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DOE OSTI2023

Validation of MFUEL Metal Fuel Performance Models of SAS4A/SASSYS-1

The fuel characterization models of SAS4A/SASSYS-1 (SAS) have recently been extended to include a new U-Pu-Zr metal fuel model, MFUEL. MFUEL is equipped with mechanistic physics based models to predict the pre-transient characterization and transient response of metal fuel, with emphasis on fuel melting, cladding failure, and the metal fuel’s impact on core reactivity. The MFUEL model will be available in the full version of SAS4A/SASSYS-1 5.7, which is scheduled to be released in June 2023. Fast reactor fuel pins that operated in EBR-II and FFTF with low smear density U-Zr and U-Pu-Zr metal fuels and irradiation resistant ferritic-martensitic cladding showed significant advantages in achieving high burnups and assuring inherent safety characteristics during anticipated transients, design basis events, and beyond design basis events. For safety analysis, a fuel performance model must be able to predict (1) Fuel pin mechanics and compositional and dimensional changes, (2) Clad failure, and (3) Fuel pin thermal resistance. Achieving these high level goals accurately is strongly related to the model performance of individual physical processes taking place within a fuel pin during its lifetime. Metal fuels typically operate above the mid-point temperature of melting during normal, as well as off-normal, conditions. At these elevated temperatures, the availability of thermal activation provides a driving force for various diffusional processes leading to complex phase transformations, micro-structure evolution, significant amounts of fuel swelling, interconnected porosity formation, excessive amounts of fission gas release, and fuel clad chemical interactions. Clad failure in fast reactors primarily occurs as a result of creep rupture augmented by clad wastage formation. The reaction is driven by thermal creep induced dislocation motion, grain boundary cavity nucleation, growth and breakup of grain boundaries. The high level complexity and limited available data requires introducing physics-based modeling approaches to gain extrapolation ability and sensitivity with respect to various conditions. The objective of this report is to perform validation of MFUEL using the experimental data for (1) Normal operation EBR-II fuel behavior, (2) Normal operation PHENIX fuel behavior, (3) HT9 Pressure tube ramp-and-hold creep rupture tests, (4) Whole Pin Furnace (WPF) creep strain, creep rupture and eutectic tests, (5) Fuel Behavior Test Apparatus (FBTA) eutectic tests, and (6) TREAT M5-7 OverPower tests up to clad failure. Section-2 includes a brief description of the MFUEL models. A detailed description of the MFUEL physics-based, semi-empirical models will be presented in the SAS V 5.7 theory manual. Section-3, 4, and 5 describes the validation effort for the pre-transient irradiation, furnace transients, and TREAT M-Series transients, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Survey of Uranium Nitride and Mixed Oxide Fuels for Microreactor Applications

Different fuel systems are being proposed for use in microreactors including TRISO fuel and metallic U-Zr based fuel. TRISO fuel is considered the main fuel option for most of the industry teams. Given the need for compact size core and potential for long fuel life of a microreactor, high density fuels such as metallic and nitride fuels are potential options. Of interest here is the uranium nitride (UN) option, which can allow for higher fissile material content and higher thermal conductivity compared to conventional uranium oxide fuel, and also has higher melting temperature compared to metallic fuel. Meanwhile, the current availability is limited for high assay low enriched uranium (HALEU) that is needed for high density fuels, which motivates the consideration of using Pu as a potential replacement for HALEU until adequate production capacity is in place. Current Pu availability is mainly attributed to the inventory of excess weapons Pu rather than through reprocessing of spent nuclear fuel. This inventory of excess Pu can be used in both metallic and oxide fuel systems to replace HALEU. Of interest here, the oxide form, that is the mixed oxide form of PuO 2 and UO 2 (MOX). In this report, the options of using UN or MOX, in both pellet and TRISO fuel forms, in microreactors are evaluated in relation to their properties, fuel performance and irradiation data, as well as fabrication. Gaps related to those areas are identified for both fuel systems, to guide future activities by DOE programs such as the advanced fuels campaign (AFC), to enable their use in microreactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Characterization of MC&A for the Molten Salt Fuel Cycle

Advanced reactor developers are exploring diverse reactor designs, including molten salt reactors (MSRs). These advanced reactors are considered for wider applications and a range of deployment locations, including supporting the integration of renewable energy sources in the grid. There are three main types of MSRs: (1) reactors in which the fuel salt freely circulates within the core; (2) reactors with the fuel salt contained within vented fuel tubes; and (3) reactors that use molten salt solely as a coolant, with the fuel in a separate, solid form. In this document, the term MSR refers specifically to the first two types, which use fuel salt—special nuclear material (enriched uranium, plutonium, and 233 U) in chloride or fluoride form mixed with chloride- or fluoride-based carrier salt in a peritectic mixture—as the primary medium for fission. The composition of fuel salt, both at startup and for makeup or refueling, varies depending on the MSR design and the chosen fuel cycle approach, which can be either once-through or closed. For MSRs, a variety of fuel cycle approaches (e.g., U, U–Pu, U–Pu–TRU, U–Th, U–Pu–Th) are being considered. Fuel in MSRs is much different than traditional solid fuel, including its preparation. The uniqueness warrants investigation into characterizing fuel preparation processes, known as fuel salt synthesis . This effort characterized major fuel preparation and synthesis processes, identifying temperature, equipment, and environmental requirements for uranium-, plutonium-, and thorium-based fuel preparation and synthesis. Because MSR fuel salt synthesis facilities handle special nuclear material in loose, bulk form, a material control and accounting plan will be required for licensing from the US Nuclear Regulatory Commission or under the US Department of Energy authorization. This effort serves as a foundation to investigate material control and accounting approaches for synthesis facilities, including determining measurement points and techniques. Because several MSR developers are planning demonstration facilities in the coming years, this effort will support stakeholders with preparing or reviewing material control and accounting plans for providing assurance that all special nuclear material is accounted for at fuel salt synthesis facilities. This report was produced for Materials Protection, Accounting, and Control Technologies (MPACT) program under the US Department of Energy (DOE), Office of Nuclear Energy, Nuclear Fuel Cycle and Supply Chain.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

An Overview of Material Models in the BISON Fuel Performance Code

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and supports the use of one-, two-, and three-dimensional meshes. BISON is designed to be a multi-fuels code. Although primarily applied to Light Water Reactor (LWR) fuel to date, the code 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. Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models which are often complex and nonlinear. Nearly all of the material models currently available in BISON are empirical, using correlations from existing fuel performance codes or material property libraries. Although these models perform adequately under many conditions, they have well-understood limitations. One obvious weakness is that the models are only valid in the domain of the associated experimental data, and extrapolating beyond this domain can result in substantial errors. A further weakness is that these models are often correlated in terms of imprecise quantities, such as burnup. Materials can, for example, reach a given burnup under very different circumstances, leading to very different material behavior. To overcome these weaknesses, efforts are underway with BISON to develop models based on a multi-scale approach, where material behavior is described with a series of interdependent microstructure based models. This paper provides an overview of the material models available in BISON. To keep the scope manageable, only LWR fuel will be considered. The first section focuses on UO{sub 2} fuel, identifying the models generally required and briefly describing those that have been implemented in the code. The second section repeats this approach for Zircaloy cladding. A final section provides examples of areas where multi-scale material modeling has been applied to BISON. BISON is a modern 2D and 3D nuclear fuel performance code that is applicable to multiple fuel forms. A wide variety of complex and nonlinear material models are required to simulate fuel behavior. Focusing on UO{sub 2} fuel and Zircaloy cladding (LWR fuel), the set of empirical models currently implemented in BISON have been reviewed. To overcome some of the deficiencies associated with empirical models, multi-scale material models are being developed. Two multi-scale models that have been used in BISON were reviewed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Z4B, AREVA NP's most recent material for structural fuel assembly components

Z4B is AREVA NP's advanced material for fuel channels and BWR fuel assembly structural components such as water channels, which are made of zirconium base material. As such, Z4B is envisaged to replace Zircaloy-2 and Zircaloy-4 as a fuel channel and water channel material. The chemical composition of Z4B is similar to that of Zircaloy-4 with the exception of an increased iron and chromium content slightly above the ASTM specification. The manufacturing processes applied to produce Z4B components are the same as those that have been applied for Zircaloy-2 and Zircaloy-4 materials. AREVA NP has irradiated a total of 74 lead fuel channels made of Z4B either in beta-quenched condition or in fully recrystallized condition in eight BWRs worldwide. Sixty-six of these are still in service with 26 of these lead fuel channels on fuel assemblies of various ATRIUM 10 designs and 40 on ATRIUM 11 lead fuel assemblies. In addition to Z4B fuel channels, the ATRIUM 11 design also features fully recrystallized water channels. The first ATRIUM 11 lead fuel assemblies were inserted in-core in 2012. The first Z4B beta-quenched lead fuel channels were inserted in 2009 and will reach an estimated burnup of 60 MWd/kgHM in 2016 after seven annual cycles of irradiation. The first Z4B fully recrystallized lead fuel channels were also inserted in 2009 and have been discharged with EOL burnups of about 47 MWd/kgHM after three biennial cycles of irradiation. Pool-side inspections of these leads have confirmed the superior uniform and shadow corrosion behavior of Z4B fuel channels relative to Zircaloy-4 fuel channels. This increased corrosion resistance combined with a similar hydrogen pickup fraction to Zircaloy-4 will lead to an overall decrease in total hydrogen content and minimize the potential of hydrogen assisted irradiation growth and abnormal bow of fuel channels in the recrystallized condition. The use of Z4B fuel channels in the beta-quenched condition decreases bow even further due to the resulting quasi-isotropic microstructure. Additionally, fuel channels made of Z4B tend to exhibit reduced bulge, which is valid in both conditions, beta-quenched and recrystallized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

DOE Advanced Gas Reactor Fuel Development and Qualification Program

Breakdown of the DOE Advanced Gas Reactor Fuel Development and Qualification Program. Including discussion topics on TRISO technology Status Circa 2000, New Production Reactor (NPR) Fuel Experience, Fuel Qualification, US DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program, Initial AGR Program Reference HTGR Design, Fuel Qualification Approach, Fuel Performance Modeling, Fuel Fabrication, Selected AGR-1, AGR-2, and AGR-5/6/7 Fuel Property Means, AGR Program TRISO Fuel Key Performance Data, Irradiation Performance: Fission Gas R/B, Irradiation Testing Results, Kernel and Coating Behavior During Irradiation, Locating and Studying Failed Particles Greatly Improves Understanding of Fuel Performance, Fission Product Release from UCO Fuel Compacts: AGR-1 and AGR-2 Examples, HTGR Accident Safety Testing of TRISO Fuel, Evaluating Behavior During D-LOFC Accidents, Safety Test Results for US UCO Fuel, Particle Failure Evaluation, Fuel Performance Summary, Ongoing Work and Outstanding Data Needs, Core Oxidation, Industry Engagement, and Coated-Particle-Fueled Reactor Concepts and Fuel Designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Dry Storage of Encapsulated Defective Fuel Rods at Doel 1 and 2 in Belgium - 17581

The project of removing all fissile and non-fissile material out of the spent fuel ponds of Belgian Doel power plant is on-going and in that framework, a solution was needed for evacuating all used fuel elements from Doel 1 and 2 storage pool, including Defective Fuel Assemblies (DFA) and Defective Fuel Rods (DFR). Following the moving out of storage pool, all used fuel elements are intended to be loaded in dry storage casks, this used fuel management option being the one implemented today in Belgium while waiting for final management. Synatom, in charge of all fuel cycle related activities for the nuclear power plants of Belgium and Engie-Electrabel, the operator of both Belgium power plants (Doel and Tihange), had then to find an industrial solution for emptying Doel 1 and 2 used fuel pool and loading DFR in dry storage casks. After having thoroughly reviewed the possible options for implementing dry storage of DFR, Synatom selected AREVA technology. It consists in: - encapsulating each DFR by loading the fuel rod in a capsule, drying the fuel rod using a process which allows to physically verify with high accuracy the residual water content inside the DFR, and welding the capsule, - assembling the capsules in a capsule canister (a skeleton with external dimensions similar to fuel assembly's) and - transferring this capsule canister along with intact fuel assemblies in an associated dry transport and storage (dual purpose) cask, as already used for interim storage of used fuel at Doel site. The French as well as the Belgian competent authorities are parties to the licensing process project. The purpose of the paper is to present the on-going industrial project related to dry storage of dried and encapsulated DFR - along with intact used fuel in dual purpose casks - at Doel 1 and 2 in Belgium, its challenges and progress. The paper presents: - why the drying method, welded capsule and associated cask is the best available technology for dry storage of DFR; - a focus on fuel assembly repair operations (initial step before encapsulation) per-formed in 2015 which consisted in successful extraction of four DFR from their original fuel assembly, using proven and experienced methods and tools; - in conclusion, an update of Doel 1 and 2 DFR dry storage project: reaching of major qualifications and licensing milestones and overview of the project perspectives. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Damaged Fuel Rod Service with German Turnkey Technology NCS 45 Transport Casks and Under-Water-Brazing Technology - 18182

Transportation of damaged irradiated fuel rods with the package NCS 45 developed and operated by DAHER Nuclear Technologies GmbH (DNT) saves considerable technical and financial resources in the German phase-out scenario. The technology can be used world-wide to speed up decommissioning of NPPs. In each German reactor (and most likely in all reactors in the world) a small number of damaged fuel rods have accumulated over the operation phase. In general, transport and storage solutions for fuel are concentrated on the bulk of material, i. e. complete fuel assemblies containing undamaged fuel rods. It is hence a rather straightforward process to remove almost all the fuel from the reactor pools. However, damaged fuel rods - which have to be removed from the pool prior to dismantling - can cause considerable delay in dismantling and high costs for safe-guarding and specialized solutions, even though they only contain a comparatively very tiny amount of fuel. DNT has all required tools, equipment and experience to mitigate the described problem: - The package NCS 45 licensed for the transport of damaged Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel rods with a burn-up of up to 120 GWd/MgHM; - A patented underwater brazing (encapsulation) technology to render very high burn-up damaged fuel rods fit for transport; - Qualified personnel to plan, license and operate such encapsulation process and the NCS 45 in a NPP and to carry out the transports; - Proven record of successful removal of damaged fuel rods from several shut-down reactors in Germany. The package NCS 45 is currently licensed in several European countries. Due to its state-of-technique design - type B(U)F-96 approval - it could be licensed and used for world-wide applications. The underwater brazing technology was developed and patented by DNT to address the challenge of transporting high burn-up and/or damaged UOX and MOX fuel rods. Above a burn-up of 55 GWd/MgHM there are neither enough data to assess the behavior of the cladding under stress nor sufficient benchmarks available to validate the source terms of the fuel. Hence, high burn-up fuel rods are encapsulated in the fuel pool by using the underwater-brazing-technology before they are loaded into the package NCS 45. Furthermore, the capacity of the NCS 45 to transport damaged fuel rods can be considerably enhanced by encapsulating these rods. (authors)

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

Enhanced Thermal Conductivity UO{sub 2}-BeO and UO{sub 2}-SiC Fuels Behavior

Commercial light-water reactors currently use UO{sub 2} as a nuclear fuel, which has demonstrated several desirable characteristics. However, UO{sub 2} has a low thermal conductivity that leads to the development of a large temperature gradient across the fuel pellet. This temperature gradient contributes to limits in the operational performance of the reactor due to effects that include thermal stresses causing pellet cladding interaction and the release of fission product gases as demonstrated. These high fuel temperatures can be decreased and reactor performance improved by developing an enhanced thermal conductivity nuclear fuel. A high thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling as described. Additionally, fission gas release would be decreased allowing for higher fuel burn-up, and the safety of the reactor would be improved with a faster thermal response and less stored energy in the fuel pins. A ceramic-ceramic composite nuclear fuel has an increased effective thermal conductivity due to adding a second compatible and high thermal conductivity ceramic to UO{sub 2}. SiC and BeO are two materials that demonstrate compatibility with UO{sub 2} and have a high thermal conductivity. In summary, the enhanced thermal conductivity UO{sub 2}-BeO and UO{sub 2}-SiC fuel performance has been modelled in a light water reactor, and a more complete and detailed understanding was achieved for enhanced thermal conductivity UO{sub 2}-BeO and UO{sub 2}-SiC fuels. BeO and SiC have high melting point, low neutron absorption, low thermal expansion coefficient and chemical compatibility with current UO{sub 2} fuels which makes them a suitable material for thermal conductivity enhancement. In our modeling and simulation work, the enhanced thermal conductivity nuclear fuel showed the capability of decreasing the fuel temperatures, increasing the gap total conductance, reducing the fission gas release, decreasing the gap/plenum pressure, lessening oxygen redistribution, and facilitating a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling. The safety of the reactor operation would be then improved. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Reactor Physics Analysis of Transitioning to a Thorium Fuel Cycle with Molten Salt Reactors

Commercial power reactors operating in the United States are light-water reactors (LWRs) fueled with low-enriched uranium (LEU) in a once-through fuel cycle; the important fissile isotopes are {sup 235}U, {sup 239}Pu, and {sup 241}Pu. In a thorium fuel cycle, the fertile isotope {sup 232}Th is converted into {sup 233}U, a fissile isotope that is not naturally occurring. In transitioning from the once-through fuel cycle to the thorium fuel cycle, the availability of man-made {sup 233}U presents a challenge: without it, the initial fissile material fueling the first reactors built in this transition is not the fissile isotope that the reactor is breeding. The transition requires an alternate fissile material until enough surplus {sup 233}U is available to start up additional reactors. In liquid-fueled molten salt reactors (MSRs), the fuel is dissolved into molten salt, which is continuously circulated through the core, simultaneously undergoing irradiation, chemical treatments and separations, and feeds. MSRs are an attractive technology in which the thorium fuel cycle can be deployed, as the {sup 233}U bred from {sup 232}Th is continuously being recycled back through the core. The bred fissile material is immediately available to fuel the core without processing or fuel fabrication. After start-up, only a fertile {sup 232}Th feed is necessary to sustain criticality in the MSR. But, in single-fluid MSRs, the initial fissile and fertile materials are mixed into the same molten fuel salt, complicating separations of initial and bred fissile materials. In solid-fuel systems, these materials are often separate. This analysis herein builds on previous work modeling thorium-fueled MSR start-up. It is based on unit cell representations of the graphite-moderated, 2250 MWt Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR). This paper discusses the start-up of the first MSR in the transition to a thorium fuel cycle and identifies challenges and limitations of fuel cycle deployment with this reactor technology. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Advanced Simulation Aided Metallic Fuel Design

Nuclear engineering is reaching a domain space in which computational power and methods are starting to surpass the models that are used to simulate nuclear fuel rods. As these models are updated, simulations utilizing different compositions, operating conditions, fabrication techniques, and geometries can provide core designers with an important first step towards fuel qualification. Due to the cost associated with fuel irradiations, it can be expected that time and money spent building better fuel models may provide an attractive return during initial studies of advanced fuels, especially with regards to fuels beyond the extensively studied uranium oxide system. In general, the primary U.S. experience with irradiated fuel rods beyond traditional oxide fuel is with binary (U-Zr) and ternary (U-Pu-Zr) metal fuels. The bulk of the experimental data for metal fuels is obtained from rods irradiated in the EBR-II sodium fast reactor (SFR), thus many of the correlations developed for metal fuels comes from the post-irradiation (PIE) examination of EBR-II rods. One of the unique characteristics that set U-Pu-Zr fuels apart from traditional oxide fuels is the presence of different phases that develop in the fuel during irradiation. Each of these different phases has unique thermal, mechanical, and diffusive behavior. The transitions between these phases are not static as the initially homogenous zirconium diffuses towards the center and edge of the fuel rod, resulting in a low-Zr, high-U ring. Along with the differing material properties that occur in each phase, the increase in uranium corresponds to an increase in localized fission and power production. The direct coupling between the thermal and diffusion solves highlights one of the many examples of tightly coupled phenomenon that are present in nuclear fuel, and motivation for the development of codes that simultaneously solve each piece of physics in a fully-coupled manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A numerical investigation on methane combustion and emissions from a natural gas-diesel dual fuel engine using CFD model

Natural gas (NG)-diesel dual fuel engines have been criticized for their high emissions of unburned methane. The past research on methane emissions from dual fuel engines has focused on the measurement of methane concentration in exhaust gases. The development of approaches capable of minimizing methane emissions requests the detailed spatial distribution of methane in-cylinder during the combustion and post combustion processes. However, it is difficult to experimentally measure the spatial distribution of methane in-cylinder. Here this research presents a numerical study on the combustion process of a NG-diesel dual fuel engine using the computational fluids dynamics (CFD) model CONVERGE coupled with a reduced primary reference fuel (PRF) mechanism. The model was validated against the heat release process and the emissions of nitrogen oxide, methane and carbon monoxide measured in a single cylinder dual fuel engine. The validated CFD model was applied to investigate the combustion of methane and n-heptane and the spatial distribution of methane in the dual fuel engine. This is most likely the first attempt to visualize the spatial distribution of methane in dual fuel engines using CFD. The objective of this study is to numerically simulate the methane combustion process, especially the methane present outside the pilot spray, quantify the methane combustion in each combustion stage, and visualize the spatial methane distribution in cylinder. The results showed that the momentum produced by the pilot fuel injection and combustion pushed the combustion products of pilot fuel and methane within the pilot spray plume toward the unburned methane-air mixture. Such a movement enhanced the mixing of the hot combustion products and the relatively cold unburned methane-air mixture during the main combustion process and dominated the combustion of methane presented outside the pilot fuel spray plume. Based on the simulation results at a low load condition (4.05 bar), the main combustion process consumed 43–53% of the methane fumigated into the intake mixture. The post-combustion oxidation process consumed 17–29% of the intake methane, which was 36.2–51.8% methane that survived the main combustion process. In comparison, 27–35% methane emitted the engine without participating the combustion process. The unburned methane at exhaust valve opening was mainly observed at the center of the cylinder. In comparison, the contribution of the crevice and boundary layer around the cylinder liner to methane emissions was relatively small. The slip of methane through the dual fuel engines was due to the fact that the premixed mixture was too lean to support the propagation of the turbulent flame initiated by the pilot fuel and the lack of pilot fuel vapor reaching the center of the combustion chamber because of the geometric limitations of the fuel injection system and the reduced mass of pilot fuel injected into the cylinder. The approaches aiming to enhance the combustion of methane and minimize methane emissions from dual fuel engines should focus on those capable of increasing the volume of pilot fuel vapor formed after injected into the cylinder.

42 ENGINEERING↗
DOE OSTI2025

Development of the Well-to-Wake Life Cycle Assessment Model for Marine Fuels in R&D GREET 2025

The Marine analysis capabilities in the 2025 Research and Development Greenhouse gases, Regulated Emissions, and Energy use in Technologies (R&D GREET) have been updated to incorporate the latest findings from the Fourth International Maritime Organization (IMO) GHG Study.1 To support these updates and enhance user flexibility, the former single marine module used in previous R&D GREET versions has been split into two workbook tabs: The Marine_Fuel tab, which covers the fuel-cycle, and the Marine_Trip tab, which covers the trip-cycle. In the Marine_Fuel tab, fuel cycle results for 39 marine fuels are available—assuming a slow speed diesel engine—and can be seen in Appendix A. However, there are options for changing feedstocks for multiple fuels, ability to view results with or without pilot fuels for fuels such as methanol and ammonia, and carbon capture options. All these combinations make the number of total possible pathways over one hundred. For each pathway, life cycle emission results are divided into feedstock, conversion, and combustion categories, along with pilot fuel supply chain emissions for fuels used in dual fuel engines. In the newly created Marine_Trip tab, users can view the life cycle emissions of using these fuels from a trip perspective. Compared to R&D GREET 2024, 16 new types of vessels are added in R&D GREET 2025. Examples of results assuming the default size for each vessel type are available in Appendix A. Additional information categories on vessels are included such as vessel size, main propulsion type, engine power rating, combustion cycle, etc. based on the Fourth IMO GHG Study. Operational mode of the trip, fuel consumption, and load calculation were updated based on the report. The fuel consumption calculation also incorporates low-load adjustment, speed–power correction, weather correction, and fouling correction factors. Furthermore, fuel consumption in boilers or steam turbines is also added for the first time, besides fuel use in main propulsion engines and auxiliary engines. This effort makes the R&D GREET Marine capabilities more robust than before. Users will be able to estimate fuel consumption more accurately for more diverse choices of vessel categories, vessel sizes, and energy converters.

09 BIOMASS FUELS↗
DOE OSTI2017

Microalgae Feedstocks for Aviation Fuels

There is significant global interest in developing, testing, and using alternative jet fuels for both commercial and military use in an effort to create a sustainable and stable fuel supply while reducing greenhouse gas emissions. Currently, the aviation industry is entirely dependent on a finite-supply of petroleum based fuel sourced in part by politically and economically unstable regions of the world. Commercial jet fuel use within the contiguous United States (CONUS) was 17.8 billion gallons per year (BGY) in 2009, while jet fuel use in 2010 by the U.S. Air Force (USAF), Navy, and Army was 1.5 BGY, 0.6 BGY, and 0.8 BGY, respectively (Carter et al., 2011). U.S. commercial and military aviation sectors have set ambitious near-term alternative fuel and environmental performance targets. This includes a tentative Federal Aviation Administration (FAA) goal of 1 BGY alternative fuel use by commercial aircraft by 2018. The USAF has set a target of 50% for USAF domestic aviation via alternative fuels by 2016 (0.73 BGY), and 50% of the Navy’s total energy consumption afloat (0.3 BGY) will come from alternative fuels by 2020 (Carter et al., 2011). If these targets become policy, at least 2 BGY of domestically-produced alternative jet fuel will be required by 2020. The Energy Independence and Security Act (EISA) of 2007 established production requirements for domestic alternative fuels under the Renewable Fuel Standard (RFS). For example, 36 billion gallons of renewable fuel must be produced by 2022, of which 21 billion gallons shall be advanced biofuels. EISA defines advanced biofuels as non-corn starch derived biofuels having lifecycle greenhouse gas emissions 50% lower than gasoline. There a number of potential fuel pathways for meeting the RFS. One of these is biomass-based diesel, including jet fuel (Schnepf and Yacobucci, 2013). The U.S. Department of Energy (DOE) Bioenergy Technologies Office (BETO) has a stated goal in its 2013 Multi-Year Program Plan (USDOE, 2013) to support the RFS through development of “…commercially viable biomass utilization technologies to encourage the creation of a new domestic bioenergy industry…”. BETO has also recognized the potential for aviation biofuels to support the bioenergy industry, seeing drop-in bio-based jet fuels one of the viable alternatives for the aviation industry and the military to meet their ambitious near-term GHG reduction targets (USDOE, 2014). One of the important Multi-year Program Plan Targets (USDOE, 2013) is to establish feedstock resource assessment models to evaluate the geographic, economic, quality and environmental criteria for which 20 million metric tons of ash free dry weight (AFDW) algal biomass can be produced by 2022. Toward meeting the EISA requirements, algal biofuels may offer a number of advantages. They can produce a range of biofuel feedstocks suitable for diesel and aviation fuels. Microalgae, on a strain-specific basis, can be cultivated using impaired water including saline, and/or brackish pumped groundwater or seawater, treated industrial wastewater, municipal sewage effluent, and produced water generated from oil and gas drilling operations. Additionally, microalgae require nitrogen and phosphates as essential nutrients and could provide water treatment co-benefits to municipalities, industry, and the environment.

algae↗
DOE OSTI2013

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

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

Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in this paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light-water reactor approaches an equilibrium composition after 20 depletion steps, demonstrating the potential for the longer time scales required to achieve equilibrium for solid-fueled systems over liquid fuel systems. This time to equilibrium can be reduced by starting with an initial fuel composition closer to that of the equilibrium fuel, reducing the need to handle time-dependent fuel compositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Fuel Fabrication Facility Study for FCF HALEU

Nuclear energy technology developers seek to bring to market new advanced reactor designs and new light water reactor fuel designs, many of which would deploy fuel made from uranium with 235 U enrichment ranging from 5% to 19.75%. There is currently no commercial supply of uranium in this enrichment range, which is higher than used in today’s light water reactors but still lower than the safeguards limit of 20% 235 U. This limitation is a barrier to development of demonstration of these new nuclear energy technologies. As a means to address this near-term need, DOE and Idaho National Laboratory (INL) has considered how the HALEU product from sodium-bonded spent fuel treatment in the INL Fuel Conditioning Facility (FCF), located at the Materials and Fuels Complex (MFC), might be provided and fabricated into some of the new fuel designs. The evaluation presented in this report addresses the potential establishment of two types of fuel fabrication capability at INL, represented generically as a ceramic-type (or pellet-type) fuel fabrication line and a metallic fuel fabrication line. Because the FCF HALEU feedstock contains residual transuranic contamination and a relatively small amount of residual fission products, fabrication with that material is necessarily performed inside of engineered confinement barriers, such as gloveboxes, which would be filled with inert gas for most of the fuel types considered. The throughput rate of the fabrication lines envisioned is 2.5 metric tons of HALEU fuel per year, which is judged to be sufficiently representative of engineering-scale. Space in three existing buildings was identified for consideration in this study of pre-conceptual equipment layouts. The scope and nature of the work needed to prepare each of the three buildings for each type of fuel fabrication was identified (though not in engineering detail), and three-dimensional models of envisioned fabrication equipment were laid out in each of the buildings to assess adequacy of space. The study concluded that either type of fuel fabrication line can feasibly be installed and operated in either of the three buildings. However, each of the three buildings presents different challenges in preparation, due to accommodating current missions that would need to be relocated or necessary decontamination and decommissioning work to free-up space for fuel fabrication, or due to complications of accessibility. Rough order-of-magnitude (ROM) cost estimates for preparing buildings begin at $10M ( 20%/+50%) and would increase with increased complexity of building modifications. The cost of fabrication equipment and gloveboxes is roughly estimated to be $21M (-20%/+50%) for a metallic fuel fabrication line and $27M (-20%/+50%) for a ceramic/intermetallic fuel fabrication line. Nominal operating costs are estimated to be $7M per year (in 2019 dollars), increasing from lower amounts needed as the operating crew is being assembled and trained.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗