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DOE OSTI2016

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impact of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.

Accident tolerance fuel↗
DOE OSTI2015

EVALUATION OF ENHANCED ACCIDENT TOLERANT FUELS

The U.S. Department of Energy Advanced Fuels Campaign (AFC) is currently funding research and development on a number of fuel technologies for light water reactors (LWRs) that could provide enhanced performance under accident scenarios relative to the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the accident tolerant fuel (ATF) research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. Evaluating the performance of candidate accident tolerant fuel concepts is a topic of significant discussion among researchers, fuel vendors and utilities to ensure that the developed concepts can meet minimum performance and economic requirements, can be developed within a reasonable time, and can be approved for insertion as a lead fuel rod in a commercial reactor by the 2022 goal. Characterizing the performance enhancements of candidate ATF first requires understanding the performance of the current Zr-UO2 system under equivalent operations and accident scenarios. Proper evaluation of each concept is dependent on development of data through focused out-of-pile and in-core experiments to support modeling of the fuel and cladding behavior in fuel performance and systems analysis codes; complimentary irradiation of ATF concepts will be discussed in a separate paper. This paper will provide an overview of the key evaluation tools and evaluation scenarios currently being considered within the LWR community for ATF. The evaluation toolset includes standard neutronic and thermal-hydraulic analysis for normal operating conditions and transient/accident conditions and analysis of severe accident behavior using a modified version of the MELCOR code. Development of advanced fuel performance analysis using the BISON application based on the Multi-physics Object-Oriented Simulation Environment (MOOSE) environment at Idaho National Laboratory (INL) has also been initiated. Evaluation scenarios (e.g. accident scenarios) currently being discussed for use across the international ATF development teams are also presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

U-10Mo Monolithic Fuel Qualification Plan

The Material Management and Minimization (M3) Program’s primary objective within the U.S. Department of Energy/National Nuclear Security Administration is to achieve permanent threat reduction by minimizing and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project, is pursuing a fuel qualification and licensing effort focused on converting high-performance research reactors in the United States from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. The body of this document is focused on defining the activities required for generic qualification of high density, LEU fuel that then allows reactor conversions to proceed for the four high-performance research reactors and the critical assembly that will operate using the U-10Mo monolithic fuel design (see Subsection 1.3): MITR, MURR, NBSR, ATR, and ATRC. Fuel qualification in this report means generic fuel acceptance of the U-10Mo monolithic fuel form by the NRC for use in reactor conversions in the United States, which allows the fuel to be used in subsequent reactor specific licensing requests. The U 10Mo Monolithic Fuel Research, Development, and Qualification Plan utilizes the functions and requirements of the USHPRR Project that was established in the Functions and Requirements Document (F&RD) [1] and expands on these requirements to ensure that planned tests have traceable results that will ensure the requirement has been met. The methods by which data will be collected to show that these requirements have been met are described in this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Nuclear fuel behavior at high temperature: new insights from post-test examinations on the VERDON-1 sample

In order to reduce the uncertainties concerning Source Term assessment under Light Water Reactor (LWR) severe accident conditions, 4 VERDON tests were recently performed. They have been defined in the framework of the International Source Term Program and carried out at CEA Cadarache, in the LECA-STAR facility. Among these 4 VERDON tests, VERDON-1 test was devoted to high burn-up UO{sub 2} fuel behavior and FP release under H{sub 2} reducing conditions at very high temperature. This test was performed with the Release Circuit, which is devoted to the precise characterization of the FP release, in terms of FP release kinetics and fission products (FPs) released fraction. The main components of the circuits are a furnace heating the fuel sample, a filter trapping the emitted aerosols and gamma spectrometry detectors for online and post-test measurements. VERDON-1 test was conducted up to about 2600 Celsius degrees without reaching global fuel relocation, allowing thereby post-test micro-analyses. After a brief description of the fuel sample nature and the experimental set-up, we present the progress of the experimental sequence and the associated gamma spectrometry measurements. We then focus on the major results of the VERDON-1 tests regarding the fuel global behavior. These results are obtained by two complementary techniques. First, the fuel sight gamma station can be used either on-line to detect fuel relocation or just after the test, as a longitudinal bench, to get preliminary information on the global shape of the fuel sample inside the furnace. Secondly, these information are completed by post-test examinations performed on the fuel sample after extraction from the furnace. Complete qualitative and quantitative characterizations, including EPMA (Electron Probe Micro-Analysis) and SEM (Scanning Electron Microscopy) analysis, were performed for the first time at the CEA Cadarache on a sample that has been submitted to such severe accident conditions. These results show an advanced state of sample degradation, with locally previously molten areas, meaning that fuel global relocation was imminent by the end of thermal-hydraulic sequence. The relocation temperature is compared to that of previous VERCORS tests and discussed regarding the atmosphere of the test and possible UO{sub 2}-ZrO{sub 2} interaction. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Selected Profiles of Burnup and Isotopic Concentrations in Pebble Bed High Temperature Reactor Fuel

Pebble Bed High Temperature Reactors (PB-HTRs) feature a unique fuel loading, burnup, and discharge process that is markedly different that a light water reactor or any other batch-loaded reactor that uses fixed fuel elements. During PB-HTR operation, small spherical fuel elements are dropped onto the top of the bed at small and regular intervals, 'trickle' through the core, and pass out of the core via a discharge chute at the bottom. Partially burnt pebbles are circulated back to the top of the core for another pass through the flux field. The process continues until the pebble achieves a specified discharge burnup. This results in high fuel utilization relative to a comparable batch-loaded fuel assembly which may or may not achieve full burnup depending upon where it is placed in the core. On the other hand, the radial and axial variation in the flux field and the random placement of each element on top of the bed mean that there will be a distribution of the final burnup and isotopic concentrations in spent fuel. This leads to some uncertainty in burnup analysis that has consequences for burnup monitoring (and thus efficient operation), source term quantification, and material accountability. The PEBBED code was developed specifically for the design and analysis of PB-HTRs. Although it is still under development, the code captures the important physics of cores fueled by dropping in, and if so configured, recirculating small graphite spheres in which the Tri-Isotropic (TRISO) coated particle fuel is embedded. In this study, the capabilities of the code for computing the isotopic contents of PBR fuel were exploited to examine and compare the variability of selected isotopes, in this case U-235, Pu-239, and Cs-137, in the fuel discharged from different PBR concepts. Results indicate the sensitivity of certain isotopes to the core geometry, mainly the presence of an inner reflector. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A Preliminary Overview of Metal Fuel Models in BISON

BISON is a modern nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. It is a finite element-based code capable of solving the energy, mechanics, and species diffusion equations in a fully-coupled sense. BISON also utilizes thermal and mechanical contact algorithms and variety of nonlinear material models. BISON is applicable to steady and transient fuel behavior and supports the use of one-, two-, and three-dimensional meshes. Although thus far used primarily to simulate light water reactor fuel, BISON is designed to be a multi-fuels code. As such, BISON has been used to analyze TRISO-coated particle fuel and metal fuel in rod and plate form, design and interpret fuel irradiation experiments and investigate novel fuel concepts. This publication provides a preliminary list of metal fuel-related material models recently implemented in BISON. Some preliminary results from a simulation of the EBR-II X447 experiment are shown, which begins to demonstrate how all these models work together. A follow-up publication will be presented with more detail at the ANS meeting in New Orleans during the embedded topical meeting titled 'Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors'. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermal Hydraulics Safety Analyses of TRIGA Reactor Upon Fuel Reconfiguration

Safety analyses of the Training, Research, Isotope, General Atomics (TRIGA) reactor at Kansas State University(KSU) is performed using thermal hydraulic simulations based on changes to reactivity or the volumetric heat generation rate due to changes in the positioning of the enriched fuel elements in the reactor core. If the reactivity or neutronics parameters are known, then they can be used to simulate the thermal energy state of the material which is crucial to assess the safety of the nuclear fuel inside a nuclear reactor. Several thermal hydraulics models and codes have been developed in the past to assess reactor safety. RELAP (Reactor Excursion and Leak Analysis Program) is one such tool widely used throughout the world for safety analyses. Although RELAP is considered a reliable tool by regulators, vendors, and operators, it lacks the numerical fidelity provided by 3-D computational fluid dynamics (CFD) models. This work involves the critical task of assessing safety of a research reactor upon fuel reconfiguration based on models developed in RELAP and comparing them with a commercially available 3-D CFD program. The TRIGA Mark-II reactor, designed by General Atomics and located at Kansas State University, is a class of small pool type inherently safe research reactor licensed to operate at steady state power of up to 1.25 MW. It has a cylindrical core measuring 0.45 meters diametrically and 0.38 meters axially positioned in a tank measuring 1.98 meters diametrically and 6.25 meters axially. It has uranium zirconium hydride (UZrH) fuel rods housed in stainless steel cladding material arranged in five concentric rings (B,C,D,E and F rings, in that order, moving outwards from the center). TRIGA fuel is nominally 20% enriched, but the reactivity of the fuel can be changed by using burnable poisons (not used at KSU), or by varying the total weight percent of uranium in the UZrH fuel meat. The TRIGA reactor is designed to be an inherently safe reactor and is designed to be cooled by natural convection. Thus it is of paramount importance to avoid the film boiling regime; failing to do so will subject the fuel to severe conditions, possibly resulting in its failure due to fuel/clad interactions. A high DNBR ratio is a primary consideration in the reactor core design and thus any modification in the core should ensure that this criteria is always met. In order to fully utilize the capabilities of KSU TRIGA reactor facility, it should be operated at its full capacity, i.e. 1.25 MW, which requires increasing the fuel loading in some of the E- and F-ring elements from 8.5% U (nominal) to 12.0% U (nominal). Only four elements are to be switched to 12% loading, causing local power peaking in these elements. The degree of power peaking to be expected in these 12%-loaded fuel elements was calculated based on a detailed MCNP (Monte Carlo N-Particle code) model of the reactor core. Some preliminary results of the calculations are presented in this paper based on changes in peaking factors and power levels. To predict reactor thermal behavior, axial changes in the peaking factors obtained from MCNP models were used to calculate volumetric energy density functions for a single channel one dimensional RELAP model and a three dimensional COMSOL model of the TRIGA MARK-II reactor. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Exploring the Use of Muon Momentum for Detection of Nuclear Material Within Shielded Spent Nuclear Fuel Dry Casks

The inspection of the storage containers that house used nuclear fuel is an important issue facing the nuclear industry; currently, there are limited options available to provide for even minimal inspections. During the long storage life of the fuel, intermediate handling and transportation activities may take place that can create concerns regarding fuel integrity (safety) and continuity of knowledge (safeguards and security). Accidents and natural disasters also contribute to this concern. The reliable and safe management of used nuclear fuel is among the critical tasks to be addressed for the advancement of fission-based nuclear energy in the United States and elsewhere. Since the early 1950's, when the first nuclear power plant began to produce electricity, approximately 65,000 metric tons of used nuclear fuel have been generated in the United States. The postponement of the opening of a permanent repository until 2048 will lead to increased used fuel accumulation at reactor sites. A rapid increase in storage installations with more than 9,000 casks storing {approx}420,000 assemblies is anticipated by 2050 even if no new nuclear power plants are built. After used nuclear fuel has been placed inside a cask, the cask is sealed, not allowing for visual inspection. Conventional methods, e.g., x-rays, are limited in that they cannot penetrate dense, well-shielded objects. More sophisticated techniques, such as penetrating neutrons or proton radiography, require the use of an expensive accelerator. Efforts by Ziock et al. using neutron and gamma-ray imagers concluded that '...the signatures are deemed insufficient to uniquely identify individual casks of the same type...'. It was found that the gamma-ray and neutron images were dominated by diffusion and thermalization effects resulting from the large amount of shielding. One promising mode of inspection would use cosmic-ray muons. These muons are naturally generated, are able to penetrate dense materials, and are incident from all directions above the horizon. Tomography performed using these muons could potentially provide efficient, inexpensive, remote assessment of used nuclear fuel. However, current muon applications are limited in that they have not included momentum measurement. If momentum measurement could be incorporated, that would represent a big step forward because it would allow more accurate estimation of the muon observables, e.g., scattering variance. This approach has the potential to reconstruct the contents of a cask with improved resolution and identify partial defects, fuel assemblies, and individual fuel rods.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

LOCA and LOFA Safety Margin Analysis for U-10Mo Monolithic LEU Fuel Conversion of the MURR{sup R} Reactor

The University of Missouri Research Reactor (MURR{sup R}) is a multi-disciplinary research and education facility providing a broad range of analytical and irradiation services to the research community and the commercial sector. The source of neutrons for the experimental facilities of MURR{sup R} is a compact 10 MW core fueled with highly-enriched uranium (HEU) fuel. The continued presence of HEU fuel in civilian installations such as research and test reactors poses a threat to national and international security. Therefore, most research and test reactors, both domestic and international, have started a program of conversion to the use of low-enriched uranium (LEU) fuel. MURR{sup R} is one of five U.S. high-performance research and test reactors (USHPRRs) that are actively collaborating with the U.S. National Nuclear Security Agency (NNSA) Office of Material Management and Minimization (M3) to develop a suitable LEU fuel replacement for the currently required HEU fuel. A new type of high-density LEU fuel based on an alloy of uranium and molybdenum is currently undergoing qualification testing, and is expected to allow the conversion of all USHPRRs. As a part of the LEU conversion of MURR{sup R}, an analysis of core behavior under a set of accident conditions for a core fueled with an LEU monolithic alloy of uranium and 10 wt% molybdenum (U-10Mo) has been conducted. This paper presents the newest results for the loss-of-coolant accidents (LOCAs) and loss-of-flow accidents (LOFAs). Other classes of accidents have also been investigated but were not included here. Namely, reactivity insertion accidents (RIAs), fuel failure during operations accidents (FFAs), and fuel handling accidents (FHAs)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates

Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. Here, the swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

New insights into fuel blending effects: Intermolecular chemical kinetic interactions affecting autoignition times and intermediate-temperature heat release

Fuel blending effects on chemically-dominated fuel properties, such as gasoline anti-knock quality, are influenced by fundamental chemical kinetic interactions between the blending agent and the base fuel. Historically, quantification of such interactions has focused on changes to the radical pool, including $\dot{O}$H and HO 2 , while intermolecular interactions pertaining to carbonated, non-fuel-specific intermediates are typically overlooked. In this regard, this work aims to derive new insight into intermolecular chemical kinetic interactions that are intrinsic to fuel blending effects via a case study on blends of 0–30% ethanol (by volume) into FGF-LLNL (a multi-component gasoline surrogate for FACE-F research gasoline) using a rapid compression machine at a diluted/stoichiometric fuel loading, compressed pressure of 40 bar and low- to intermediate-temperature regimes that are representative of boosted SI engine operation. Ethanol blending effects on the intermediate temperature heat release (ITHR) of FGF-LLNL are characterized using experimental measurements, where ethanol is found to promote the extent of ITHR and suppress the transition from ITHR to main ignition. Chemical kinetic modeling is undertaken using a recently updated gasoline surrogate model. Sensitivity analyses on ITHR characteristics further corroborate the ethanol blending effects, and highlight the significant dependence of ITHR on both fuel-specific and non-fuel-specific reactions. An approach allowing comprehensive characterization of the complex intermolecular chemical kinetic interactions between constitutes in a fuel blend is then proposed. Application of the approach to FGF-LLNL/E0–E30 reveals that ethanol perturbs the heat release and autoignition characteristics of FGF-LLNL not only by directly changing the $\dot{O}$H and HO 2 radical pools via fuel-specific reactions, but also through intermolecular interactions where participating intermediates can be produced and consumed by various sub-chemistries. Disabling the intermolecular interactions in carbonated species between ethanol and FGF-LLNL sub-chemistries leads to somewhat slower ITHR evolution and lower ignition reactivity. The role of the individual intermolecular interaction is also characterized using the proposed approach. Lastly, implications of intermolecular interactions for future studies that aim to improve model performance are highlighted, where it is found that further investigations on the core C0-C4 chemistry are needed for developing highly accurate chemistry models for complex fuel blends.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Life-cycle analysis of fuels from post-use non-recycled plastics

Plastic-to-fuel (PTF) technology uses pyrolysis to convert plastic waste—especially non-recycled plastics (NRP)—into ultra-low sulfur diesel (ULSD) fuel. To assess the potential energy and environmental benefits associated with PTF technology, we calculated the energy, water consumption, and greenhouse gas emissions of NRP-derived ULSD and compared the results to those metrics for conventional ULSD fuel. For these analyses, we used the Greenhouse gases, Regulated Emissions and Energy use in Transportation (GREET®) model. Five companies provided pyrolysis process product yields and material and energy consumption data. Co-products of the process included char and fuel gas. Char can be landfilled, which, per the company responses, is the most common practice for this co-product, or it may be sold as an energy product. Fuel gas can be combusted to internally generate process heat and electricity. Sensitivity analyses investigated the influence of co-product handling methodology, product yield, electric grid composition, and assumed efficiency of char combustion technology on life-cycle greenhouse gas emissions. The sensitivity analysis indicates that the GHG emissions would likely be reduced up to 14% when it is compared to conventional ULSD, depending on the co-product treatment method used. NRP-derived ULSD fuel could therefore be considered at a minimum carbon neutral with the potential to offer a modest GHG reduction. Moreover, this waste-derived fuel had 58% lower water consumption and up to 96% lower fossil fuel consumption than conventional ULSD fuel in the base case. In addition to the comparison of PTF fuels with conventional transportation fuels, we also compare the results with alternative scenarios for managing NRP including power generation and landfilling in the United States.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Investigating the potential of a higher reactivity fuel to achieve faster heat-up of aftertreatment systems

Shortening the aftertreatment system heat-up period will be crucial to meet upcoming emissions regulations, which mandate significant reductions in tail-pipe emissions for on-road heavy-duty vehicles. This work studies the impact of increasing fuel reactivity on the combustion of post injections during the expansion stroke. The work demonstrates an operating strategy using a blend of di-butyl ether (a potential low-carbon biofuel component) and #2 diesel that achieves higher exhaust enthalpy, while maintaining similar engine-out emissions (NOx and combustion efficiency) in comparison to the baseline #2 diesel fuel (cetane number 41). On average, a 12% increase in exhaust enthalpy was observed for the blend (cetane number 60) relative to the baseline fuel. The ability to operate the higher reactivity fuel at a more retarded post-injection timing, where combustion with #2 diesel was found to be more unstable, and less complete was key to achieving higher exhaust enthalpy. However, a small fuel penalty (3%–5%) was also associated with the higher reactivity fuel with this operating strategy. The formation of fuel lean regions, due to the long ignition delays at late injection conditions, was predicted to be the leading contributor to the increased emissions of products of partial combustion. Here, chemical kinetic simulations using surrogates for the fuels used in the experiments, demonstrated the ability of a higher cetane fuel to outperform the baseline fuel under fuel lean conditions with equivalence ratios less than 0.7.

09 BIOMASS FUELS↗
DOE OSTI2018

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on “fingerprinting” casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17 x 17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4 He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244 Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Role of Accelerated Burnup Irradiation Testing in Support of Accelerated Fuel Qualification

Accelerated fuel qualification has gained attention as a means to reduce the time needed to realize new nuclear fuel concepts and expand the operating windows of existing fuel forms. A key component of this approach is accelerated burnup irradiation testing. Although the concept of accelerated burnup has been familiar to the community for many decades, the specifics about how the increasing fission rate may be used as a qualification tool have not yet been elucidated. The present work provides a vision of how accelerated fission rate testing can enable accelerated fuel qualification. Technology readiness levels (TRLs) are reintroduced to demarcate the stages of traditional fuel qualification, and accelerated fuel qualification is presented in this context. The critical steps needed to achieve each TRL are reframed within the context of modern nuclear materials research and development, as revolutionary fuel concepts are more common than previous eras. The practical impacts of accelerated fuel qualification approaches as applied to contemporary fuel qualification efforts are illustrated. Examples are given to illustrate how accelerated burnup irradiations are being used currently and could be applied in the future to support qualification and licensure. Finally, outstanding challenges in the application of accelerated burnup methods to nuclear fuel qualification are summarized, with priority placed on understanding how fission rate impacts diffusion, microstructure evolution, and other critical mechanisms that dictate fuel performance.

Accelerated fuel qualification↗
DOE OSTI2019

Preface - JES Focus Issue on Advances in Modern Polymer Electrolyte Fuel Cells in Honor of Shimshon Gottesfeld

The advances in polymer electrolyte fuel cell (PEFC) technologies over the past few decades and particularly over the past few years, have been very impressive. The promise of PEFC-based fuel cell electric vehicles (FCEVs) as long range EVs of zero emissions, high conversion efficiency, and fit for use of fuels from renewable sources, was recognized by pioneers of this technology some 40 years ago, facing at the time significant skepticism. The eventual commercial introduction of FCEVs in 2015 was enabled by the impressive accomplishments of scientists, engineers, and policymakers over that period of time. Among those responsible for the fuel cell community's success, few, if any, have had more impact than Dr. Shimshon Gottesfeld. He is a distinguished fuel cell educator, a passionate fuel cell and alternative fuel advocate, and a pioneering fuel cell innovator, responsible for several critical contributions that helped demonstrate the commercial viability of polymer electrolyte fuel cells. His impact on the field is seen through specific technical accomplishments, the scientists he has mentored, the teams he has led in different topical areas, and his activities as organizer and speaker in technical fuel cell meetings. We are honored to dedicate this focus issue of the Journal of The Electrochemical Society, entitled 'Advances in Modern Polymer Electrolyte Fuel Cells in Honor of Shimshon Gottesfeld,' to Dr. Gottesfeld for his accomplishments in the area of PEFC science and technology.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2018

Navigation Application Programming Interface Route Fuel Saving Opportunity Assessment on Large-Scale Real-World Travel Data for Conventional Vehicles and Hybrid Electric Vehicles

The green-routing strategy instructing a vehicle to select a fuel-efficient route benefits the current transportation system with fuel-saving opportunities. This paper introduces a navigation application programming interface (API), route fuel-saving evaluation framework for estimating fuel advantages of alternative API routes based on large-scale, real-world travel data for conventional vehicles (CVs) and hybrid electric vehicles (HEVs). Navigation APIs, such as Google Directions API, integrate traffic conditions and provide feasible alternative routes for origin-destination pairs. This paper develops two link-based fuel-consumption models stratified by link-level speed, road grade, and functional class (local/non-local), one for CVs and the other for HEVs. The link-based fuel-consumption models are built by assigning travel from many global positioning system driving traces to the links in TomTom MultiNet and road grade data from the U.S. Geological Survey elevation data set. Fuel consumption on a link is computed by the proposed model. This paper envisions two kinds of applications: (1) identifying alternate routes that save fuel, and (2) quantifying the potential fuel savings for large amounts of travel. An experiment based on a large-scale California Household Travel Survey global positioning system trajectory data set is conducted. The fuel consumption and savings of CVs and HEVs are investigated. At the same time, the trade-off between fuel saving and travel time due to choosing different routes is also examined for both powertrains.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2016

Vehicle Technologies and Fuel Cell Technologies Program: Prospective Benefits Assessment Report for Fiscal Year 2016

Under a diverse set of programs, the Vehicle Technologies and Fuel Cell Technologies offices of DOE’s Office of Energy Efficiency and Renewable Energy invest in research, development, demonstration, and deployment of advanced vehicle, hydrogen production, delivery and storage, and fuel cell technologies. This report estimates the benefits of successfully developing and deploying these technologies (a “Program Success” case) relative to a base case (the “No Program” case). The Program Success case represents the future with completely successful deployment of Vehicle Technologies Office (VTO) and Fuel Cell Technologies Office (FCTO) technologies. The No Program case represents a future in which there is no contribution after FY 2016 by the VTO or FCTO to these technologies. The benefits of advanced vehicle, hydrogen production, delivery and storage, and fuel cell technologies were estimated on the basis of differences in fuel use, primary energy use, and greenhouse gas (GHG) emissions from light-, medium- and heavy-duty vehicles, including energy and emissions from fuel production, between the base case and the Program Success case. Improvements in fuel economy of various vehicle types, growth in the stock of fuel cell vehicles and other advanced technology vehicles, and decreased GHG intensity of hydrogen production and delivery in the Program Success case over the No Program case were projected to result in savings in petroleum use and GHG emissions. Benefits were disaggregated by individual program technology areas, which included the FCTO program and the VTO subprograms of batteries and electric drives; advanced combustion engines; fuels and lubricants; materials (for reduction in vehicle mass, or “lightweighting”); and, for medium- and heavy-duty vehicles, reduction in rolling and aerodynamic resistance. Projections for the Program Success case indicate that by 2035, the average fuel economy of on-road, light-duty vehicle stock could be 47% to 76% higher than in the No Program case. On-road medium- and heavy-duty vehicle stock could be as much as 39% higher. The resulting petroleum savings in 2035 were estimated to be as high as 3.1 million barrels per day, and reductions in GHG emissions were estimated to be as high as 500 million metric tons of CO2 equivalent per year. The benefits of continuing to invest government resources in advanced vehicle and fuel cell technologies would have significant economic value in the U.S. transportation sector and reduce its dependency on oil and its vulnerability to oil price shocks.

33 ADVANCED PROPULSION SYSTEMS↗