developed during the German HTR program in a highly integrated manner and contains new features. Within the HCP, all aspects related to fission product release and the redistribution within the primary circuit are handled by the module STACY (Source Term Analysis Code system). STACY is the successor of the codes FRESCO-I/II (dealing with fission product release) and PANAMA (dealing with fuel performance issues). Compared to these predecessor codes, STACY can take amongst others a time-dependent nuclide inventory and a radial temperature profile within the fuel element into account. So far, this version of STACY performs an individual fission product release calculation for a representative number of tracer pebbles. While this version of STACY still relies on codes like VSOP and MGT-3D and contains features such as basic fuel management which are also contained in the before mentioned codes, STACY now will be coupled to forms an integral part of the HCP. Thus, fission product release calculations will be examined with one consistent reactor input model. This allows assessing the release of fission products already during the design phase, besides parameters like the maximum fuel temperature during a transient. Furthermore the total nuclide inventory of a fuel element is distributed among fuel kernels, coating layers, graphite grains and pores of the surrounding matrix graphite. Formerly, this total inventory calculated by the burnup code TNT was distributed according to the distribution of Uranium. As a new feature, STACY now keeps track of the distributions of nuclide inventories within the fuel element. Especially for nuclides mainly formed by activation and decay processes like Ag-110m, this approach is more accurate. Further, the knowledge about the actual nuclide inventories improves the accuracy of fuel performance calculations. In addition, the internal coupling allows a feedback from the fission product calculation to other code modules. Nuclides amounts being released can be considered in e.g. the burnup calculation while calculating the source term of for example tritium or within the calculation which takes care of the redistribution of species by the coolant. Within this paper, the application of the STACY to HTR cores with special emphasis on fission product release to reactor models such as the HTR-Modul and HTR-10 will be discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Bioenergy sorghum crop model predicts VPD-limited transpiration traits enhance biomass yield in water-limited environments

Bioenergy sorghum is targeted for production in water-limited annual cropland therefore traits that improve plant water capture, water use efficiency, and resilience to water deficit are necessary to maximize productivity. A crop modeling framework, APSIM, was adapted to predict the growth and biomass yield of energy sorghum and to identify potentially useful traits for crop improvement. APSIM simulations of energy sorghum development and biomass accumulation replicated results from field experiments across multiple years, patterns of rainfall, and irrigation schemes. Modeling showed that energy sorghum’s long duration of vegetative growth increased water capture and biomass yield by ~30% compared to short season crops in a water-limited production region. Additionally, APSIM was extended to enable modeling of VPD-limited transpiration traits that reduce crop water use under high vapor pressure deficits (VPDs). The response of transpiration rate to increasing VPD was modeled as a linear response until a VPD threshold was reached, at which the slope of the response decreases, representing a range of responses to VPD observed in sorghum germplasm. Simulation results indicated that the VPD-limited transpiration trait is most beneficial in hot and dry regions of production where crops are exposed to extended periods without rainfall during the season or to a terminal drought. In these environments, slower but more efficient transpiration increases biomass yield and prevents or delays the exhaustion of soil water and onset of leaf senescence. The VPD-limited transpiration responses observed in sorghum germplasm increased biomass accumulation by 20% in years with lower summer rainfall, and the ability to drastically reduce transpiration under high VPD conditions could increase biomass by 6% on average across all years. This work indicates that the productivity and resilience of bioenergy sorghum grown in water-limited environments could be further enhanced by development of genotypes with optimized VPD-limited transpiration traits and deployment of these crops in water limited growing environments. As a result, the energy sorghum model and VPD-limited transpiration trait implementation aremade available to simulate performance in other target environments.

09 BIOMASS FUELS↗
DOE OSTI2020

BISON High Burnup Structure Modeling Capabilities Validated with a Selection of the Halden IFA-650 Rods

The U.S. Department of Energy (DOE)’s Nuclear Energy Advanced Modeling and Simulation (NEAMS)program aims to develop predictive capabilities using computational methods for the analysis and design of advanced reactor and fuel cycle systems. This program has been supporting the development of BISON, a high-fidelity and high-resolution fuel performance tool at the engineering scale. This document continues analysis and refinement of capabilities added to BISON early this calendar year in regards to the incorporation of capabilities applicable to extended burnups in response to industry interest. Details are provided on high burnup thermal conductivity models, a refitting of the high-burnup structure (HBS) porosity formation model to include additional data, the coupling of the HBS volume fraction model to thermal conductivity and fine fragmentation models, and validation activities. The IFA-650.4 and IFA-650.9 loss of coolant accident (LOCA) analyses are revisited with the latest developments in this report. A new validation case, IFA-650.14 has also been added to the BISON test suite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Ethanol to diesel: a sustainable alternative for the heavy-duty transportation sector

The combustion of middle distillates (diesel and jet fuel) is responsible for the emission of more than 2 GTon of CO 2 per year worldwide. While sustainable alternatives exist for gasoline and jet fuel, we still lack sustainable alternatives for diesel. This fact is especially relevant if we consider that electrification of sectors where diesel is used is challenging. One sustainable approach for diesel production is the catalytic upgrading of ethanol. While most work in this field has focused on the dehydration/oligomerization of ethanol, this approach is limited to producing fuels that have a high degree of branching and low cetane number. Another approach is the sequential use of ethanol Guerbet coupling, leading to higher alcohols, followed by etherification, leading to large ethers which, importantly, result in a product with high cetane number. In this work, we explore the catalytic upgrading of ethanol into diesel following an approach based on the initial Guerbet coupling of ethanol followed by etherification. The results presented are a collaborative and synergic effort among process and systems engineers, experimentalist in the area of catalysis, and fuel property modelers. Here, we demonstrate experimentally the feasibility of upgrading ethanol into a diesel fuel with properties that surpass its fossil counterpart. The diesel produced has a predicted cetane number of ~70 and outstanding cold flow properties, while maintaining other properties (viscosity, density, and flash point) within expected ranges. A technoeconomic analysis performed based on a detailed biorefinery model shows that the MFSP is ~5.89$ per Gal in 2021 dollars when lignocellulosic ethanol is used, with the most relevant economic driver the cost of the ethanol feedstock. The upgrading process can be performed with a net energy gain (EROI = 1.49 > 1). An LCA analysis of greenhouse gas (GHG) emissions reveals that the use of lignocellulosic ethanol may lead to more than 50% reduction in GHG emissions compared to fossil diesel. Depending on the CO 2 emissions associated with the production of ethanol we show that in some instances the production of carbon neutral diesel fuel is possible.

09 BIOMASS FUELS↗
DOE OSTI2022

Using Calibrated Water Data for Preliminary Validation of the SRT Code for Advanced Reactors

Recent interest and corresponding progress worldwide regarding advanced nuclear reactors has renewed focus on their performance and related safety assessments. Specifically, the U.S. Nuclear Regulatory Commission has emphasized the importance of mechanistic approaches to source term analysis for advanced reactor licensing applications, which attempt to realistically account radionuclide transport and retention phenomena. Further model development is required due to the numerous and complex physical and chemical phenomena associated with mechanistic source term analyses. Reflecting the need for modeling advancement, Argonne National Laboratory developed a mechanistic source term analysis tool for sodium fast reactors. The Simplified Radionuclide Transport (SRT) code describes fuel pin failure (for simulating the initial condition at the point of fuel pin breach), bubble scrubbing, deposition, leakage and following environmental impact. In the current work, a validation study of the SRT bubble scrubbing model is performed using a water-loop experiment performed at the University of Wisconsin-Madison. Through the analysis, the approach and fundamental bubble scrubbing models in SRT, which examine the removal of aerosols within the bubble as it is transported through a pool, have been widely evaluated. The results of the assessment demonstrate a high level of agreement in the regions of greater aerosol size. In the parameter range of minimal aerosol removal, the simulation slightly underpredicts the experiment results; however, considering the scale of plots and huge uncertainties inherently included, the deviation can be judged to be minor and would produce a conservative result. In addition, uncertainty analysis has been further refined to reflect the experimental distribution of parameters including aerosol sizes, which induces a span of performance for each representative aerosol size. Based upon the initial validation results along with uncertainty effects, SRT is expected to provide meaningful insights for the analysis of bubble scrubbing. Future sodium-loop tests will provide further validation basis.

Kam, Dong Hoon↗
DOE OSTI2020

Impact of Grain Size on Performance Degradation of TREAT LEU

We argue that radiation damage induced degradation of thermal conductivity does not set a lower limit on fuel grain sizes for the low enriched uranium fuel design of the Transient Reactor Test Facility (TREAT). Earlier work reports that smaller grains cause a larger degradation of thermal conductivity than larger grains constraining the smallest feasible size of fuel grains. This work assesses TREAT’s transient performance in the presence of radiation damage. The difference between the two studies is in treating damaged and fresh graphite as serial (this work) or parallel (previous) heat resistors. We use a multiphysics model of TREAT fuel grains to compute the reduction in transient capability measured by the total deposited energy as a function of irradiation dose. We find that radiation damage has a negligible effect on energy deposition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗