Engineering Papers⌕ Search

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.

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

At least 1,729 records · Page 96
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Auto Stop-Start Fuel Consumption Benefits

With increasingly stringent regulations mandating the improvement of vehicle fuel economy, automotive manufacturers face growing pressure to develop and implement technologies that improve overall system efficiency. One such technology is an automatic (auto) stop-start feature. Auto stop-start reduces idle time and reduces fuel use by temporarily shutting the engine off when the vehicle comes to a stop and automatically re-starting it when the brake is released, or the accelerator is pressed. As mandated by the U.S. Congress, the U.S. Environmental Protection Agency (EPA) is required to keep the public informed about fuel saving practices. This is done, in partnership with the U.S. Department of Energy (DOE), through the fueleconomy.gov website. The “Fuel-Saving Technologies” and “Gas Mileage Tips” sections of the website are focused on helping the public make informed purchasing decisions and encouraging fuel-saving driving habits. Here, in order to provide users with accurate information about the auto stop-start feature, experiments were conducted to determine its fuel economy effect. Four vehicles were tested both with and without the feature enabled under three test cycles: the Federal Test Procedure (FTP) city fuel economy test, the US06 high acceleration aggressive driving schedule that is often identified as the “Supplemental FTP” driving schedule, and the EPA New York City Cycle (NYCC). The results were compared to measure the fuel economy and consumption effects of using the auto stop-start feature. It was found that the fuel economy improvement varied significantly between drive cycles depending on the amount and percentage of idle time during the test. The largest fuel economy improvements were 7.27% and 26.4% for the FTP and NYCC, respectively.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2014

Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Separation of the rare-earth fission product poisons from spent nuclear fuel

A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Bison Fuel Performance Modeling on RIA in a PWR

This report is prepared to fulfill a Level 2 milestone for CASL on “deliver and document RIA fuel performance modeling using Bison fuel performance code.” The work performed is focused on modeling the full length commercial fuel rod response under a design basis Reactivity Initiated Accident (RIA), in particular, a hypothetical Control Rod Ejection Accident (CREA) in a Pressurized Water Reactor (PWR). Fuel rods with different burnups, 0 (fresh fuel), 20, 40, and 60 GWd/tU have been modeled with a reactivity worth insertion of $1.24 at Hot Zero Power (HZP) condition. Bison failure modeling method, and a brief summary of validations on selected RIA cases from CABRI and NSRR testing facilities are also described. The work performed has demonstrated the successful application of Bison code in the modeling of thermal and mechanical responses in RIAs for PWRs. In particular, Bison code has a more mature capability in the modeling of the frictional contact between pellet and cladding of the high burnup fuels, which is critical for modeling the Pellet Clad Mechanical Interaction (PCMI). The modeling of RIA using Bison code shows fuels in PWR environment could have different responses than in the simulated RIA tests, and results indicate that a transition from PCMI to Departure from Nucleate Boiling (DNB) could happen at relatively lower enthalpies: a better heat transfer from the fuel to coolant could increase cladding temperature and ductility, but at the same time also reduces the margin to DNB, which could cause high temperature oxidation and ballooning type of failures. The precise determination of the transition into DNB may rely on a more accurate input of the power histories in the simulated RIAs from VERA, the core simulator of CASL. The limiting conditions in PWRs could be quite different than those tests with non-typical testing conditions, and the improvement of the modeling of thermal responses, such as, clad coolant heat transfer, gap heat transfer, clad thermal inertia, and radial power profiles could be important considerations of the RIAs. In modeling the fuel responses in a fast transient, a number of variables could affect the initial burnup states, and some material models, e.g., fuel mechanical models, fission gas release, and corrosion models, and parameters can affect modeling results as well. Those uncertainties and advanced modeling capabilities need to be investigated in the future study of RIAs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Neutronic Consideration of TREAT Facility Fuel SiC Recladding

The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. Benchmark efforts have demonstrated sensitivity of TREAT neutronics models to graphite cross section data and key neutron poisons found in the fuel assemblies [2]. During fabrication of TREAT fuel assemblies, a small concentration of boron was inadvertently introduced with variable concentration into the aggregate of fuel blocks prepared for TREAT use. The concentration of boron, and its respective uncertainty, represent the most sensitive component in a neutronics study; unfortunately, nothing can be done to reduce or otherwise mitigate its impact. Hydrogen in the fuel blocks reduces the mean free path of the core and its true quantity and uncertainty is approximated. The hydrogen remains from pitch binder used in the fuel fabrication process and was not completely released due to incomplete fuel graphitization. Destructive fuel testing is necessary to quantify the actual hydrogen content and to improve benchmark model quality. A third neutron poison of some significance is the hafnium impurity in the Zircaloy-3 (Zry-3) assembly cladding. The quantity of hafnium is expected to be within nominal manufacturing limits for nuclear grade zircaloy. Recladding the fuel assemblies with material that is less neutron absorbing is expected to benefit reactor neutron economy. Moreover, TREAT’s total transient energy is restricted based on the rate of Zry-3 oxidation in TREAT’s air coolant environment. Hence, cladding with more benign air reaction kinetics could dramatically uprate the facility’s experimental capability. Silicon carbide (SiC) has been proposed as a candidate material for recladding TREAT assemblies, as SiC/SiC demonstrate favorable properties within nuclear environments [3]. The current operational limits of TREAT are restricted by the oxidation of Zry-3. Change of the cladding to SiC could feasibly double the current temperature limitation. Recladding opportunities allow for previously damaged fuel assemblies to be returned to service, as Zry-3 is not commercially available. The initial effort required to consider proceeding with any recladding effort is to neutronically characterize key reactor parameters while replacing Zry-3 with SiC. A characterization summary, with respective results, is provided herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Researching the Effects of Long-Term Dry Storage on Commercial Spent Nuclear Fuel

When the fuel in a nuclear reactor no longer contributes efficiently to power production, it is discharged and stored onsite in a water-filled spent fuel pool. When nuclear power plants in the United States were originally designed, it was expected that onsite storage would be temporary. The used nuclear fuel, also known as spent nuclear fuel (SNF), was to be removed from the storage pools after only a few years of cooling and sent for reprocessing. However, commercial reprocessing was not implemented in the United States, and utility spent fuel pools began running out of room. In order to continue power production, nuclear utilities began expanding onsite fuel storage capacities, first by increasing the density of fuel in the fuel pool racks, and then by developing and implementing dry storage casks. The dry storage casks are stored onsite by each utility at a licensed and secured independent spent fuel storage installation (ISFSI). Dry storage casks are robust systems typically consisting of a sealed metal cylinder surrounded by a combination of steel and concrete to provide radiation shielding. The first dry storage casks were loaded with SNF in 1986. Currently there are over 3,000 loaded dry storage casks, and approximately 200 new casks are being placed into service annually at operating nuclear power plants nationwide. Multiple nuclear power plant sites in the United States have shut down and completed decommissioning activities except for the removal of fuel in dry storage. Eventually, the 75,000+ tons of discharged SNF at more than 70 secured locations will need to be safely and securely transported to a new location(s) for eventual disposition. The continued safety and security of this material must be assured.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Simulation and Optimization of Volatile Fatty Acid Upgrading Strategies for Sustainable Transportation Fuel Production

Underutilized wet waste is a promising feedstock for production of carbon-neutral or carbon-negative liquid transportation fuels. Arrested methanogenesis of wet wastes by microbes, which produces mixtures of C2-C8 carboxylic acids (volatile fatty acids, VFAs), is practiced industrially at a pilot scale and the VFA products can be catalytically upgraded to molecules suitable for use as transportation fuels (alcohols and alkanes) via sequential ketonization and hydrogenation, both steps of which have been demonstrated with high (>90%) yield at the lab scale. We present in this work a simulation-based decision-making framework which evaluates upgrading strategies of VFAs to alcohol- and hydrocarbon-based neat and blended diesel, jet, and gasoline fuels. The processes utilize sequential upgrading steps (ketonization and hydrogenation) accompanied by one distillation step (either separating VFAs before ketonization or alcohols after ketone hydrogenation) to form two liquid streams: a light alcohol stream best suited as a light-duty fuel and a heavy alcohol or hydrocarbon stream best suited as a diesel or jet fuel. Suitability of the liquid products as transportation fuels and maximum blending levels with petrofuels comes from simulation of critical fuel properties (boiling point, flash point, lower heating value, viscosity, melting point, water solubility, and cetane/octane number). Catalytic upgrading steps and fuel property mixing models used by the simulation are experimentally validated. We demonstrate the flexibility of the decision-making algorithm by evaluating processing scenarios for experimentally-observed VFA profiles with varied carbon chain length distributions. Processing scenarios are optimized for either maximum total renewable carbon utilization or production of heavy-duty (jet or diesel) fuels. We evaluate the economic viability and CO2 emissions of each proposed upgrading scenario using techno-economic and life cycle analyses. The decision-making framework developed in this work can also be used to down-select promising strategies for upgrading of other bio-based feedstocks to sustainable fuels and chemicals.

aviation fuel↗
DOE OSTI2023

Marine Alternative Fuel Pricing, Supply, and Demand

The International Maritime Organization low sulfur fuel rules and greenhouse gas-reduction strategy may create a significant perturbation to global refinery operations, impacting volumes and prices of marine fuel globally and within the United States and incentivizing the transition to alternative fuel use. The goals of this project are to (1) enhance understanding of how very low-sulfur fuel-primarily fuel oil and diesel-and low carbon fuel requirements and promising biofuel processes will affect the marine fuel supply chain; and (2) explore how these perturbations interact with and impact indicators such as pricing, number of trips, and demand behavior, along with the potential to meet low sulfur and carbon fuel demand with biofuel supply chains, (3) merge innovative thinking in the area of marine fuels within research centers of DOE (NREL) and the U.S. Department of Transportation (Volpe). These goals will be accomplished through the combination of detailed refinery, marine fuel burn, system dynamics, and geospatially explicit linear programming models.

biofuel↗
DOE OSTI2023

Effective Lanthanide Diffusivity through U-Zr Metallic Fuel

Fuel-cladding chemical interaction (FCCI) is one of the main concerns for the performance of U-Zr metallic fuels. Although FCCI results in complex phase transformations and morphological changes in both the fuel and cladding sides of the interface, the formation of a brittle layer in the cladding side of the interface (sometimes referred to as wastage) has one of the most significant impacts on the cladding mechanical integrity. This brittle layer is associated with the formation of intermetallic compounds between the cladding constituents Fe, Cr and the lanthanide fission product species. The formation of this cladding wastage layer has been the focus of recent efforts to develop a mechanistic modeling framework, including production of lanthanides in the fuel, transport through the fuel to the fuel-cladding interface, and phase transformation to form intermetallic phases. To calculate the rate of transport of lanthanides through the fuel, a multi-scale computational approach has been used. Atomistic calculations have been performed to determine the diffusivities of Nd, the most prevalent lanthanide species, through the solid fuel matrix and along pore surfaces. These diffusivities have been used as input parameters for a mesoscale model to determine an effective diffusion coefficient that accounts for the formation of porosity, its interconnection, and infiltration with bond sodium from the fuel-cladding gap. The effective diffusivity has been used in engineering-scale simulations using the BISON fuel performance code, and BISON simulations have been compared to representative validation cases from the EBR-II reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Experimental test plan for PWR sister rods in the high burnup spent fuel data project

The High Burnup Spent Fuel Data Project (herein referred to as the 'project'), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally defining the effects of long-term storage and transportation on high burnup (HBU) pressurized water reactor (PWR) fuel (>45 gigawatt days per metric ton uranium). The goals of the project are to provide confirmatory data for model validation and potential improvement; to provide input to future dry storage cask design; to support license renewals and new licenses for Independent Spent Fuel Storage Installation facilities; and to support transportation licensing for high burnup used fuel. In support of these project goals, an instrumented TN-32B bolted-lid cask at the North Anna reactor site will be loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO, and M5. In approximately 10 years, the instrumented cask will be opened for inspection and several fuel rods from the project assemblies will be characterized. To support the project goals, the project pulled 25 'sister rods', 9 from the project assemblies and 16 from similar HBU assemblies, for characterization. The 25 sister rods are all high burnup and cover the range of cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed nondestructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of the storage period (the 'cask rods') to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup used fuel. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Neutronics Analysis Supporting Baseline Fuel-Rodlet Irradiation Design in TREAT Multi-SERTTA Vehicle

The Transient Reactor Test (TREAT) Facility is a thermal-spectrum test reactor designed to evaluate reactor fuels and structural material under simulated nuclear excursion and transient power/cooling mismatch situations in nuclear reactors. TREAT was historically operated from 1959 through 1994. There are ongoing activities to resume its transient testing capabilities to support advanced reactor designs and validate computational predictions of fuel and core behavior. Such validation will be an important factor for review and approval of next generation power reactor design and safety evaluations. Historic methods used in TREAT were very approximate and served to estimate a predicted transient power shape. A number of calibration test experiments were then required prior to the actual transient test to better estimate the power that would have been deposited in the experiment during a transient pulse. Much of the experience in the use of these legacy methods has also been lost over the past two decades. Monte Carlo criticality codes are being utilized for calculation of the initial critical state, control rod reactivity worth, and power coupling between core and experiments. Current efforts primarily support neutronics characterization with safety and design analyses for the Multi-SERTTA (Static Environment Rodlet Transient Test Apparatuses) experiment test vehicle, which can accommodate up to four concurrent rodlets under separate environmental conditions. Ongoing efforts are necessary in preparation of transient testing support for Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. The power coupling factor (PCF) is one of two key parameters utilized in designing and characterizing the performance of a TREAT fuel irradiation experiment. The second key parameter is the Transient Correction Factor (TCF) which accounts for time-dependent effects. The PCF is the effective quantity of fission-generated energy per mass of fuel specimen per total core energy. The quantity of energy deposited into a fuel rodlet under various test environments can be used to evaluate overall fuel performance, including failure mechanisms and thresholds. Baseline fuel rodlet calculations have been performed to estimate the effectiveness of the Multi-SERTTA test vehicle and optimization of PCF for various test conditions. The baseline serves as reference fuel pin conditions against which calibration and baseline experiments can be performed prior to actual ATF transient testing. Baseline experiments further serve to validate computational analyses and experimental test methods prior to design planning, experimentation, and evaluation of the ATF candidates. This paper serves to document activities performed to evaluate a reference Pressurized Water Reactor (PWR) rodlet design in the TREAT Multi-SERTTA test vehicle. (authors)

42 ENGINEERING↗
DOE OSTI2019

Simulation of Spent Nuclear Fuel loading into a Final Waste Repository

Previous work towards studying repository loading have used: - spent nuclear fuel (SNF) with an average burnup composition; - a lumped capacitance thermal model for calculating temperature in a Cyclus repository model. The goal of this work is to improve on the repository models and use U.S. historical SNF inventory data in simulations to more accurately study the loading of a waste repository. These goals will be achieved by: - using UNF-ST-and-DARDS Unified Database (UDB) that has historic assembly-specific data (e.g, isotopic composition, heat) in Cyclus simulations; - implementing a more accurate thermal model within a Cyclus repository model. Objectives: - Create a Cyclus spent fuel conditioning model that packages spent fuel bundles into packages which have user-defined properties. - Create a Cyclus medium-fidelity repository model that gives accurate time and spatial dependent temperature values and loads the repository based on a user-selected loading strategy. Cyclus is an agent-based extensible framework for modeling flow of material through user-defined nuclear fuel cycles. In Cyclus, each facility in the fuel cycle is modeled individually and the facilities interact with one another as independent agents. The spent fuel conditioning model accepts spent fuel bundles and puts them into a cylindrical waste package. In the spent fuel conditioning model, the user can define variables: For each layer, radius, thermal conductivity, thermal diffusivity; For each package, Number of spent fuel bundles, Radius and height. The waste repository model accepts waste packages and emplaces them into available positions within the waste repository based on a thermal criteria. The thermal criteria is a temperature limit at the interface between the waste package surface and the host geology, that is set based on the repository's host geology. In the waste repository model, the user can define the variables: Capacity, Distance between waste packages, Distance between drifts, Repository host geology, Loading Strategy. After the addition of new waste packages at each time step, the waste repository model recalculates the temperature at each location in the repository. If the addition of this new package causes its temperature to exceed the thermal limit, it will be placed back into the buffer. A thermal model that relies on a transient 'outside' model and quasi-steady-state 'inside' model is used to accurately determine the temperature in the repository. The 'outside' model assumes a homogenous medium with the Engineered Barrier System (EBS) replaced by the geologic medium. Temperature solutions for the central waste package, adjacent point and line sources are superimposed to calculate the temperature at specific points in the repository. The central drift consists of one finite line source which represents the central waste package. The central drift also consists of point sources that represent neighboring waste packages in the central drift. The neighboring drifts are represented by infinite line sources. The 'inside' model is considered to be at a quasi-steady-state condition because EBS has a relatively low thermal mass compared to the infinite geologic medium. The steady state calculation is performed at each time step with the heat source and interface temperature as boundary conditions. Future Work: Run Cyclus simulations with U.S. historical SNF inventory data, the spent fuel conditioning and repository models to study how waste package acceptance strategies impact repository loading.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel

This research demonstrated the first time that synchrotron X-ray micro-computed tomography (μ-CT) was utilized to analyze the three-dimensional (3-D) morphology of the microstructure of irradiated nuclear fuel. Both absorption contrast and propagation-based phase contrast enhanced μ-CT techniques were evaluated. Porosity and phase region volumes were determined in neutron irradiated uranium 10 wt.% zirconium (U-10Zr) using propagation-based phase contrast enhanced μ-CT. Volumes were determined by image processing software and algorithms, with the porosity volume subsequently compared to the ASTM E562 method. The 3-D porosity of a small region (~ 8 x 10 5 µm 3 ) located in the intermediate fuel redistribution zone that was removed via a focused ion beam was determined to be 7.2% via image processing. A detailed 3-D analysis of the porosity with respect to volume and morphology was provided. Five growth stages of pore evolution were observed including: nucleation, growth, coalescence, interconnected porosity, and extended/interconnected porosity. Four distinct phase regions were identified, including the pores, via both absorption contrast and propagation-based phase contrast enhanced μ-CT. A quantitative comparison of fuel swelling between this localized microscopic U-10Zr fuel region, previously irradiated in the Fast Flux Test Facility, and historic U-10Zr fuels irradiated in the Experimental Breeder Reactor-II (EBR-II) was conducted. It was determined that for similar burnups, the localized swelling in this microscopic fuel region was 7.7%, which was approximately one-half of the value of the macroscopic swelling reported in EBR-II irradiated fuel. With the advent of new specimen preparation tools for irradiated materials, as well as advances in characterization techniques, researchers can now obtain previously inaccessible microstructural insight into irradiated fuels. This possibility naturally lends itself to improved fuel performance models for the nuclear community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Characterization of a minor actinides bearing metallic fuel pin irradiated in EBR-II

The X501 transmutation experiment irradiated in Experimental Breeder Reactor II (EBR-II) was intended to evaluate the safety and performance of the addition of minor actinides (Am, Np) into the metallic nuclear fuel system. The irradiation of this experiment is also of paramount importance for comparisons between true fast spectrum irradiations (e.g. EBR-II) and shrouded irradiation capsules performed at the thermal neutron spectrum Idaho National Laboratory Advanced Test Reactor. The X501 experimental fuel pins were irradiated to a peak measured burnup of 6.2% fission per initial heavy metal atom. This paper reports and discusses Postirradiation Examination results for one of two minor actinide bearing pins from the EBR-II X-501 experiment (X501-G591) with a focus on fuel microstructure evolution. Additionally, fuel swelling, fission product distribution, cladding strain, fission gas release, fuel microstructure and fuel cladding chemical interaction were all evaluated. The fission gas release was 79.9% and the microstructure along the fuel pin presented several characteristics very similar to other EBR-II U-Pu-Zr ternary metallic fuel experience. Minor actinides seem to not dramatically affect the overall performance of this candidate transmutation fuel. Scanning Electron Microscopy was also performed and an in-depth characterization of the different regions and phases is presented. Performance data from these irradiations can be used to inform the feasibility of minor actinide transmutation in future reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Development and formulation of physics based metallic fuel models and comparison to integral irradiation data

Metallic fuel has an important historical significance in the development of nuclear reactors and continues to be relevant to the progression of advanced test and power reactors. A number of models, ranging from empirical to mechanistic, have been developed and implemented in various fuel performance codes to describe U-Zr and U-Pu-Zr fuel and typical fast reactor cladding materials. One challenge of using these models to simulate fuel performance is the inevitable tangling of coupled phenomena that can cloud proper implementation, calibration, and eventual utilization of new models. Here in an effort to provide a baseline capability that will facilitate the use of advanced models, new capabilities have been implemented into the fuel performance code BISON specific to metallic fuel simulations, ranging from materials properties, fission gas release and swelling calculations, coolant channel models, and cladding correlations. These models have been applied to the X441/X441A EBR-II experimental assembly data, a set of irradiated metallic UPuZr fuel rods of varying pin designs. The models implemented in BISON are able to capture the general trend of the expected response of the fuel and cladding to irradiation in EBR-II, especially when considering the spread in experimental measurements and the uncertainties inherited from the historical material models. Ultimately, the models outlined here provide the baseline capabilities on which new models can build upon in order to improve the prediction of metallic fuel performance simulations in off-normal designs or operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

An Alternative to the Continued Accumulation of Separated Plutonium in Japan: Dry Cask Storage of Spent Fuel

As of the end of 2017, Japan had a stockpile of 48 tons of separated plutonium. It will take more than a decade for Japan to convert most of that plutonium into “mixed-oxide” (MOX) fuel and load it into power reactors licensed to use such fuel. This is a very costly program. Including the cost of reprocessing, MOX fuel costs about ten times more than the low-enriched-uranium fuel that otherwise would be used by these reactors. Yet Japan has a policy to start separating more plutonium from spent reactor fuel as soon as the Rokkasho Reprocessing Plant can be put into operation – currently projected for 2021 after 24 years of delay due to a variety of technical problems and upgrades in safety requirements. Japan’s government remains committed to this costly and troubled program primarily because, pending the availability of additional spent-fuel storage space on or off site, reprocessing is seen as the only way to get spent fuel out of reactor pools that are getting full. Because of concerns about the safety of dense-packed spent-fuel pools, however, the prefectures and towns hosting Japan’s power reactors are beginning to accept on-site storage of spent fuel in dry casks. This is the option the United States and many other countries chose after they cancelled their reprocessing programs and their powerplant pools filled up. Its increasing availability in Japan, if backed by a determined policy to accelerate its introduction, should make it possible to cancel reprocessing in Japan as well.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Developing the User Experience for a Next Generation Nuclear Fuel Cycle Simulator (NGFCS)

This project made substantial progress on its original aim for providing a modern user experience for nuclear fuel cycle analysis while also creating a robust and functional next- generation fuel cycle simulator. The Cyclus kernel experienced a dramatic clari cation of its interfaces and data model, becoming a full- edged agent-based framework, with strong support for third party developers of novel archetypes. The most important contribution of this project to the the development of Cyclus was the introduction of tools to facilitate archetype development. These include automated code generation of routine archetype components, metadata annotations to provide re ection and rich description of each data member's purpose, and mechanisms for input validation and output of complex data. A comprehensive social science investigation of decision makers' interests in nuclear fuel cycles, and speci cally their interests in nuclear fuel cycle simulators (NFCSs) as tools for understanding nuclear fuel cycle options, was conducted. This included document review and analysis, stakeholder interviews, and a survey of decision makers. This information was used to study the role of visualization formats and features in communicating information about nuclear fuel cycles. A exible and user-friendly tool was developed for building Cyclus analysis models, featuring a drag-and-drop interface and automatic input form generation for novel archetypes. Cycic allows users to design fuel cycles from arbitrary collections of facilities for the rst time, with mechanisms that contribute to consistency within that fuel cycle. Interacting with some of the metadata capabilities introduced in the above-mentioned tools to support archetype development, Cycic also automates the generation of user input forms for novel archetypes with little to no special knowledge required by the archetype developers. Translation of the fundamental metrics of Cyclus into more interesting quantities is accomplished in the Cymetric python package. This package is speci cally designed to support the introduction of new metrics by building upon existing metrics. This concept allows for multiple dependencies and encourages building complex metrics out of incremental transformations to those prior metrics. New archetype developers can contribute their own archetype-speci c metric using the same capability. A simple demonstration of this capability focused on generating time-dependent cash ows for reactor deployment that could then be analyzed in di erent ways. Cyclist, a dedicated application for exploration of Cyclus results, was developed. It's primary capabilities at this stage are best-suited to experienced fuel cycle analysts, but it provides a basic platform for simpler visualizations for other audiences. An important part of its interface is the ability to uidly examine di erent slices of what is fundamentally a ve-dimensional sparse data set. A drag-and-drop interface simpli es the process of selecting which data is displayed in the plot as well as which dimensions are being used for

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗