SEARCH · Engineering Papers

Results for “Fuels”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,747 records · Page 97

All sources NASA NTRS DOE OSTI

DOE OSTI2016

Dissolution flowsheet for high flux isotope reactor fuel

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3 O 8 ) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2 . The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2 ) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg concentration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Sustainable Aviation Fuel Blending and Logistics

Worldwide, aviation accounts for 2% of all manmade carbon dioxide emissions and 12% of all transportation CO2 emissions. In 2023, the U.S. accounted for 27% of the world jet fuel consumption. The aviation industry has set sustainability goals, and mandates. Sustainable aviation fuel (SAF), made from nonpetroleum feedstocks, significantly reduces aviation emissions. SAF must be blended with petroleum-based jet fuel prior to its use in aircraft. Jet fuel quality standards and certification documents are essential to fuel performance, operability, and safety and are the primary driver in determining locations for blending. Several locations were evaluated including terminals, airports, refineries, and greenfield/brownfield sites. While all the locations evaluated are technically capable of blending fuel, there are practical considerations that make terminals the optimal location for blending. The report includes background on jet fuel and SAF use, fuel quality standards, modes of transport for both conventional jet fuel and SAF, and terminal information.

09 BIOMASS FUELS↗

DOE OSTI2015

In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel verification technique.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Dry Pretreatment of Used Nuclear Fuel for Tritium Removal: A Review

Virtually all the elements in the periodic chart are generated as a result of atomic fission in nuclear fuel. Several of the elements are volatile or can readily react with other chemicals to form volatile molecules that can be released during the processing of used nuclear fuel for the recovery of the valuable products. Among the volatile fission products, {sup 129}I, {sup 14}C, {sup 3}H, {sup 85}Kr, and {sup 79}Se constitute the main environmental concern for uncontrolled release. Storage of used fuel allows the short-lived isotopes to decay significantly (e.g., {sup 3}H decays in half every 12.32 years and {sup 85}Kr every 10.76 years). However, depending on regulatory limits they may remain an environmental concern up to about 50-100 years of fuel storage. In contrast, the concern for environmental release of the highly mobile {sup 129}I and {sup 14}C remains for geological times (half-lives of 15.7 million years for {sup 129}I and 5730 years for {sup 14}C). Processing of used commercial nuclear fuel requires significant head-end operations to remove the fuel from the protective coverings (e.g., assemblies). The current industrial process includes cropping the end caps of the fuel assemblies, chopping of the fuel pins and/or bundles into short segments, and prolonged leaching of the segments using concentrated nitric acid (HNO{sub 3}). The infrastructure, processes, and waste treatment systems in the head-end operations, including the associated off-gas system, are some of the larger and more costly components of the entire plant. The bulk of the volatile fission products are released to the off-gas during the leaching and dissolution. However, a small but significant amount will transfer to the following steps of processing such as solvent extraction and will show up in many different waste streams. The effective capture of {sup 3}H and iodine throughout the different processes and disposition of associated waste streams is costly and complicated. Up-front removal and capture is highly desirable. Currently, at the U.S. Department of Energy national laboratories and internationally, several dry pretreatment technologies are being studied and developed with the potential to simplify and increase the safety of subsequent storage or chemical recycling processes. These studies include (1) alternative decladding methods, (2) oxidation of exposed fuel following either conventional shearing or chemical decladding, and (3) the collection and stabilization of released volatile radionuclides. Several dry processes are being developed for the up-front removal of tritium and other fission products. They can be combined with different decladding alternatives and, in general, produce a fine powder that can be fed to aqueous or pyrochemical separations processes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Investigation of Imaging Spent Nuclear Fuel Dry Casks using Cosmic Ray Muons

Recently, the use of cosmic ray muons for cargo scanning applications has been demonstrated and muons have been shown to have the potential to allow for nondestructive assessment of nuclear material accountancy. Relativistic muons have the ability to penetrate dense materials and monitoring the subsequent scattering of muons can provide information about the structure and composition of materials stored in cargo containers. The suitability of cosmic muons for a number of monitoring and imaging applications has been investigated over the years and includes applications to archaeology, volcano imaging, material identification and medical diagnosis. Novel applications of cosmic ray muons have been proposed for examination of comatose patients towards bone density monitoring and also for determination of the molten nuclear fuel location in nuclear reactors having suffered from the effects of a severe accident similar to the one happened in Fukushima. There is a need to develop a capability to monitor nuclear waste and investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Monitoring nuclear waste and controlling nuclear material at its source is one of the main strategies to minimize the risks of nuclear proliferation and reduce potential homeland threats. Since the early 1950's, when the first nuclear power plant began to produce electricity, approximately 65,000 metric tons of spent nuclear fuel have been generated, 25% of which are under dry storage conditions. After the spent nuclear fuel has been placed inside the dry cask, the cask is welded, not allowing for visual inspection. During the long storage life of the cask, intermediate handling and transportation activities take place which does create concerns regarding fuel integrity. Accidents and natural disasters also contribute to this concern. Previous work on spent nuclear fuel dry cask monitoring using muons has shown that the muon scattering distributions are adequately separated and a and a decision rule derived from Bayes' Theorem rendering the classification of different dry casks feasible. Muons present significant advantages over the existing imaging techniques such as the utilization of the passive nature of muons, the lack of radiological sources and consequently the absence of any artificial radiological dose. In the present work, a methodology is developed to demonstrate the applicability of muons for imaging of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amounts of loading. Monte Carlo simulations of dry casks with different spent nuclear fuel loading are performed to simulate the passage of polyenergetic muons having energies in the range of 1-60 GeV through matter. A 'Muon Generator' is developed to generate muons that follow the actual sea level energy distribution. The 'Muon Generator' is validated against a series of actual spectrum measurements and then coupled with the Monte Carlo particle transport code GEANT4. The accumulated dataset of scattering angles were processed with the Point-of- the-Closest-Approach (PoCA) imaging algorithm and the obtained images demonstrate that an empty dry cask can be clearly distinguished from a half-loaded or a fully loaded one. The apparent separation of the images demonstrates that the muon scattering can be used as a feature for imaging spent nuclear fuel dry casks. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Multiscale Modeling of Accident Tolerant Fuels under the NEAMS ATF Program

The U.S. Department of Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has been developing computational analysis tools for nuclear fuel simulation. These codes include BISON and Marmot for investigation of fuel behavior at the continuum and microstructural scales, respectively. In 2014, NEAMS introduced what it calls High Impact Problems (HIPs) into its development plan. These HIPs are intended to make significant advancements in a particular area of nuclear power research in a short period of time (3 years or less). NEAMS has chosen an accident tolerant fuel (ATF) project, which emphasizes utilizing BISON and Marmot to model proposed materials, as its first HIP. Since the events that occurred at the Fukushima-Daiichi nuclear power plant in 2011, the U.S. Department of Energy's Office of Nuclear Energy accelerated research in nuclear fuel concepts with enhanced accident tolerance as part of its Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC). Accident tolerant fuels are expected to perform at least as well as conventional fuel during normal conditions and provide significantly increased response time in the event of an accident. Given the many unknowns associated with proposed ATF concepts, the research and development being undertaken by AFC in this area is a challenge, particularly considering the desire to have a lead test rod or assembly in a commercial plant by 2022. With the AFC as its primary stakeholder, the NEAMS ATF HIP is developing material models (using a multi-scale approach) for multiple candidate materials. Through a variety of analyses employing these models, the ATF HIP is providing insight and support to the AFC as it strives to understand these materials and prioritize further development. Use of computational tools such as BISON and Marmot are a vital component of the AFC's strategy since experimental data will not be available broadly and quickly enough to understand the proposed materials fully by the time the AFC must make further development decisions. This summary reviews the overall approach of the NEAMS ATF HIP. First, we give an overview of the current materials chosen for investigation. We also review the overall multi-scale approach to model development. We briefly review the work underway at multiple national laboratories to develop mechanistic engineering scale models based on lower length scale insight. Note that coordinated work is also underway at multiple universities, but those efforts are not detailed here. The NEAMS ATF HIP is assisting the AFC in assessing two accident tolerant materials, uranium silicide fuel and FeCrAl cladding. A multi-scale material model development approach is being undertaken to formulate the models necessary to investigate the materials over a range of normal and off-normal scenarios. This approach includes molecular dynamics, density functional theory, mesoscale modeling, and engineering scale modeling. Sensitivity studies are underway to understand the most critical scenarios and material model input parameters. The information generated is being given to the AFC for their use in selecting promising ATF concepts. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Transition to Uranium-Based Fast Reactor Fuel Cycle Options

An Evaluation and Screening (E and S) study of nuclear fuel cycle options was recently conducted for the U.S. Department of Energy to identify promising fuel cycle options with the potential for achieving substantial improvements compared to the current nuclear fuel cycle in the U.S. This study considered the entire fuel cycle, i.e., the complete nuclear energy system from mining to disposal including both once-through and recycle fuel cycles. The E and S study considered performance at steady-state when evaluating the potential benefits and challenges to achieving that eventual end state. Based on the steady-state findings from the E and S, a set of four fuel cycle options (FCOs) were identified as most promising, in terms of offering the most benefits compared to the current cycle. These FCOs have natural uranium (or existing stocks of uranium such as depleted uranium) as the makeup feed sent to the fast reactor (FR) fleet combined with two dimensions. The first dimension is whether the FR is a single-stage (produces just enough fuel resources to sustain operation and growth of the FR fleet) or a two-stage (produces additional fuel resources to also sustain a companion fleet of thermal reactors) fuel cycle. In all four cases, all of the uranium is continuously recycled. The second dimension is whether only the plutonium is co-extracted and recycled with the uranium (with minor actinides disposed as waste) or all transuranic elements are co-extracted and recycled. The four most promising options are: 1) single-stage FR recycling of U/Pu (EG23), 2) recycling of U/TRU in single-stage FR (EG24), 3) recycling of U/Pu in two-stage FR/thermal reactors (EG29), and 4) recycling of U/TRU in two-stage FR/thermal reactors (EG30). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Impact of Isotope Fidelity on Fuel Cycle Calculations

The purpose of this study is to examine the importance of the isotopic composition of plutonium on PWR MOX fuel fabrication as part of a multi-tiered fuel cycle, and the effect of its evolution during a transition. Two main effects can modify the isotopic composition of the plutonium used to build the PWR-MOX fuel: decay and depletion. This analysis is based on a transition of the nuclear fuel cycle to a configuration identified as 'Evaluation Group 29' or 'EG29', in the recently-published Evaluation and Screening (E and S) report. The EG29 fuel cycle corresponds to a double strata fleet. The first stratum is composed of sodium-cooled fast reactors (SFRs) multi-recycling plutonium. Surplus plutonium is sent to the second stratum composed of pressurized water reactors with full mixed oxide (U/Pu)O{sub 2} cores (MOX-PWRs). The PWR stratum is also multi-recycling its plutonium. The target energy generation ratio for the final EG29 system is 70:30 (SFR:PWR), with a 1% annual growth in nuclear energy demand. All the calculation of this study have been performed using the Cyclus fuel cycle simulator framework. The facility models are from the Cycamore package, developed by the Cyclus team, and containing a Pu-equivalent fuel fabrication model as well as a simple fixed mixing ratio fuel fabrication model. A new neural network model has also been contributed to the Cyclus environment using the cyCLASS tool, allowing the use of the fuel fabrication and depletion models from the CLASS tool in Cyclus. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2016

Fundamental Properties for Evaluation of Uranium-free TRU-Zr Metal Fuel Performance

FR (Fast Reactor) is widely recognized to effectively burn plutonium and MA (Minor Actinides) (i.e., TRU: Transuranic) due to its higher fission-to-neutron-capture ratio compared to LWR. When TRU in LWR (Light Water Reactor) spent fuel is recycled into FR cycle, the radiotoxicity of the residual nuclear wastes from LWR spent fuel is reduced to natural uranium level after several hundred years instead of more than hundreds of thousand years for no recycling case. It could significantly improve public perception to nuclear waste disposal and might reduce the burden of the disposal. However, U-TRU-fueled fast reactors also produce TRU while burning TRU because they use uranium together with TRU. For example, FR, even with low conversion ratio of 0.75, burns only 0.18 t TRU/GWe/y, while LWRs produces about 0.22 t TRU/GWe/y. It is not a realistic scenario which needs almost the same capacity of fast reactors as that of LWRs to continuously burn TRU at LWR-dominant era in the near term at least. Therefore, a FR system to burn TRU from LWR with the minimum reactor units is desirable. The most effective way is to use uranium-free TRU fuel since it does not produce any new TRU. It can reduce the capacity of the TRU burner units and the associated fuel cycle facilities to about 1/5 and 1/8 respectively. Metal fuel FR cycle is considered to be one of the most promising ways because uranium-free TRU metal fuel could be recycled by pyro-process for U-Pu-Zr metal fuel without large modifications. This paper shows the recent research progress on fundamental properties of uranium-free TRU metal fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Material Balance Area Design for the Transuranic Fuel Cycle Employing High Temperature Gas Cooled Reactors

A major impetus has been to ensure safety in nuclear power plant systems from the beginning of its development to the current period of nuclear renaissance. As one of the next generation reactors with a high level of nuclear safety, high temperature gas cooled reactor (HTGR) has attracted large attention because of the possibility to shut down the reactor safely and remove decay heat without the use of active cooling. A deep-burn concept of fuel utilization is proposed and studied for HTGRs to alleviate plutonium proliferation concern. The idea is to utilize the plutonium extracted by reprocessing the light water reactor (LWR) spent fuel and discharge it only at higher fuel burn-ups of up to 90 atom %. In addition HTGRs have been studied to utilize transuranic (TRU) fuel. A large amount of special nuclear material (SNM) will be used in the HTGR advanced fuel cycle. The inherent proliferation resistance and material attractiveness of the fresh fuel and discharged spent fuel have been discussed against the diversion from peaceful to military uses. On the other hand safeguards of SNM is also important to prevent nuclear proliferation under the Non-Proliferation Treaty between the state and International Atomic Energy Agency (IAEA). Facility operators are responsible to design material balance area (MBA) and provide enough information on material balance of SNMs for the safeguards conclusion. The material unaccounted for (MUF) and measurement uncertainty in MUF (σ{sub MUF}) represent accuracy of material balance. The MBAs have to be designed and established to achieve enough confidence level on MUF and σ{sub MUF}. The MBA design is challenging for nuclear facilities managing large amount of bulk (instead of items) SNM because of potentially large MUF and σ{sub MUF}. A concept and guidance of MBA design for uranium fueled HTGRs has been discussed. In this study, a MBA design is presented for the transuranic fuel cycle employing HTGRs. The MUF and the σ{sub MUF} is calculated and evaluated using international target values (ITVs) provided by the IAEA. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Status Report on Fast Flux Test Facility Mechanistic Fuel Failure Experiment Analysis with BISON for Post Irradiation Examination Support

The renewed interest in metallic U-Zr nuclear fuel alloy has led to a drive for deeper understanding of the mechanisms driving the phenomena observed under irradiation conditions. The Department of Energy Advanced Fuel Campaign has developed infrastructure to support metallic fuel development, including Post Irradiation Examination (PIE) of legacy Fast Flux Test Facility (FFTF) Mechanistic Fuel Failure (MFF) experiments. The PIE performed on legacy FFTF MFF experiments gives insight on metallic fuel performance and can address the lack of knowledge and scarcity of reliable data identified in several studies over recent years. Unfortunately, PIE efforts can cost significant time and resources which can impede the progress of metallic U-Zr fuel development. Metallic U-Zr fuel performance modeling can be used to inform PIE efforts on regions of interest for relevant investigations and can help understand phenomena observed in PIE. This report demonstrates the current progress of FFTF MFF fuel performance simulations using the BISON fuel performance code and discusses the support provided by simulation to PIE efforts. Progress in temperature, profilometry, fission gas release, plenum pressure, and zirconium redistribution simulation results have been demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Fuel Performance Modeling Internship Final Presentation

This study examines the performance of U-Zr and U-Pd-Zr annular metallic fuel rodlets and details the current status of modeling efforts regarding U-Pu-Zr solid metallic fuel rodlets which were experimentally evaluated as part of the Advanced Fuels Campaign (AFC). The AFC mission is to develop novel fuel technologies and facilitate the implementation of those technologies by industry partners. A key objective is to improve steady-state and transient performance over current fuel types. The experiments of interest in this study included metallic fuel rodlets within HT-9 cladding which were placed in SS-316 capsules and inserted in the Advanced Test Reactor (ATR). Certain mechanical and thermal conditions cannot be directly evaluated through experiments and fuel performance modeling is used to shed light on this evolution over time. In this study, BISON Multiphysics simulations are leveraged to investigate the state of the fuel system throughout and after the experimental conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Decay heat analysis for advanced reactor spent fuel transportation and storage applications

Accurate characterization of nuclide inventories and decay heat in spent nuclear fuel is critical for ensuring its safe handling, storage, transportation, and disposal. Although extensive research has been conducted on light-water reactor fuel, advanced reactors present unique challenges due to their diverse core configurations, fuel characteristics, neutron energy spectra, and burnup levels. Building upon previous efforts that developed representative reactor core models for various advanced reactor types and fuels, this study evaluates reactor-specific decay heat characteristics. The results highlight significant variations across advanced reactor types as well as across reactor designs within the same reactor type, and they provide comparison to typical commercial light-water reactor fuel. For example, thermal-spectrum reactor fuels were observed to have an approximately 100-fold decrease in decay heat over the first decade of cooling, whereas the reduction was 10-fold for fast-spectrum reactor fuels. Mass-specific decay heat at discharge can differ by three orders of magnitude among the fast and thermal reactor systems considered. Overall, for the analyzed advanced reactor fuel, fewer than 17 nuclides account for over 99% of total decay heat at 0.5 years, and that number drops to fewer than 7 nuclides at 100 years of cooling. By quantifying reactor-specific decay heat trends and nuclide contributions, this work provides a technical basis to support the development of spent fuel management strategies for advanced reactor fuels as well as safety evaluations for storage, transportation, and long-term waste disposal.

Advanced reactors↗

DOE OSTI2024

Numerical investigation of equivalence ratio effects on Flex-Fuel mixing controlled combustion enabled by prechamber ignition

There is interest in adapting renewable and low-carbon intensity fuels to heavy-duty engines to help displace criteria and greenhouse gas emissions associated with diesel combustion. Low-carbon fuels have inherently low cetane numbers and thus pose a significant challenge when considering direct substitution in diesel combustion systems. Use of an actively fueled prechamber as an ignition source to initiate mixing-controlled combustion (PC-MCC) of low-cetane fuels is becoming an attractive combustion mode to alleviate the identified reactivity deficit. This work focuses on the utilization of an active prechamber to facilitate diesel-like combustion of bioethanol-gasoline fuel blends in a heavy-duty engine. Recent results in this emerging technology suggest that ignition quality is uniquely coupled to prechamber equivalence ratio. In this numerical study, the fundamental implications of prechamber equivalence ratio on the ignition performance of direct injected fuels ranging from E10 to E100 are investigated using CONVERGE. Parametric studies of the prechamber operating strategy were assessed at diesel-like conditions to characterize the performance trends relative to a diesel baseline at the same boundary conditions. Simulation results indicate that PC-MCC is flex-fuel capable and achieves diesel-like ignition qualities and combustion processes for all fuels considered under stoichiometric and rich prechamber conditions. To characterize the equivalence ratio trends observed, a novel inflow boundary modeling technique was utilized to prescribe turbulent jets in place of the prechamber where selective speciation of the jet composition was conducted to isolate the ignition contributions of radical and reacting species in lean, stoichiometric, and rich jets. The inflow boundary modeling suggests that excess fuel and combustion intermediates present in jets produced from stratified and fuel-rich prechamber operation promotes higher jet temperatures and as such are superior ignition sources. Relative to a lean prechamber jet, the peak temperature of a reacting jet from a rich prechamber was up to 600 K hotter at fixed distances from the orifice exit. Furthermore, radicals also demonstrated an influence on the ignition process, but the combustion mode was identified to be thermally dominant.

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2016

Comparison of the safety-related physical and combustion properties of liquid hydrogen and liquid natural gas in the context of the SF-BREEZE high-speed fuel-cell ferry

Here, we review liquid hydrogen (LH 2 ) as a maritime vessel fuel, from descriptions of its fundamental properties to its practical application and safety aspects, in the context of the San Francisco Bay Renewable Energy Electric Vessel with Zero Emissions (SF-BREEZE) high-speed fuel-cell ferry. Since marine regulations have been formulated to cover liquid natural gas (LNG) as a primary propulsion fuel, we frame our examination of LH 2 as a comparison to LNG, for both maritime use in general, and the SF-BREEZE in particular. Due to weaker attractions between molecules, LH 2 is colder than LNG, and evaporates more easily. We describe the consequences of these physical differences for the size and duration of spills of the two cryogenic fuels. The classical flammability ranges are reviewed, with a focus on how fuel buoyancy modifies these combustion limits. We examine the conditions for direct fuel explosion (detonation) and contrast them with initiation of normal (laminar) combustion. Direct fuel detonation is not a credible accident scenario for the SF-BREEZE. For both fuels, we review experiments and theory elucidating the deflagration to detonation transition (DDT). LH 2 fires have a shorter duration than energy-equivalent LNG fires, and produce significantly less thermal radiation. The thermal (infrared) radiation from hydrogen fires is also strongly absorbed by humidity in the air. Hydrogen permeability is not a leak issue for practical hydrogen plumbing. We describe the chemistry of hydrogen and methane at iron surfaces, clarifying their impact on steel-based hydrogen storage and transport materials. These physical, chemical and combustion properties are pulled together in a comparison of how a LH 2 or LNG pool fire on the Top Deck of the SF-BREEZE might influence the structural integrity of the aluminum deck. Neither pool fire scenario leads to net heating of the aluminum decking. Overall, LH 2 and LNG are very similar in their physical and combustion properties, thereby posing similar safety risks. For ships utilizing LH 2 or LNG, precautions are needed to avoid fuel leaks, minimize ignition sources, minimize confined spaces, provide ample ventilation for required confined spaces, and to monitor the enclosed spaces to ensure any fuel accumulation is detected far below the fuel/air mix threshold for any type of combustion.

08 HYDROGEN↗

DOE OSTI2021

Performance and cost analysis of liquid fuel production from H 2 and CO 2 based on the Fischer-Tropsch process

Electro-fuels can be produced from concentrated sources of carbon dioxide and hydrogen using electricity generated from renewable sources; this process enables energy storage at high volumetric energy density. Among the electro-fuels options, FT (Fischer-Tropsch) fuel is attractive for heavy-duty trucks and non-road transportation applications. This study conducts a techno-economic analysis of FT liquid fuel production from H 2 and CO 2 using a detailed performance analysis. Minimum fuel selling price is estimated for a broad range of H 2 and CO 2 prices and for a range of potential CO 2 credits. The analysis indicates that H 2 price has the largest impact on the minimum selling price of FT fuel. FT fuel production with a CO 2 price of $17.3/metric ton requires an H 2 price of $0.8/kg to be cost-competitive with the pre-tax petroleum diesel price of $3.1/gal in 2050 (before the application of any CO 2 credits). When the H 2 price is $2.0/kg from central water electrolysis (2020 target), the minimum selling price of the FT fuel is $5.4–5.9/gal. A sensitivity analysis shows that future system optimization of FT fuel production could focus on improving the H 2 and CO 2 recycle contributions and FT fuel conversion ratio. The analysis results can be combined with various upstream systems for H 2 and CO 2 production.

08 HYDROGEN↗

DOE OSTI2022

Transmission Electron Microscopy based Characterization of a U-20Pu-10Zr Fuel Irradiated in Experimental Breeder Reactor-II

U-Pu-Zr metallic fuels are important fuel candidate for future advanced and/or test reactors. To better understand fuel performance and guide future fuel development, a U-20Pu-10Zr (in weight) metallic fuel irradiated in the Experimental Breeder Reactor was revisited using advanced electron microscopies. This fuel was irradiated to a burnup of 7.6 % at a peak cladding temperature of 521 °C. In this research, several transmission electron microscopy samples were extracted from different but representative radial locations of the fuel cross section using focused ion beam technique. Phase identification and chemical analysis with sub-micron spatial resolution was carried out using scanning transmission electron microscopy. Multiple phenomena that are critical to fuel performance, such as Zr phase and its distribution in a different matrix, were revealed in an unpreceded and highly detailed manner. An improved understanding of fuel reconstruction in U-Pu-Zr metallic fuel under reactor irradiation is provided.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Comparative post-irradiation examination of high burnup U-19Pu-10Zr: Assessing steady-state irradiation behavior against historical and modeled fuel performance

Here, the development of next-generation sodium-cooled fast reactors necessitates comprehensive research on metallic fuels to maximize economic performance while ensuring safe operation. In this study, we investigated the steady-state irradiation behavior of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, in preparation for planned safety testing. Post-irradiation examination (PIE) was performed to quantify fuel column elongation, regions of low-density at the top of the fuel column, pin deformation, fission product distribution, fractional fission gas release, microstructural evolution, and fuel constituent redistribution. Benchmarking against existing PIE data from U-19Pu-10Zr fuel pins irradiated in EBR-II revealed consistent patterns in fuel column elongation and cladding diametral strain. However, both pins exhibited longer low-density structures, and destructive examination of DP-36 revealed more complex constituent redistribution patterns compared to previously reported data for ternary fuel pins. The steady-state irradiation of both pins was also modeled using BISON. Comparisons of PIE results with modeled predictions showed overall agreement in fractional fission gas release but consistent overestimation of axial and radial swelling due to gaseous and solid swelling models. These findings underscore the critical importance of pre-test characterization on test and sibling pins to accurately capture steady-state fuel behavior ahead of transient testing, thus establishing a baseline for post-test comparison. Additionally, these analyses identified key data gaps that warrant further investigation to improve the understanding and prediction of fuel swelling, thereby enhancing the synergy between modeling and experimental efforts in supporting accident testing.

BISON↗