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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Emergency response and solutions for damaged fuel assembly handling events - Lessons learned from a 2013 fuel positioning issue

In 2013, during the refueling of a French 1300 MWe reactor, a fuel assembly (FX3599) was damaged by a fuel assembly positioning tool in the reactor vessel. The upper part of a corner rod of the fuel assembly was twisted and the top end grid was torn off. Because of the protruding fuel rod, the assembly could not be easily transferred for repair to the fuel storage pool via the transfer tube. This unexpected and rarely encountered situation was assessed jointly by EDF and AREVA NP. The chosen option was to directly perform the field service inside the reactor building in cells generally used for intermediate storage during refueling. This operation, performed by EDF and AREVA, involved removing the twisted rod and positioning it in a special-purpose canister by means of a gripper-type tool for removal (KIPCI gripper) and a gripper for holding the rod in place. The special-purpose canister was designed, ordered and supplied by a sub-supplier within one week. The operation went through 3 phases. First a preparation step in the fuel building: The stripper was disassembled and technical checks carried out to ensure correct operation. The gripper tightening force was checked in real-life conditions with the gripper suspended from auxiliary crane. The fuel rod removal was simulated several times by trained operators. Second, the equipment was installed in the reactor building. And third, the operation took place in the reactor cavity: the special-purpose canister was opened, the radioactive waste canister placed in the special-purpose canister, the twisted fuel rod was then grasped by the KIPCI gripper and removed from the fuel assembly. The twisted fuel rod was positioned above the canister and dropped into it. Once the fuel rod had been removed, the canister and fuel assembly FX3599 were transferred to the fuel building and refueling resumed. The canister containing the fuel rod is currently in interim storage in the reactor pool until appropriate action is taken for it to be shipped and eventually reprocessed. EDF has performed improvement in its handling procedures and in its handling tools to avoid a similar new event.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effects of fuel thickness on thermal and mechanical performance of U10-Mo monolithic fuel plates

Monolithic fuels are being considered for conversion of high performance research and test reactors. These plate-type fuels are comprised of a low enrichment, high density U-Mo alloy fuels within an aluminum cladding. Although the plates have demonstrated satisfactory performance; still, use of a high density fuel in a foil form retains technical challenges that could affect the overall performance. To understand performance of this new design during fabrication and irradiation, the plates have been evaluated for various geometrical and operational variables. As a part of these set of parametric sensitivity studies, effects of foil thickness on performance were studied. Because high performance research reactors will utilize different fuel thicknesses, it is necessary to evaluate possible effects for a range of thicknesses that are being considered for various designs. Based on the preliminary design specifications, the fuel thicknesses were varied between the limiting cases. The bounding fuel thicknesses were 0.203 mm as minimum that is projected for ATR, and 0.635 mm as maximum that is projected for MITR. To study possible effects, a behavioral model was developed for a selected plate from RERTR-12 experiments and the plate was simulated with as run irradiation history. The simulations were repeated for a range of fuel thicknesses, while keeping the cladding thickness and the operational parameters the same. The results have indicated that the plates with thicker fuels would have higher temperatures, deformations and shutdown stresses. To investigate the effects of fuel thickness exclusively, operational parameters were scaled. In particular, a second set of simulations with prorated volumetric heat generation rates were performed. The results indicated that the fabrication stresses in the fuel foil decrease with an increasing fuel thickness. On contrary, irradiation stresses of the fuel at shutdown are higher for the plates with thicker fuels. Greater peak deformations occurred in plates with thicker fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Potential Impacts of High-Octane Fuel Introduction in a Naturally Aspirated, Port Fuel-Injected Legacy Vehicle

In recent years there has been an increased interest in raising the octane level of gasoline to enable higher compression ratios (CR) in spark-ignition engines to improve vehicle fuel efficiency. A number of studies have examined opportunities to increase efficiency in future vehicles, but potential impacts on the legacy fleet have not received as much attention. In this study, our effort focused on experimental studies on an engine using high-octane fuels without changing the engine’s CR. Spark timing was advanced until maximum torque was reached or knock was encountered for each engine condition, using each individual fuel to maximize engine efficiency. Knock-limited conditions occurred as the output brake mean effective pressure (BMEP) neared the maximum attainable output at a given engine speed. Increasing research octane numbers generally enabled knock-free operation under a greater number of operating conditions. Vehicle modeling using Autonomie was used to project vehicle energy use, fuel economy, and tailpipe CO 2 emissions for the Urban Dynamometer Driving Schedule (UDDS), the Highway Fuel Economy Test (HWFET), and the US06 cycle. Results show that decreases in energy consumption of up to 2% for a small SUV are possible through the use of a 97 RON fuel compared to a baseline using 91 RON fuel, provided that the formulation of the fuel does not cause unanticipated operational issues such as lower maximum BMEP. Greater improvements using high-octane fuels are possible if the CR is increased, but there is no opportunity to increase the CR in legacy vehicles. Thus, these vehicles realize an improvement from increased octane rating in accordance with their ability to spark advance to take advantage of a fuel with a higher octane rating. For the modeled vehicle, improvements of up to 2% in volumetric fuel economy may be possible through the use of a 97 RON fuel with the largest gains expected on the US06 cycle. Fuel economy impacts are strongly coupled to the heating value of the fuels in addition to changes in engine efficiency. Similarly, decreases in tailpipe CO 2 emissions are also achievable. However, simultaneous improvements in energy consumption, fuel economy, and tailpipe CO 2 emissions are not guaranteed and are dependent upon fuel formulation.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2019

Fuel performance evaluation of annular metallic fuels for an advanced fast reactor concept

The annular metallic fuel design has been considered as a promising solution to relieve the fuel-cladding mechanical interaction (FCMI) caused by significant swelling of metallic fuels at high burnup. In this paper, the fuel performance of annular metallic fuel with HT9 cladding in a novel advanced fast reactor design optimized for high-burnup performance was evaluated. The BISON code was utilized to perform fuel performance simulations using a modified fuel swelling correlation for conservative evaluation. Three representative fuel pins, which feature highest peak burnup (26.8 % FIMA, fission per initial metal atom), highest fuel temperature (645 degrees C), and highest fast neutron flux (2.68 x 10 15 n/cm 2 ∙s), were investigated. A series of key fuel performance parameters, such as peak fuel temperature, reduction of fuel central void, and cumulative damage fraction (CDF) of the cladding, were predicted for both upper and lower plenum configurations. It is evident that annular fuel design with a lower smeared density is capable of accommodating the significant swelling of metallic fuel at high burnup. Finally, the adoption of a lower plenum configuration effectively lowers the gas pressure inside the rod and reduces the creep damage accumulation in the HT9 cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development and preliminary validation of a mechanistic multiscale model for fuel-cladding chemical interaction in metallic nuclear fuels

Despite decades of fuel rod material and design improvements, fuel-cladding chemical interaction (FCCI) remains the single-most lifetime-limiting behavior for modern metallic fuel rods. Constraining fuel lifetime increases operating costs, limiting the economic viability of commercializing metallic nuclear fuel technology. A mechanistic multiscale model utilizing the finite element method-based MARMOT and BISON codes was developed to more confidently predict cladding-side FCCI and its impact on fuel performance. The new BISON model incorporates mesoscale models for the effects of fuel microstructure evolution on the transport of wastage-inducing lanthanides through the fuel and for the kinetics of cladding wastage layer growth. The mesoscale models, in turn, build on lanthanide transport property data obtained from the atomistic scale. Preliminary validation studies using wastage thickness and cladding profilometry data from four fuel rods irradiated in Experimental Breeder Reactor II experiment X447 and one fuel rod from Fast Flux Test Facility experiment IFR1 show that the new model predicts cladding wastage and its effects on cladding deformation as well as existing empirical FCCI correlations. The new model is expected to aid in the design of new metallic fuel concepts, including fuel additives, cladding liners, and sodium-free annular fuel geometries. In conclusion, future work will focus on broader validation and refinement of the model’s treatment of different fuel alloys and cladding materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fuel cycle impacts of accident tolerant fuels

This paper discusses the impact of several accident-tolerant fuel (ATF) options on the fuel cycle performance of once-through LWRs compared to current uranium oxide-Zircaloy fueled commercial LWRs. Desirable characteristics of proposed ATF concepts include higher thermal conductivity, enhanced fission product retention, and reduced reaction kinetics and hydrogen generation in loss of coolant conditions. ATF concepts that have been proposed target changes to the fuel and/or cladding. Broadly speaking, the options can be summarized as follows: -) UO{sub 2} pellet fuels with increased thermal conductivity to lower temperatures, -) high-density pellet fuels (e.g., nitrides and/or silicides), -) fully ceramic microencapsulated (FCM) fuels, -) metallic fuel with high fissile density and high thermal conductivity, -) coatings on Zircaloy cladding, -) total elimination of the Zircaloy as cladding and replacement with a ceramic, and -) composite cladding including for instance layered Zircaloy and SiC and molybdenum. We can draw 4 conclusions. First, other than SiC, the other cladding options will result in a reactivity penalty requiring either thinner cladding or increased fissile loading. Secondly, silicide fuels have a higher density than UO{sub 2} and can allow longer cycle lengths for the same enrichment or a reduction in the enrichment for a given cycle length. Thirdly, while the pellet fuels yield very similar results, the FCM fuel is dramatically different due to the low fuel loading. Fourthly, we have to note that in order to achieve the maximum benefit from a particular fuel-clad combination, it requires consideration of the characteristics of each. For example, a cladding with a relatively high absorption (e.g., FeCrAl) would benefit from a high density fuel such as U{sub 3}Si{sub 2}.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

EXPERIMENTAL STUDY OF THE EFFECT OF START OF INJECTION AND BLEND RATIO ON SINGLE FUEL REFORMATE RCCI

A new concept of single-fuel RCCI has been proposed through the catalytic partial oxidation reformation of diesel fuel. The reformed fuel mixture is then used as the low reactivity fuel and diesel itself is used as the high reactivity fuel. In this paper, two selected reformates mixture from the reformation of diesel were selected for further analysis. Each reformate fuel mixture contained a significant fraction of inert gases (89% and 81%). The effects of the difference in the molar concentrations of the reformate mixtures were studied by experimenting with diesel as the direct injected fuel in RCCI over a varying start of injection timings and different blend ratios (i.e., the fraction of low and high reactivities fuels). The reformate mixture with the lower inert gas concentration had earlier combustion phasing and shorter combustion duration at any given diesel start of injection timing. The higher reactivity separation between reformate mixture and diesel, compared with gasoline and diesel, causes the combustion phasing of reformate-diesel RCCI to be more sensitive to the start of injection timing. The maximum combustion efficiency was found at a CA50 before TDC, whereas the maximum thermal efficiency occurs at a CA50 after TDC. The range of energy-based blend ratios in which reformate-diesel RCCI is possible is between 25% and 45%, limited by ringing intensity (RI) at the low limit of blend ratios, and COV of IMEP and combustion efficiency at the high limit. Intake boosting becomes necessary due to the oxygen deficiency caused by the low energy density of the reformate mixtures as it displaces intake air.

Hariharan, Deivanayagam↗
DOE OSTI2021

Mapping the combustion modes of a dual-fuel compression ignition engine

Compression-ignition (CI) engines can produce higher thermal efficiency (TE) and thus lower carbon dioxide (CO 2 ) emissions than spark-ignition (SI) engines. Unfortunately, the overall fuel economy of CI engine vehicles is limited by their emissions of nitrogen oxides (NO x ) and soot, which must be mitigated with costly, resource- and energy-intensive aftertreatment. NO x and soot could also be mitigated by adding premixed gasoline to complement the conventional, non-premixed direct injection (DI) of diesel fuel in CI engines. Several such “dual-fuel” combustion modes have been introduced in recent years, but these modes are usually studied individually at discrete conditions. This paper introduces a mapping system for dual-fuel CI modes that links together several previously studied modes across a continuous two-dimensional diagram. This system includes the conventional diesel combustion (CDC) and conventional dual-fuel (CDF) modes; the well-explored advanced combustion modes of HCCI, RCCI, PCCI, and PPCI; and a previously discovered but relatively unexplored combustion mode that is herein titled “Piston-split Dual-Fuel Combustion” or PDFC. Tests show that dual-fuel CI engines can simultaneously increase TE and lower NO x and/or soot emissions at high loads through the use of Partial HCCI (PHCCI). At low loads, PHCCI is not possible, but either PDFC or RCCI can be used to further improve NO x and/or soot emissions, albeit at slightly lower TE. These results lead to a “partial dual-fuel” multi-mode strategy of PHCCI at high loads and CDC at low loads, linked together by PDFC. Drive cycle simulations show that this strategy, when tuned to balance NO x and soot reductions, can reduce engine-out CO 2 emissions by about 1% while reducing NO x and soot by about 20% each with respect to CDC. This increases emissions of unburnt hydrocarbons (UHC), still in a treatable range (2.0 g/kWh) but five times as high as CDC, requiring changes in aftertreatment strategy.

Engineering↗
DOE OSTI2025

Modeling a Thermal-Spectrum LEU-fueled Molten Salt Reactor Co-fueled with Thorium in SCALE 6.3.1 and Serpent-2

Interest in the development of advanced nuclear fission reactors for commercial electricity production has risen in recent years, with Generation IV reactor designs offering numerous advantages in safety, efficiency, fuel cycle sustainability, and waste management. In particular, thorium-fueled molten salt reactors (MSRs) are considerably promising for enhancing fuel cycle sustainability in their ability to breed fissile 233U fuel from thorium, a presently untapped and widely abundant resource. A novel MSR fuel cycle concept, the ”Sourdough” refueling and waste management strategy, has previously been demonstrated with a traditional uranium-based fuel cycle in a thermal-spectrum MSR operating with low enriched uranium (LEU) fuel with favorable neutronic performance. However, the ability to use this unique fuel cycle approach with thorium-based molten salt fuels has not yet been studied. In this work, the Sourdough fuel cycle was implemented in a small, thermal-spectrum MSR fueled with high assay low enriched uranium (HALEU) and fertile 232Th for breeding 233U fuel. Relevant neutronic data, including fuel and isothermal temperature feedback behavior, was studied using the SCALE 6.3.1 and Serpent-2 code systems, and keff data was measured during simulated depletion at 400 MWth. The Sourdough fuel cycle concept is shown to perform favorably with a thorium-fueled MSR model, thus warranting further study into its use in other MSR designs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

An integrated approach to examine fuel-cladding chemical interaction in HT9/U-10Zr metallic fast reactor fuels: Coupling machine learning with electron microscopy and local mechanical properties analysis

The metallic U-Zr nuclear fuel alloy has garnered renewed interest as a promising candidate for next-generation sodium-cooled fast reactors. Recent studies and technology assessments have identified several areas requiring improvements, enhanced knowledge, and reliable data to strengthen the U-Zr fuel design basis for qualification and commercial applications. One of the most challenging phenomena impacting this fuel system’s performance is fuel-cladding chemical interaction (FCCI). This work aimed to harvest FCCI data by examining selected HT9/U-10Zr (wt. %) fuel samples of prototypic full-length fuel pins through an integrated approach. This approach integrated scanning electron microscopy (SEM) microstructure characterization with localized mechanical properties examination to deepen understanding of FCCI phenomenon in HT9/U-10Zr fuel system. Particularly, this study focused on MFF fuel pins irradiated at Fast Flux Test Facility (FFTF), which aimed to qualify metallic fuel as a driver fuel for FFTF and to assess its viability for larger-scale fast reactors. Electron microscopy provided high confidence in detecting and distinguishing the different FCCI layers, while small-scale mechanical testing (SSMT) probed the mechanical properties of these layers. SEM examination of a MFF-2 pin 192167, with a time averaged inner cladding temperature (TICT) slightly over 500°C, revealed minimal cladding-side FCCI (cladding wastage). In contrast, significantly thicker cladding wastage comprising two distinct sublayers was observed in samples from the thermally hot MFF-3 pin 193045 and MFF-5 pin 195011 where the TICT ranged from 610-635°C. SSMT indicated complete embrittlement in the sublayer adjacent to the fuel and a tendency toward embrittlement in the other sublayer. Additionally, a new machine learning method was developed, validated, and used to quantify cladding wastage thickness. The machine learning method reliably predicted the wastage thickness across various fuel pins and sample cross-sections. Furthermore, the available cladding wastage data from HT9/U-10Zr fuel system demonstrated a strong temperature dependency. However, the dataset remains small, and ongoing research activities are essential to further understand the FCCI phenomenon and develop a reliable FCCI model for enhanced fuel performance simulation under various conditions.

36 - MATERIALS SCIENCE↗
DOE OSTI2017

Multivariate analysis of gamma spectra to characterize used nuclear fuel

Abstract—The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of reprocessing streams in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type. Locally weighted PLS models were fitted on-the-fly to estimate continuous fuel characteristics. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE. This automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters and material diversions.

Multivariate Analysis for Characterization of Used↗
DOE OSTI2021

Initial demonstration of automated fuel performance modeling with 1977 EBR-II metallic fuel pins using BISON code with FIPD and IMIS databases

Using the BISON fuel performance code, simulations were conducted using an automated process to read initial and operating conditions from the Fuels Irradiation and Physics Database (FIPD) and Integral Fast Reactor materials information system (IMIS) database, which contains metallic fuel data from the Experimental Breeder Reactor-II (EBR-II). This work demonstrates use of an integrated framework to access the vast majority of EBR-II experimental fuel pin data to support rapid development of fuel performance models for next-generation metallic fuel systems. With this capability, validation for fuel qualification can be performed rapidly. Between IMIS and FIPD, there is enough information to conduct 1977 unique EBR-II metallic fuel pin histories from 24 different experiments, at varying levels of detail between the two databases. Each of these histories includes a high-resolution power history, flux history, coolant channel flow rates, and coolant channel temperatures. Fission gas release (FGR), cumulative damage fraction (CDF), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON. The results were compared to post-irradiation examination (PIE) results for the initial demonstration of automated BISON modeling. BISON simulations conducted with IMIS and FIPD were in rough agreement with PIE measurements and calculations. Cladding profilometry, FGR, and fuel axial swelling were found to be in rough agreement with PIE measurements, depending on the physics used within the BISON input files. Here, the mechanical contact solver chosen was found to significantly impact axial fuel swelling and cladding strain predictions. CDF values were assessed to see whether pin failure may have been predicted (CDF ≥ 1). This work suggests that continued development of an automated tool for BISON should focus on inclusion of the Fast Flux Test Facility (FFTF) experimental data for a larger database for metallic fuel, improved physical models to better capture fuel performance, such as fuel-cladding interactions, and a more detailed comparison with available PIE data to further the BISON model development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Degradation and Failure Phenomena of Advanced Reactor Fuel Concepts: Sodium-Cooled Fast Reactor Metallic Fuel

The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of new fuel types in advanced nuclear power reactors that would be designed and built in the United States. Pacific Northwest National Laboratory (PNNL) is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs that would be deployed in these reactors. This report focuses specifically on the metallic fuel that is being considered for sodium-cooled fast reactors and specifically on mechanisms that would cause damage or failure to the fuel under reactor operating conditions and design basis accident conditions. There is historic experience with both metallic and oxide fuels in experimental sodium-cooled fast reactors, but this report will focus solely on metallic fuel. Currently two U.S.-based reactor designers are engaged with the NRC in the application or pre-application review and considering a sodium-cooled fast reactor. TerraPower is engaged with NRC in pre-application review of its Natrium reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel clad in HT9 stainless steel. ARC Clean Technology is engaged with NRC in pre-application review of its ARC-100 reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel with steel cladding. This report will focus on this fuel system specifically, with broader information given regarding other metallic fuel systems with other stainless steel alloy claddings. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of metallic fuel concepts for sodium-cooled fast reactors, including fuel performance characteristics that may not be addressed within existing regulatory documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup fuel storage and transportation. This paper discusses the staff's preliminary considerations on the safety implication of fuel reconfiguration with respect to nuclear safety (subcriticality control), radiation shielding, containment, the performance of the thermal functions of the packages, and the retrievability of the contents from regulatory perspective. (authors)

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

Accelerated Fuel Qualification of Fast Modular Reactor Fuel in a Thermal Reactor: Modeling and Simulation Paired with Irradiation Testing

The accelerated fuel qualification (AFQ) methodology is applied by simulating accelerated fuel tests of the General Atomics Electromagnetic Systems’ fuel system for its 44-MW(electric) gas-cooled, fast-spectrum fast modular reactor (FMR). This fuel is comprised of UO 2 pellets in SiGA® cladding, a silicon carbide ceramic matrix composite. Fast reactors, like the FMR, offer many benefits, including high fuel utilization and flexibility, but may require a lengthy material design process if tests are performed using fast neutron irradiation alone. A thermal neutron irradiation can instead be used to rapidly test how well key components of the current material models extend to high burnup. Thermal neutrons produce a different radial power distribution within the pin than fast neutrons. However, the temperature and burnup values for the two neutron types are comparable, and the differences between the simulated fuel responses are relatively small, demonstrating the weak sensitivity of the physics-based fuel model calculations on the neutron type and the irradiation rate. Furthermore, the deformation of the SiGA cladding saturates after about 1 displacement per atom for both neutron spectra. In an accelerated fuel test, the irradiation time required to reach the target fuel burnup can be reduced by a factor of 3 by using a small rodlet with a 45% smaller pellet diameter while maintaining the same linear power. Therefore, the time for data collection up to high burnup can be significantly reduced while maintaining the same temperature profile, which largely determines the material response. Tests of fuel rodlets of standard and compact size will be carried out in the Idaho National Laboratory’s Advanced Test Reactor (ATR), including full size and compact rodlets with varying gap sizes. By applying physics-based mechanistic modeling and simulation in accordance with the AFQ methodology, this type of compact rodlet testing in a thermal test reactor captures the necessary phenomena to test fuel material models up to high burnup and to simulate the expected impact of fast neutron radiation on the fuel in FMR operations. Finally, this approach to testing fast reactor fuels in existing thermal test reactors, paired with advanced physics-based mechanistic modeling and simulation, is expected to be applicable to a range of advanced fuels and will decrease the overall fuel qualification timeframe from decades to years.

Advanced test reactor (ATR)↗
NASA NTRS1984

Fuel system design concepts for broad property fuels

The results of a study assessing the impact of using jet fuel with relaxed specification properties on an aircraft fuel system are given. The study objectives were to identify credible values for specific fuel properties which might be relaxed, to evolve advanced fuel system designs for airframe and engines which would permit use of the specified relaxed properties fuels, and to evaluate performance of the candidate advanced fuel systems and the relaxed property fuels in a typical transport aircraft. The study used, as a baseline, the fuel system incorporated in the Lockheed Tristar. This aircraft is powered by three RB.211-524 Rolls-Royce engines and incorporates a Pratt and Whitney ST6C-421 auxiliary power unit for engine starting and inflight emergency electrical power. The fuel property limits examined are compared with commercial Jet A kerosene and the NASA RFP fuel properties. A screening of these properties established that a higher freezing point and a lower thermal stability would impact fuel system design more significantly than any of the other property changes. Three candidate fuel systems which combine the ability to operate with fuels having both a high freeze point and a low thermal stability are described. All candidates employ bleed air to melt fuel freeze-out prior to starting the APU or an inoperable engine. The effects of incorporating these systems on aircraft weight and engine specific fuel consumption are given.

Versaw, E. F.↗
DOE OSTI2015

THE MISSION AND ACCOMPLISHMENTS FROM DOE’S FUEL CYCLE RESEARCH AND DEVELOPMENT (FCRD) ADVANCED FUELS CAMPAIGN

The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors, enhance proliferation resistance of nuclear fuel, effectively utilize nuclear energy resources, and address the longer-term waste management challenges. This includes development of a state of the art Research and Development (R&D) infrastructure to support the use of a “goal oriented science based approach.” AFC uses a “goal oriented, science based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R&D activities across multiple organizations. AFC interfaces and collaborates with Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Challenged with the research and development of fuels for two different reactor technology platforms, AFC targeted transmutation fuel development and focused ceramic fuel development for Advanced LWR Fuels.

AFC Mission↗
DOE OSTI2021

Performance of a Natural Gas Solid Oxide Fuel Cell System With and Without Carbon Capture

The fuel cell program at the United States Department of Energy (DOE) National Energy Technology Laboratory (NETL) is focused on the development of low-cost, highly efficient, and reliable fossil-fuel-based solid oxide fuel cell (SOFC) power systems that can generate environmentally-friendly electric power with at least 90 percent carbon capture. NETL’s SOFC technology development roadmap is aligned with near-term market opportunities in the distributed generation sector to validate and advance the technology while paving the way for utility-scale natural gas (NG)- and coal-derived synthesis gas-fueled applications via progressively larger system demonstrations. The present study represents a part of a series of system evaluations being carried out at NETL to aid in prioritizing technological advances along research pathways to the realization of utility-scale SOFC systems, a transformational goal of the fuel cell program. In particular, the system performance of utility-scale NG fuel cell (NGFC) systems with and without carbon dioxide (CO2) capture is presented. The NGFC system analyzed features an external auto-thermal reformer (ATR) feeding the fuel to the SOFC system consisting of planar anode-supported SOFC with separated anode and cathode off-gas streams. In systems with CO2 capture, an air separation unit (ASU) is used to provide the oxygen for the ATR and for the combustion of unutilized fuel in the SOFC anode exhaust along with a CO2 purification unit to provide a nearly pure CO2 stream suitable for transport for usage in enhanced oil recovery operations or for storage in underground saline formations. Remaining thermal energy in the exhaust gases is recovered in a bottoming steam Rankine cycle while supplying any process heat requirements. A reduced order model (ROM) developed at the Pacific Northwest National Laboratory (PNNL) is used to predict the SOFC performance. The ROM, while being computationally effective for system studies, provides other detailed information about the state of the stack, such as the internal temperature gradient, generally not available from simple performance models often used to represent the SOFC. Such additional information can be important in system optimization studies to preclude operation under off-design conditions that can adversely impact overall system reliability. The NGFC system performance was analyzed by varying salient system parameters, including the percent of internal (to the SOFC module) NG reformation—ranging from 0 to 100 percent—fuel utilization, and current density. The impact of advances in underlying SOFC technology on electrical performance was also explored.

solid oxide fuel cell (SOFC), natural gas fuel cel↗