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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

A low-cost solid–liquid separation process for enzymatically hydrolyzed corn stover slurries

Solid-liquid separation of intermediate process slurries is required in some process configurations for the conversion of lignocellulosic biomass to transportation fuels. Thermochemically pretreated and enzymatically hydrolyzed corn stover slurries have proven difficult to filter due to formation of very low permeability cakes that are rich in lignin. Treatment of two different slurries with polyelectrolyte flocculant was demonstrated to increase mean particle size and filterability. Filtration flux was greatly improved, and thus scaled filter unit capacity was increased approximately 40-fold compared with unflocculated slurry. Although additional costs were accrued using polyelectrolyte, techno-economic analysis revealed that the increase in filter capacity significantly reduced overall production costs. Fuel production cost at 95% sugar recovery was reduced by $1.35 US per gallon gasoline equivalent for dilute-acid pretreated and enzymatically hydrolyzed slurries and $3.40 for slurries produced using an additional alkaline de-acetylation preprocessing step that is even more difficult to natively filter.

09 BIOMASS FUELS↗
DOE OSTI2025

Operational Experience and Development of a Reliability Model for the ATR Demineralizer System (Slides)

The Advanced Test Reactor (ATR) is a light water reactor with aluminum clad driver fuel. Strict limits are implemented on pH, conductivity, and filterable solids to assure the performance of the driver fuel limit corrosion of the primary coolant system (PCS) pressure boundary components. A bypass demineralizer system is used to maintain PCS coolant within allowable bands for each of the aforementioned variables. Unlike many other reactors, the ATR does not have a filtration system. As such, failure of ion exchange resin in the bypass demineralizer can defeat the purpose of the system and result in high filterable solids in the PCS. Power operations with high filterable solids is prohibited by a technical safety requirement. This paper discusses recent failures of the ATR bypass demineralizer system. The testing and analysis done to deduce the cause of the failures is covered, as well as short term mitigative steps that were taken. To prevent the likelihood of future issues, historical data from the ATR and industry is used to develop a reliability model of the ATR bypass demineralizer system. Evidence of radiolytic decomposition of ion exchange resin was found in the cause investigation. As such, special attention is given to ion exchange resin failure due to high radiation environments. The reliability model is used to develop the ATR testing and maintenance program. Specifically, limits on resin service life are discussed.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2017

Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report

The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach has not been to create testing conditions that are fully prototypic of a fast reactor, but to create a thermal profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of limited fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The available evidence supports the assertion that use of cadmium-filtering creates irradiation conditions in the ATR that are both relevant to, and suitable for, fast reactor fuel performance testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report

The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach has not been to create testing conditions that are fully prototypic of a fast reactor, but to create a thermal profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of limited fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The available evidence supports the assertion that use of cadmium-filtering creates irradiation conditions in the ATR that are both relevant to, and suitable for, fast reactor fuel performance testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Meeting the industry challenge to achieve leak free performance

Deployment of highly reliability fuel products is one of AREVA NP's overarching philosophies. Any fuel defect brings appreciable burden to utility and fuel vendors alike. AREVA NP's achievements in leaker free performance are direct results of rigorous investigations and development of innovative products and solutions to address longstanding industry fuel reliability concerns. An illustrative example of this commitment can be found in the AREVA NP's BWR fuel product line. PCI had been a leading failure mechanism in BWR fuel supplied in the United States. In response, AREVA NP implemented improved pellet designs, tighter manufacturing standards, and industry-leading power maneuvering methodologies that have eliminated PCI failures altogether. Likewise, failures resulting from small debris forms in the primary system have been a leading failure mechanism in the BWR fuel. The implementation of advanced debris filters such as the Improved FUELGUARD and 3. Generation FUELGUARD have virtually eliminated debris-related failures. AREVA NP's commitment to fuel reliability is not limited to the BWR fuel products. Grid-to-Rod Fretting (GTRF) has long been a dominant PWR failure mechanism in the industry and results from challenging cross-flows within specific reactor designs that excite and ultimately cause the fuel rod to fret against the spacer contacts to the point of failure. In 2010 alone, 32 PWR fuel rods were discharged in the United States region that had failed due to fretting-related conditions. To eliminate this longstanding industry challenge, a suite of fuel products developed around the HTP (High Thermal Performance) spacer platform were implemented. The HTP grid design, with its multiple line contacts, serves to cradle the fuel rod and dampen vibrations in the presence of challenging flow conditions. When coupled with the Alloy 718 HMP (High Mechanical Performance) lower grid, HTP fuel has never experienced a GTRF failure. The AREVA NP's commitment to fuel reliability is successful: in 2010 a total of 41 failed fuel rods were discharged from reactors in the United States whereas in 2015 no failed fuel rods were discharged out of the approximately 1.3 million fuel rods in operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1983

Atomization and combustion performance of antimisting kerosene and jet fuel

Combustion performance of antimisting kerosene (AMK) containing FM-9 polymer was investigated at various levels of degradation (restoration of AMK for normal use in a gas turbine engine). To establish the relationship of degradation and atomization to performance in an aircraft gas turbine combustor, sprays formed by the nozzle of a JT8-D combustor with Jet A and AMK at 1 atmosphere (atm) (14.1 lb/square in absolute) pressure and 22 C at several degradation levels were analyzed. A new spray characterization technique based on digital image analysis of high resolution, wide field spray images formed under pulsed ruby laser sheet illumination was developed. Combustion tests were performed for these fuels in a JT8-D single can combustor facility to measure combustion efficiency and the lean extinction limit. Correlation of combustion performance under simulated engine operating conditions with nozzle spray Sauter mean diameter (SMD) measured at 1 atm and 22 C were observed. Fuel spray SMD and hence the combustion efficiency are strongly influenced by fuel degradation level. Use of even the most highly degraded AMK tested (filter ratio = 1.2) resulted in an increase in fuel consumption of 0.08% to 0.20% at engine cruise conditions.

Fleeter, R.↗
DOE OSTI2016

Computed Tomography of Dry Cask Storage for Used Fuel

The demand for dry cask storage for used nuclear fuel is on the rise due to the continuous operation of currently existing nuclear plants which are reaching or have reached the capacity of their used fuel pools. Dry cask storage allows used fuel that has already been cooled in the used fuel pool for at least five years to be surrounded by an inert gas inside a specially designed and licensed cask. Currently, 75 of the 78 U.S.A reactor sites either have or are pursuing an independent spent fuel storage installation (ISFSI) for dry cask fuel storage [1] with over 2,000 casks already loaded [2]. Cask loading will only continue to increase until a final repository solution is operational or the U.S. begins to recycle fuel. The challenge of monitoring the condition of the used fuel inside the cask without exposing the used fuel has been apparent since the 1980s. The objective of the University of Florida (UF) project is ultimately to investigate the feasibility of imaging the used fuel assemblies’ location and evaluating the inner structural condition using emission source computed tomography (CT). A combination of radiation transport models will be used to simulate gamma-rays and neutrons traversing through the used fuel, structure, and cask body materials. In the simulations, multiple linear detector arrays will record the radiation outside the cask wall and will be placed at various angles in order to model rotating a linear detector array around the entire cask. The combination of data from all the detectors will give the planar radiation projection onto the detectors, more commonly called the Radon transform, of the structure of used fuel assemblies. This project can be processed with filtered back projection or other more advanced reconstructive imaging algorithms to recreate the fuel structures. A majority of the historical work done for used fuel has been on verification, diversion detection, and burnup credit validation in order to meet regulatory guidelines. One of the more commonly used systems for single assembly verification is the Fork Detector developed at Los Alamos National Laboratory in 1988 [3]. A group at the Idaho National Laboratory developed a Compton Dry-Cask Imaging Scanner that has the ability to detect empty vs full storage positions but it suffered a setback when transported to the Doel Nuclear Power Plant near Antwerp, Belgium; the system was unable to detect full energy peaks because of the additional scattering caused by a thick ballistic steel shield [4]. As detectors, computational power, and modeling improves, a few groups have leveraged these advancements and started to investigate CT hoping, that despite the difficulties, the technique can give information on fuel condition. So far the groups have mostly limited the tomographic work to a per assembly basis [5] [6] [7], rather than the full dry cask as the University of Florida plans to investigate. The downsides of CT versus simpler verification and validation are the amount of data required and the time it takes to acquire that data. Given that dry casks are intended to sit undistributed for up to 60 years, a scanning time of several weeks to several months is not unreasonable if such systems do not require human operators beyond the initial setup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Computed Tomography of Dry Cask Storage for Used Fuel

The demand for dry cask storage for used nuclear fuel is on the rise due to the continuous operation of currently existing nuclear plants which are reaching or have reached the capacity of their used fuel pools. Dry cask storage allows used fuel that has already been cooled in the used fuel pool for at least five years to be surrounded by an inert gas inside a specially designed and licensed cask. Currently, 75 of the 78 U.S.A reactor sites either have or are pursuing an independent spent fuel storage installation (ISFSI) for dry cask fuel storage with over 2,000 casks already loaded. Cask loading will only continue to increase until a final repository solution is operational or the U.S. begins to recycle fuel. The challenge of monitoring the condition of the used fuel inside the cask without exposing the used fuel has been apparent since the 1980's. The objective of the University of Florida (UF) project is ultimately to investigate the feasibility of imaging the used fuel assemblies' location and evaluating the inner structural condition using emission source computed tomography (CT). A combination of radiation transport models will be used to simulate gamma-rays and neutrons traversing through the used fuel, structure, and cask body materials. In the simulations, multiple linear detector arrays will record the radiation outside the cask wall and will be placed at various angles in order to model rotating a linear detector array around the entire cask. The combination of data from all the detectors will give the planar radiation projection onto the detectors, more commonly called the Radon transform, of the structure of used fuel assemblies. This project can be processed with filtered back projection or other more advanced reconstructive imaging algorithms to recreate the fuel structures. A majority of the historical work done for used fuel has been on verification, diversion detection, and burnup credit validation in order to meet regulatory guidelines. One of the more commonly used systems for single assembly verification is the Fork Detector developed at Los Alamos National Laboratory in 1988. A group at the Idaho National Laboratory developed a Compton Dry-Cask Imaging Scanner that has the ability to detect empty vs full storage positions but it suffered a setback when transported to the Doel Nuclear Power Plant near Antwerp, Belgium; the system was unable to detect full energy peaks because of the additional scattering caused by a thick ballistic steel shield. As detectors, computational power, and modeling improves, a few groups have leveraged these advancements and started to investigate CT hoping, that despite the difficulties, the technique can give information on fuel condition. So far the groups have mostly limited the tomographic work to a per assembly basis , rather than the full dry cask as the University of Florida plans to investigate. The downsides of CT versus simpler verification and validation are the amount of data required and the time it takes to acquire that data. Given that dry casks are intended to sit undistributed for up to 60 years, a scanning time of several weeks to several months is not unreasonable if such systems do not require human operators beyond the initial setup. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Fracture strength and elastic modulus as a function of porosity for hydroxyapatite and other brittle materials Part I

Porous brittle materials are used as filters, catalyst supports, solid oxide fuel cells and biomedical materials. However the literature on the Weibull modulus, m, versus volume fraction porosity, P, is extremely limited despite the importance of m as a gauge of mechanical reliability. In Part I of this study, m is determined for 441 sintered hydroxyapatite (HA) specimens fractured in biaxial flexure for 0.08 P 0.62. In this study, we analyze a combined data set collected from the literature that represents work from a total of 17 different research groups (including the present authors), eight different materials and more than 1560 oxide and non-oxide specimens, the m versus P plot is Ushaped with a wide band of m values for P < 0.1 (Region I) and P > 0.55 (Region III), and a narrower band of m values in the intermediate porosity region of 0.1 < P < 0.55 (Region II). The limited range of m ( 4 < m < 11) in Region II has important implications since Region II includes the P range for the majority of the applications of porous brittle materials. Part II of this study focuses on the P dependence of the mean fracture strength,< f >, and the Young s modulus E for the HA specimens tested in Part I along with literature data for other brittle materials. Both < f > and E are power law functions of the degree of densification, , where = 1 P/PG and PG is the green (unfired) porosity.

30 DIRECT ENERGY CONVERSION↗
NASA NTRS1966

Valve effectively controls amount of contaminant in flow stream

Contaminant valve with a coaxial groove rotor uniformly deposits contaminant into a flow stream under full pressure and flow conditions. The valve tests filters and filter elements of hydraulic oil, fuel, or lubricant systems without any detrimental effect on the performance.

Schnitzer, T. E.↗
DOE OSTI2020

Performance Evaluation of AG-1 FC HEPA Filters and Medium in Nuclear Complex Facilities - 20434

High Efficiency Particulate Air (HEPA) filters are credited as the final barrier against the release of radioactive aerosol contamination in nearly every operating U. S. Department of Energy (DOE) and National Nuclear Security Agency (NNSA) nuclear facility. The Institute for Clean Energy Technology (ICET) at Mississippi State University maintains a research program that studies various aspects of these components of containment systems and seeks to answer key questions from across the industry. This programmatic overview will include results from recent and ongoing studies, including a study on filter design and performance envelope, degradation due to aging or fatigue, fire event impact on HEPA filters, and accelerated aging of medium and component parts. ICET has completed a study helping to define the threshold for combined elevated temperature and relative humidity resistance of separator style and separator-less style HEPA filters that are used in DOE complexes. The study provides experimental data that considers elevated temperature, elevated relative humidity (RH), and target differential pressure (dP) in an attempt to gain a better understanding or more comprehensive insight into the operating envelope of different configurations of ASME (American Society for Mechanical Engineers) AG-1 Section FC filters. Test variables including air temperature ranges of 48.88 deg. C, 54.44 deg. C, or 60 deg. C (120, 130, or 140 deg. F); air RH ranges including 60-70%, 80%, or 90+% with initial filter dP of either 497.68 Pa, 746.52 Pa, or 995.36 Pa (2 inches water column (in. w. c.), 3 in. w. c., or 4 in. w. c.) included for full bracketing of each set of conditions. A study examining the effects of aging and fatigue on nuclear grade HEPA filters and medium elucidates the physical properties of media along with testing of new and aged ASME AG-1 HEPA filters. In addition to performing autopsies of the tested, aged filters to help better understand the service life of their performance, newly designed accelerated aging chambers allow new medium to undergo accelerated aging and exposure treatments. Evaluation of degradation in physical properties and functionality requires a high population of aged filters of different ages, designs, manufacturers, and operational histories. The limited availability of this population and the impropriety of looking solely at formulations and components from past years requires experimental design that will provide foresight into future filter performance. This undertaking therefore involves not only properties analysis of aged medium from different operational histories, ages, designs, and manufacturers, but also prescience in analytical goals via accelerated aging studies of newly manufactured media. A more complete understanding of the mechanisms of degradation from past formulations as well as current and future formulations is possible. Accelerated Aging of newly manufactured media utilizes exposure treatments based on the Arrhenius equation to artificially age medium and therefore glimpse into the future. Evaluation of current media properties This testing allows for the comparison of performance and durability of new filters under upset or design basis conditions with aged filters that were in service under ambient conditions and other aged filters retained in storage. Susceptibility of HEPA filters to the effects of combustion byproducts, heat, and water is also being studied in an attempt to gain understanding on the prevention of filter failures during fire events in nuclear facilities. ICET seeks to determine the effect of filter performance due to smoke loading using characterized smoke from variable fuel compositions, geometries, and loads; the smoke capacity of various filters with different burning conditions, heat release rates and transport, as well as mass transport from the air stream will be studied in low flow containment systems. (authors)

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

Low-Temperature Combustion of High Octane Fuels in a Gasoline Compression Ignition Engine

Gasoline Compression Ignition (GCI) has been shown as one of the advanced combustion concepts that could potentially provide a pathway to achieve cleaner and more efficient combustion engines. Fuel and air in GCI are not fully premixed as compared to homogeneous charge compression ignition (HCCI) which is a completely kinetic-controlled combustion system. Therefore, the combustion phasing can be controlled by the time of injection, usually post injection in a multiple-injection scheme, to mitigate combustion noise. Gasoline fuels ignite more difficult than Diesel. The autoignition quality of gasoline can be indicated by research octane number (RON). Fuels with high octane tend to have more resistance to auto-ignition, hence more time for fuel-air mixing. In this study, three fuels, namely, Aromatic, Alkylate, and E30, with similar RON value of 98 but different hydrocarbon compositions were tested in a multi-cylinder engine under GCI combustion mode. Considerations of EGR, start of injection (SOI), and boost were investigated to study the sensitivity of dilution, local stratification, and reactivity of the charge, respectively, for each fuel. Combustion phasing was kept constant during the experiments to the changes in ignition and combustion process before and after 50% of the fuel mass is burned. Emission characteristics at different levels of EGR and lambda were revealed for all fuels with E30 having the lowest filter smoke number (FSN) and was also most sensitive to the change in dilution. Reasonably low combustion noise (< 90 dB) and stable combustion (COVIMEP < 3%) were maintained during the experiments. The second part of this paper contains visualization of the combustion process obtained from endoscope imaging for each fuel at selected conditions. Soot radiation signal from GCI combustion were strong during late injection, and also more intense at low EGR conditions. Furthermore, soot/temperature profiles indicated only the high-temperature combustion period, while cylinder pressure-based heat release rate (HRR) showed a two-stage combustion phenomenon.

33 ADVANCED PROPULSION SYSTEMS↗
NASA NTRS2005

NASA Tech Briefs, August 2005

Topics include: Hidden Identification on Parts: Magnetic Machine-Readable Matrix Symbols; System for Processing Coded OFDM Under Doppler and Fading; Multipurpose Hyperspectral Imaging System; Magnetic-Flux-Compensated Voltage Divider; High-Performance Satellite/Terrestrial-Network Gateway; Internet-Based System for Voice Communication With the ISS; Stripline/Microstrip Transition in Multilayer Circuit Board; Dual-Band Feed for a Microwave Reflector Antenna; Quadratic Programming for Allocating Control Effort; Range Process Simulation Tool; Simulator of Space Communication Networks; Computing Q-D Relationships for Storage of Rocket Fuels; Contour Error Map Algorithm; Portfolio Analysis Tool; Glass Frit Filters for Collecting Metal Oxide Nanoparticles; Anhydrous Proton-Conducting Membranes for Fuel Cells; Portable Electron-Beam Free-Form Fabrication System; Miniature Laboratory for Detecting Sparse Biomolecules; Multicompartment Liquid-Cooling/Warming Protective Garments; Laser Metrology for an Optical-Path-Length Modulator; PCM Passive Cooling System Containing Active Subsystems; Automated Electrostatics Environmental Chamber; Estimating Aeroheating of a 3D Body Using a 2D Flow Solver; Artificial Immune System for Recognizing Patterns; Computing the Thermodynamic State of a Cryogenic Fluid; Safety and Mission Assurance Performance Metric; Magnetic Control of Concentration Gradient in Microgravity; Avionics for a Small Robotic Inspection Spacecraft; and Simulation of Dynamics of a Flexible Miniature Airplane.

Source record↗
DOE OSTI2016

Proven Uranium Recovery Solutions for the Nuclear Fuel Manufacturing Market - 16060

During the Nuclear Fuel Manufacturing process a number of by-products and residues are generated from various sources. In recent years, Westinghouse has characterised and processed around 50,000 drums of residue comprising of numerous different types generated over 60 years from the manufacture of numerous types of fuel for greater than 50 reactors. These include, AGR, Magnox (uranium metal fuel), PWR, fast reactors and various test reactors. Different organisations have different definitions and names for each of these categories but essentially they fall into one of the following categories: Clean Scrap (U{sub 3}O{sub 8}, UO{sub 2}, UO{sub 3}, UF{sub 4}, powders, pellets), Dirty Scrap (U{sub 3}O{sub 8}, UO{sub 2}, UO{sub 3}, UF{sub 4}, powders, pellets), Gd Contaminated Scrap (U{sub 3}O{sub 8}, UO{sub 2}, powders, pellets), Ash, Sludge, Combustible / soft residues (e.g. ventilation filters, gloves, plastic bags), Contaminated metals (including contaminated scrap zirconium tubes), Materials arising from decommissioning. Routes have been developed across the industry for a number of the categories listed above, however in numerous examples, routes have not been established and materials have been stored creating a significant legacy. Westinghouse understands the issues regarding legacy uranic material and the financial and regulatory impact of storage. Interim storage of materials may seem like a low cost option in the short term; however ongoing costs for facilities, manpower and equipment, without a permanent and final solution can result in significant lifetime material management costs. Recovery of uranium and sending materials for safe disposal of enriched uranium residues is not easy because of the criticality risk both during processing and at a disposal site. Therefore quantities of materials than can be sent via permitted disposal routes are very limited The development of regulations often means that finding solutions is becoming more difficult. It is believed that timely resolution is ultimately the most cost effective solution. In addition, the intrinsic value of recycling this material is also high and in some cases can mitigate the cost of processing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Scoping Study for Fast Flux Testing in the Advanced Test Reactor

The value of fast spectrum reactors remains prominent in the nuclear technology portfolio. The performance of these reactors can be maximized with advancements in nuclear fuel technologies, but development of these technologies is currently held back by lack of fast spectrum test reactors available to the United States. Spectral modification of experiment positions in the thermal spectrum Advanced Test Reactor (ATR) has long been used to support fast reactor fuel development, but these methods have not been progressed to their full potential. This study investigated the use of concentric rings of aluminum-clad fuel plates in ATR flux traps and thermal neutron absorbing filters to increase fast neutron flux on test specimens. This concept was termed the Boosted Energy Advanced Spectrum Test (BEAST). This approach will enable irradiation of advanced fuel designs in prototypic-length fuel pins and representative flux environment to support post irradiation exams, enable transient testing, and produce the type of data that will permit lead test assembly irradiations in true Sodium Fast Reactors (SFRs) when they become available. Neutronic predictions were performed to investigate BEAST design options and thermal hydraulic models were produced to ensure feasibility of BEAST. Two versions were considered based on the geometric limitations of ATR’s small and large flux traps. The small version was found to be preferable due to slightly higher fast flux and fast-to-thermal neutron ratio. Perhaps more influentially, the small flux trap option was also preferred to avoid conflict with ongoing very high temperature reactor fuel irradiation programs in ATR’s large northeast flux trap. The small flux trap option provided less than half the test volume of the large version, but still had adequate volume for seven SFR pins in cross section which could be stacked two-high in ATR’s 1.2m long core to accommodate up to 14 EBR-II size pins. The preference for the small flux trap configuration should be revisited if additional collaborative test programs emerge with the need to irradiate a significant volume of additional specimens. Calculations were performed regarding a lithium deuteride ring to convert thermal neutrons into 14 MeV fusion neutrons. At the time this report was written these calculations were partially complete and it remains to be seen whether the concept would be worth including in BEAST. Given the preference for the small flux trap option, which does not afford enough room for the 14 MeV ring, it was concluded to defer future work on the lithium deuteride ring. This decision could be revisited if fusion material research programs emerge for collaborative testing in BEAST. A cadmium-lined specimen holder design was found to be adequate in filtering thermal neutrons and preferred over other neutron absorbers based on past experience with cadmium baskets. It was acknowledged that cadmium-bearing hardware would become depleted and need to be replaced occasionally, which appeared feasible from a mechanical design perspective. Neutronic studies investigated different enrichment levels in the booster fuel using uranium-molybdenum alloy dispersion fuel which has performed well in past ATR irradiations. Both options were able to drive fuel pins to SFR-like fission heating rates. The high enriched booster fuel option outperformed the low enriched option by ~20% on key metrics including fast flux and fast-to-thermal ratio, but the low enriched option was favored in order to broaden options for potential fuel suppliers. The preferred BEAST design options including cadmium filter with low enriched booster fuel in the small flux trap configuration was predicted to achieve 6.2E14 n/cm2sec fast flux (>0.1 MeV) with a fast-to-thermal ratio of 44.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗