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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Effects of Critical Compression Ratio on Rating Gasoline Knock Propensity

It is common practice in the automotive industry to explore the knock limits of fuels on an engine by a comparison of the knock limited spark advance (KLSA) at threshold knock intensity. However, the knock propensity of gasolines can be rated by changing one of three metrics on a variable compression ratio Cooperative Fuels Research (CFR) octane rating engine while holding the other two variables constant: knock intensity, spark timing, and critical compression ratio. The operational differences between the standard research octane number (RON) rating and modern engine operation have been explored in three parts. The first part focused on the effects of lambda and knock characterization. The second part studied the effects of spark timing. This third part explores the knock ratings of several gasolines by comparing the critical compression ratios at constant combustion phasing and knock intensity. The threshold knock intensity was based on the standard octane rating D1 pickup or by maximum amplitude of pressure oscillations (MAPO) measured by a piezoelectric cylinder pressure transducer. Several Fuels for Advanced Combustion Engines (FACE) gasolines, primary reference fuels (PRFs), and toluene standardization fuels (TSFs) were tested on a CFR octane rating engine with advanced data acquisition equipment and a piezoelectric cylinder pressure transducer. These tests deviated from the ASTM D2699 standard octane rating procedure. For each test fuel, the CFR engine was operated at stoichiometry at a constant combustion phasing (CA50) and the compression ratio was modified until a threshold knock intensity was realized. It was found that the chemical composition of the fuels affected the relationship of critical compression ratios between the D1 knockmeter and piezoelectric pressure transducer knock intensity thresholds, as well as the measured combustion maximum pressure rise rate and spark timing setting for constant CA50. For highly aromatic fuels tested at a constant MAPO knock intensity threshold, it was found that the maximum pressure rise rate was two to three times higher than that of highly paraffinic fuels with similar RON and the spark advance was several crank angle degrees less for constant combustion phasing.

Kolodziej, Christopher P↗
DOE OSTI2026

FPMS_XPeRT_INL_Poster

As nuclear energy expands and experimental programs increasingly rely on the facilities at Idaho National Laboratory (INL) for reactor and fuel testing, research capabilities must also expand to meet these demands. A new Fission Product Monitoring System (FPMS) has been deployed at the Advanced Test Reactor (ATR) at the Auxiliary Lead-out Experiment (ALE) House to support this expanding fuel testing mission. By tracking gaseous fission products releases from test fuel in near real-time, release rates, calculated from FPMS data, can be used to characterize the effectiveness of fuel cladding, especially for Tri-structural Isotropic (TRISO) fuel concepts. The new iteration of the FPMS supports up to 14 fission product monitors for online fission-product tracking via gamma-ray spectroscopy of the experiment’s effluent gas. Each monitor consists of a nominally 10% HPGe detector housed in a copper-lined lead shield with a warm gas trap. The new system features gamma-ray count rate information with a five-second temporal resolution and provides isotopic activity every five minutes, capable of resolving multiple overlapping fission product releases over a broad range of activities in near real-time. This work includes data from ATR cycle 175D data to demonstrate these capabilities. The hourly resolution data shows general trends and significant releases over the cycle, while the 5-minute resolution data allows for a more detailed examination of events due to unexpected particle releases.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2022

Research and Test Reactor Fuels

PRO-RR is the research reactor focused program element of the broader Proliferation Resistance Optimization program (PRO-X) under the National Nuclear Safety Administration (NNSA) in the U.S. Department of Energy (DOE). PRO-X provides a framework for integrating proliferation resistance in nuclear system designs to minimize weapons usable nuclear materials (WUNM) production and diversion pathways while optimizing systems performance for peaceful use missions. PRO-RR applies the PRO-X mission objectives to research reactor system design. This document serves as one of the foundational documents for the PRO-RR-Fuel System Design technical team by documenting current research reactor fuels usage. The PRO-RR-Fuel System Design technical team consists of subject matter experts from Argonne National Laboratory (Argonne) and Savannah River National Laboratory (SRNL). In order to determine the preferred fuel of use in upcoming research and test reactors to optimize proliferation resistance, performance, and safety, it is useful to assess the fuels that have been used in the past, or are currently in use. This report reviews the historical and current fuels used in research and test reactors to inform future fuel selection. Chapter 2 discusses the low-enriched uranium (LEU) fuels currently in use in terms of thermal power level and utilization of the reactor. Chapter 3 summarizes the fabrication processes for common fuel types. Chapter 4 discusses in detail the fuel types in use in research and test reactors. A review of the cladding types in use is presented in Chapter 5, and a historical review of research and test reactor fuel fabricators is presented in Chapter 6. The data collection strategy used the International Atomic Energy Agency (IAEA) research reactor database [1] as a starting point. Information on the fuel used was gathered on research reactors (other than critical assemblies) that were listed as operational, planned, or in temporary shutdown in the IAEA database. Data on the fuel type, geometry, enrichment, uranium loading, cladding type, and fabricator were collected for each of the reactors available in the public domain. Sources of data included conference papers, journal articles, and facility and fabricator websites. Data on research reactors operating on LEU fuels are presented in Appendix A, while Appendix B presents data collected on all reactors at the time of publication of this report. Appendix C presents data collected on reactors that were part of the M3 research and test reactor conversion program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

FCET Solid Oxide Fuel Cell Testing and Development (CRADA 526) (Final Report)

Pacific Northwest National Laboratory (PNNL) tested electrolyte coatings from Fuel Cell Enabling Technologies, Inc. (FCET) for use in solid oxide fuel cells (SOFCs). The key technology held by FCET is a process to deposit thin layers of oxide materials, less than 1 μm in thickness. The range of possible materials that can be deposited with their method is broad, but this project focused on the gadolinium-doped ceria (GDC) and yttria-stabilized zirconia (YSZ) electrolytes for SOFCs. Thin, gas tight electrolyte membranes have been a long-sought target in SOFC research. The thinner the electrolyte, the lower the cell resistance, and the higher performance of the cell (or the lower the operating temperature). A YSZ thickness of 1 μm would be a step change from the state-of-the-art, tape-cast electrolytes (~10 μm). An in-house prototype SOFC stack from FCET was first tested. The sealing geometry of the prototype stack was determined to be problematic, and testing shifted to button cells. Anode-supported solid oxide electrolysis cell (SOEC) button cells without an electrolyte layer were produced at PNNL and sent to FCET for coating with electrolyte. Three cells were tested with a gadolinium-doped ceria (GDC) electrolyte applied via spin coating. GDC was chosen for its conductivity at lower temperatures than YSZ. All three cells failed during initial reduction under hydrogen at 600°C. Testing then shifted to YSZ, which is the standard SOFC electrolyte. Several button cells were coated with YSZ and examined with scanning electron microscopy (SEM). A promising coating of ~1 mm thickness was observed under SEM. A similarly coated button cell was tested and failed similarly to previous tests during reduction at 600°C. The YSZ coating appeared dense and uniform in SEM analysis. The roughness of the underlying Ni/YSZ anode is on the order of 1 μm, and that may have compromised the gas-tightness of the coating. Further development is warranted to understand and refine the coating process. Thin YSZ applied via this spin-coating technique could be used as a low-cost, drop-in replacement in large-scale SOFC manufacturing processes, improving cell performance and lowering the cost per watt of SOFCs.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2024

AGR Fuel Oxidation Testing Summary and Plans

This presentation describes the goals, scope, and plans for oxidation testing of (primarily irradiated) AGR TRISO fuels and related materials (such as graphite and graphitic matrix material). This presentation will be given at the annual program review meeting for the Gas-cooled Reactor program hosted at INL starting on July 16.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Halogenation of used aluminum matrix test reactor fuel – a bench-scale demonstration with surrogate materials

In this work, experiments with surrogate materials were performed at bench scale to demonstrate a halogenation technique applicable to treatment of used aluminum matrix test reactor fuel. The technique involves dissolution and separation of aluminum from used aluminum matrix test reactor fuel in molten-halide salt systems prior to treatment and disposition of the fuel’s uranium and fission products. Demonstration of the halogenation technique was performed with neodymium metal as a non-radiological surrogate for uranium metal. Experiments involved blending forms of aluminum and neodymium metal with ammonium and lithium chloride or ammonium and lithium bromide, which upon heating decomposed into ammonia gas and the respective hydrogen chloride or bromide gas. The latter reacted with the metals to form the respective aluminum and neodymium halides. At elevated temperatures, aluminum halides gasified away from the respective neodymium halides, which fused with their respective lithium halides. Samples of fused and distillate salts were collected and analyzed, yielding extents of aluminum removal that ranged from 94.5–98.2% for chlorination runs and 91.4–97.8% for bromination runs. No neodymium was detected in the distillate fractions. Some experiments were repeated with excess reactants, and a portion of aluminum chloride distillate was processed into a consolidated waste form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Modeling and Analysis for Spent Nuclear Fuel Seismic Testing

Note: This is the final public release version of PNNL-31671 Draft, which was previously released to the sponsor for review. The intent is the public release version will be PNNL-31671. There are no significant changes from the earlier draft version. This unlimited distribution report is the deliverable for M3SF-21PN010202014: Modeling and Analysis for Spent Nuclear Fuel Seismic Testing. This report summarizes the modeling, analysis, and test plan support completed by Pacific Northwest National Laboratory for the spent nuclear fuel dry storage system seismic test plan through May of 2021. Test plan preparation is planned to continue until the seismic test is completed in July of 2022. This report covers preliminary structural dynamic model development and computer-aided design of test hardware. Based on preliminary modeling, the strongest earthquakes under consideration in this test program provide mechanical loading on the spent nuclear fuel that is comparable to the 30 cm cask drop scenario, although the loads do not appear to be strong enough to cause significant permanent deformation of the fuel assembly spacer grids. The potential for grid deformation during the test will be assessed when the proposed shake table motion becomes available. The weakest earthquakes considered in this test are expected to be comparable to the mechanical loads witnessed in the multimodal transportation test of 2017. The horizontal canister system is predicted to provide nearly-uniform loading condition on the fuel assemblies it contains, while the vertical cask system is predicted to cause non-uniform dynamic loads on the fuel assemblies. In the horizontal case, gravity keeps the fuel assemblies in contact with one basket wall surface unless the loads are strong enough to cause a separation. In the vertical case, the fuel assemblies are relatively long and slender, and seismic motion in the anticipated test range is predicted to cause the assemblies to lean, tilt, and impact the basket walls throughout the seismic event. These gap closures are a nonlinear force transmission condition, so the vertical system is expected to have more variation and variability than the horizontal system. While the horizontal system is expected to have a relatively more linear response than the vertical system, there is still the potential for nonlinear behavior in the horizontal system because the fuel assemblies are free to slide, bounce, and impact the basket walls if the seismic loads are strong enough. One major conclusion of this study is that the use of mixed fuel assemblies in the canister will be acceptable. There are differences in overall system response when the mass, center of gravity, or gaps are changed, but the changes in response are within the bounds of a system that contains completely homogeneous fuel assemblies. One important observation is that the loading for each individual fuel assembly within the vertical canister is expected to be different from that of the others because of nonlinearities in the system. The horizontal canister system is expected to have a more uniform and more predictable response than the vertical canister, making variations in fuel assembly characteristics easier to account for. One important recommendation that comes from this modeling work is to repeat some of the strongest shake tests at different angles, particularly for the vertical cask system. Modeling predicts that strong seismic motion will cause the fuel assemblies to close initial gaps and impact the fuel basket walls when the canister is in the vertical cask configuration. Each shake test will impose a pre-defined three-dimensional time history on the cask system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Ultra-compact system for characterization of physical, chemical and ignition properties of fuels

The present disclosure relates to a miniaturized fuel laboratory system that makes use of a housing, a processor housed within the housing, and a fuel inlet port supported from the housing for receiving a quantity of fuel to be used as a fuel test sample. The system may also have at least one fuel sensor housed in the housing in communication with the fuel inlet port for receiving the fuel test sample and carrying out combustion thereof. An electronic component may be housed in the housing, which enables communication with an external remote subsystem. A database may be incorporated which contains at least one of stored fuel characteristics or stored fuel analysis models, accessible by the processor. The processor may use fuel oxidation information generated by the fuel sensor, and at least one of the stored fuel characteristics or stored combustion models, to determine at least one fuel characteristic of the fuel test sample.

McNenly, Matthew J.↗
DOE OSTI2023

A Review of Advanced Test Reactor Fuel and Assessment of Its Compatibility with the ZIRCEX Chlorination Process

Advanced Test Reactor (ATR) fuel has been identified as a resource for high-assay low-enriched uranium (HALEU) production. A survey was performed on the published literature describing ATR fuel. The geometry of the fuel is complex; different parts of the fuel compact experience differing neutron flux and burnup. The literature is sparse, and access is controlled. Therefore, fundamental studies of fuel reprocessing must use a model fuel that represents the main chemical and structural features. Advanced chlorination, or chlorination with sulfur-chlorine bearing reagents is being investigated as way to separate the fuel from metal matrix alloys. A UAl x alloy will be fabricated with x = 3, 4, and 5. The potential chlorination of individual UAl x intermetallics will be assessed in the advanced chlorination process of Al-8001 and Al-6061 as well as a representative mixture. Initial studies will track the alloying elements of the Al, which are Si, Fe, Cu, Mn, Mg, Cr, Zn, and Ti, in addition to the U itself. Further studies will include fission product simulants. Because advanced chlorination solvents include sulfur, the chemistry of sulfur with major and minor constituents will also be investigated. The experimental work accompanied by neutronic calculations will allow the assessment of the feasibility of advanced chlorination to separate aluminum from uranium. If bench-scale testing appears promising, then small-scale tests in shielded facilities with irradiated cladding, lightly irradiated fuel, and spent nuclear fuel are recommended to track the complete inventory of fissile actinides, fission product impurities, and reagent solids and liquids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Heavy Duty Hydrogen: Reference Station, Fueling Performance Test Device Concepts, and Station Capacity Model (CRADA Final Report)

This project will provide valuable information on (1) reference station design, (2) exploration of design concepts for a fueling performance test device, and (3) modeling of station capacity. The work applies to heavy-duty (HD) hydrogen fueling stations with large dispensing capacity and high flowrates servicing HD hydrogen trucks such as class 8 trucks in long haul applications. Additionally, a 4th area will provide near-real-time verification of fuel quality with on-site hydrogen contaminant detectors (HCDs) for use at both light-duty (LD) and HD stations. The work leverages national lab capabilities including staff and equipment at SNL, NLR, and ANL with collaboration and funding cost share from California agencies (CEC, SCAQMD, and GO-Biz). This project will provide tools and information that lead to more efficient design, acceptance, and commissioning of these larger capacity, higher flowrate stations serving HD applications. The HCD work will benefit both LD and HD stations. Note: The 4th area (HCD) mentioned above continued outside of this CRADA and will not be included in this report. There are no tasks included in this CRADA for HCD and the summary text for it should have been removed prior to this CRADA contract execution but remained in error and is included here for completeness.

08 HYDROGEN↗
DOE OSTI2021

FCET Solid Oxide Fuel Cell Testing and Development: CRADA 526 [Abstract only]

The purpose of the proposed project is for PNNL to test the performance of prototype solid oxide fuel cells (SOFCs) created by FCET. Such testing will provide FCET with independent performance data that can be communicated to potential clients and/or investors. Additionally, PNNL will collaborate with FCET on design changes to improve fuel cell performance. Intellectual property developed in this way will benefit FCET with improved products to market, and PNNL through royalty revenue. The key technology held by FCET is a process to deposit extremely thin layers of oxide materials, from 10-50 nm in thickness. The range of possible materials that can be deposited with their method is broad, but this project will focus on the yttria-stabilized zirconia (YSZ) electrolytes for SOFCs. Thin, gas tight YSZ membranes have been a long-sought target in SOFC research. The thinner the YSZ, the lower the cell resistance, and the higher performance of the cell (or the lower the operating temperature). 10-50 nm would be a major step change in YSZ thickness from the state of the art. PNNL can team with FCET on future R&D projects and push this technology forward to improve energy efficiency and reduce carbon emissions.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2023

Sister Rod Destructive Examinations (FY22) Appendix E: Mechanical Testing - Spent Fuel and Waste Disposition

As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This report documents the status of the ORNL Phase 1 DE activities related to the mechanical testing of selected sister rods in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Summary of Factory Fueling and Testing Risks to Support the Transformation of Regulatory Requirements for Microreactors

Microreactors have the potential to open new markets and applications for nuclear energy. New manufacturing and operational regimes, such as factory fueling and testing prior to site installation, are being proposed to meet these new markets. The radiological hazards associated with traditional commercial power reactors are not the same as the hazards of a reactor being constructed in a factory. Currently, a reactor operating license specified under 10 CFR Part 50, Part 52, or (in the near future) Part 53 is required prior to loading fuel in the reactor. This is logical for plants that are immediately to begin full-power operations soon after fuel loading. However, for microreactors, this may not be the case. Some deployment concepts may stage reactors as “ready to go” in a factory prior to site installation and startup. This approach is being considered in the case of diesel generator replacements, mobile or roaming mining operations, and emergency response situations, such as after a major disaster event. Therefore, it is necessary to evaluate hazards assuming low power operation for a short irradiation time. This report investigates the primary hazards unique to microreactors associated with low power reactor and physics tests necessary to confirm core safety assumptions, neutronic and core design performance parameters, and other assumptions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

HIGH BURNUP FUEL-COOLANT INTERACTION ANALYSIS SUPPORTING FUEL SAFETY TESTING AT IDAHO NATIONAL LABORATORY

In the near future, experiments on HBu fuel under loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA) conditions will be performed within the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL). These experiments will be performed using the Transient Water Irradiation System for TREAT (TWIST) experiment vehicle. To support these experiments, analysis of fuel-coolant interaction (FCI) energetics is underway. This paper discusses FCIs in the context of light water reactor (LWR) safety, differentiating between the severe accident focus of commercial reactors and experimental RIA test programs where FCIs have occurred. However, it is highlighted that as the nuclear industry aims for increased burnup limits, the FCI events observed in RIA test programs may become relevant to commercial LWR safety analysis. The paper then presents developments to the UW-FCI computer program to enable simulation of FCIs initiated by solid fuel particles dispersing into the coolant during RIAs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Influence of NOx Chemistry on the Prediction of Natural Gas End-Gas Autoignition

Research on the influence of NOx (NO/NO2) chemistry on HC oxidation has shown that NOx plays an important role in the promotion/inhibition of HC autoignition [1-6]. Current data for n-heptane, iso-octane, ethanol, and toluene [3, 4] have shown that NO chemistry inhibits autoignition at low temperatures and concentrations. However, at medium and high temperatures and low concentrations, autoignition is promoted, decreasing the Negative Temperature Coefficient (NTC) effect. This autoignition promotion, however, becomes weaker as NO concentrations are increased. Additionally, fundamental kinetics studies have suggested that NOx chemistry also promotes the autoignition of smaller HC, such as CH4 [7-10], C2H6 [11-14], and C3H8 [15]. Additionally to the chemical kinetics studies, a growing body of evidence supports the importance of NOx chemistry on SI engine simulations [16, 17]. Foong et al. [18] suggested that the addition of NO in their models significantly improved the agreement between simulations and experiments where the onset of knock was advanced due to the presence of NO, and Morganti et al. [19] showed that NO addition in the residual-gas was necessary for their models to obtain good agreement with experimental data. Lastly, Mohr et al. [20, 21] investigated the effect of Natural Gas (NG) reactivity and EGR substitution rate and composition on homogeneous ignition delay, flame speeds, and EGAI for stoichiometric NG/oxidizer/EGR blends. They observed that the addition of EGR composed of only inert species (CO2 and Ar) suppressed homogeneous autoignition in all NG fuels tested under all conditions. However, for all NG fuels tested, increasing EGR rates with reactive species (Ar, CO2, CO, and NO) promoted homogeneous autoignition under all conditions, thus reproducing the same trends for various NG fuel compositions that were observed in other studies and fuels. For this reason, a study is being conducted to analyze the effects of NOx chemistry on the prediction of NG/air/EGR homogeneous autoignition. This study has been divided into two tasks. First, the addition of NOx chemistry to the ARIES82 mechanism [20] is being validated using homogeneous ignition delay data collected by Mohr [20, 21]. Second, once the new modified mechanism with NOx reactions has been validated, it will be employed on engine multi-dimensional modeling presented in [22, 23] to analyze the influence of NOx chemistry on NG EGAI. Results thus far have shown that homogeneous ignition delay calculations are sensitive to NOx chemistry, where the modified mechanisms captured well all trends and closely matched the homogeneous ignition delays observed by Mohr et al. Also, the results in this work suggest that NOx chemistry is necessary to correctly capture the trends in NG homogeneous autoignition when EGR with reactive species is used since the opposite behavior is observed when only inert species are considered for the EGR composition.

03 NATURAL GAS↗
DOE OSTI2022

Chemical Kinetics Underlying the Sooting Tendency and Auto-Ignition Characteristics of Linear, Branched, and Cyclic Ether Compounds

Biofuels present opportunities for improving the performance and reducing emissions from internal combustion engines by incorporating oxygenated functional groups to the fuels. Among various oxygenates, ethers have been recognized as promising candidates for an alternative to conventional diesel fuel owing to their higher reactivity and lower sooting tendency. The detailed guidelines for designing ethers, however, have not been fully discussed, even though their combustion characteristics are sensitive to the molecular structure. This study was devoted to exploring the structure-property relationships, particularly focusing on the cetane number and yield sooting index, using five linear, branched, and cyclic ethers: di-amyl-ether, 4-butoxy-heptane, 3,3-dimethyl-oxetane, 2-ethyl-4-methyl-1,3-dioxolane, and 2-isopropyl-4-methyl-1,3-dioxolane. First, we examined the chemical kinetics underlying the sooting tendency of the test fuels. The combustion product distribution was measured from flow reactor experiments at 750-1100 K, F=3, at atmospheric pressure. As a result, it was revealed that the sooting tendency is closely related to the size of hydrocarbon intermediates in the high-temperature regime (>1000 K); that is, larger hydrocarbons lead to more soot precursor formation. The underlying chemistry determining the size of the hydrocarbons from the tested fuels was analyzed using reaction pathway analysis and quantum mechanics calculations, which showed that the branched and cyclic ether structures form abundant C3-C4 compounds. Moreover, the auto-ignition characteristics of the test fuels were studied using the flow reactor at low-temperature (400-700 K) and F=1. We found a clear difference in the combustion-product distribution from high and low reactivity fuels, which was then correlated to the systematic analysis of the key reaction energy barriers with the varying molecular structure.

BIOMASS FUELS,INORGANIC, ORGANIC, PHYSICAL, AND AN↗
DOE OSTI2020

Drop Analysis of the Advanced Test Reactor Fresh Fuel Shipping Container with Heavier Low-Enriched Uranium Fuel Contents

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is a rectangular stainless steel container used for shipping radioactive material. The container is described in the ATR FFSC Safety Analysis Report (SAR). Per the ATR FFSC SAR, the ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.4, and was originally designed to transport high enriched uranium (HEU) reactor fuel elements for the Advanced Test Reactor (ATR), the Advanced Test Reactor Critical (ATRC) facility, the Massachusetts Institute of Technology Reactor (MITR), and the University of Missouri Research Reactor (MURR). The Department of Energy, National Nuclear Security Administration’s (NNSA), Office of Material Management and Minimization (M3) is working with the Idaho National Laboratory (INL) to develop and qualify new low enriched uranium (LEU) fuels and technologies for use in the ATR, ATRC, MITR, and MURR reactors. The LEU fuel elements will weigh significantly more than the current HEU designs and, combined with their associated Fuel Handling Enclosures for packaging, some configurations will exceed the 50 lbf used in the ATR FFSC qualifying drop tests. There are LEU versions of MITR, MURR, and ATR fuel elements. However, for this evaluation, drop analysis of the ATR FFSC with only the heavier ATR Low Enrichment (LOWE) fuel element is considered in this evaluation because the LOWE fuel element is the heaviest of the considered LEU fuel elements. The ATR HEU fuel element and the ATR LOWE fuel element are identical in every design aspect except for the fuel meat inside the 19 fuel plates. The LEU fuel meats are made using a U-10Mo high-density foil rather than uranium dispersed in aluminum in the HEU fuel elements. The high density of the uranium in the LEU fuel meat increases the LOWE fuel element weight to just under 44 lbf (versus the 22.1 lbf weight of the tested ATR HEU fuel element). ATR fuel elements are placed in a thin-gauge aluminum weldment called a "Fuel Handling Enclosure" during packaging. The Fuel Handling Enclosure is used to cover and protect the element during loading and unloading operations. The ATR Fuel Handling Enclosure weighs about 15 lbf per the drawings in the ATR FFSC SAR and the weight is accounted for in this evaluation. Transporting the heavier LEU fuel elements require evaluation of two issues. The first is the effect of the increased mass of the LEU fuel elements on the survivability of the ATR FFSC package following the requisite drop qualifications. The second is the effect of the increased mass of the fuel plates on the fuel element during the same drops. The ATR FFSC containing an ATR HEU fuel element in an ATR Fuel Handling Enclosure was physically dropped multiple times to qualify the container as a Type AF-96 package. The ATR FFSC SAR describes the drop tests performed with an actual ATR HEU fuel element weighing 22.1 lbf contained in a 14.3 lbf Fuel Handling Enclosure for a total payload of 36.4 lbf. Those drop tests showed that the ATR FFSC maintained containment of the ATR HEU fuel element and the fuel element was not significantly damaged. (Containment herein is not defined as a leak tight but is retention of the radioactive contents.) The purpose of the evaluation is to analytically show that, for a similar set of tests, the ATR FFSC maintains containment of the heavier ATR LOWE fuel element and to assess the damage to the fuel element during the drops. The approach was to create finite element analysis (FEA) models that produce the same results as the physical drops. Those models were then used as the benchmarks for the follow-on analyses using the heavier contents. FEA models of the drops of ATR FFSC using up to a 115 lbf fuel element were run and evaluated. Likewise, drops of a LOWE fuel element weighing 44 lbf in the ATR FFSC were run and evaluated. It is important to note that this report was done at the quality level necessary to be included in a nuclear facility safety basis. However, it is not the intent of this report to conclude the suitability of the ATR FFSC for transporting the heavier payloads. This report only describes the results of the FEA as related to the required drop scenarios. Incorporation of the FEA into the safety basis will be evaluated by the ATR FFSC design authority. The physical drop tests of the HEU fuel element and FEA drop analysis for the LOWE fuel element showed noteworthy damage to the fuel plates. An aluminum protective block was conceived to mitigate the damage. The concept requires the blocks to be placed in the fuel element between the end boxes and fuel plates. Additional FEA drop analyses were performed using the protective block. The addition of the blocks is primarily intended to mitigate the damage to the LOWE fuel element fuel plates. However, FEA drop analyses of the ATR HEU fuel element with the blocks were also performed and included for information.

42 ENGINEERING↗
DOE OSTI2022

Thermal Modelling of Advanced Test Reactor Fuel in a Generalized Dry Storage System

Star-CCM+, a computational fluid dynamics (CFD) software was used to conduct modeling and simulation of the thermal performance of a dry storage configuration consisting of Department of Energy Standardized Canisters (DOESCs) loaded with aluminum-clad spent nuclear fuel (ASNF.) The configuration includes nine DOESCs loaded with Advanced Test Reactor ASNF contained within a stainless steel overcanister centered in a ventilated, concrete overpack. The simulations were used to estimate the maximum temperatures reached by backfill gases inside the overcanister and DOESCs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗