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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 OSTI2024

Passive Temperature Sensors for Nuclear Applications

Thermocouples are generally used to provide real-time temperature indications in instrumented tests performed at material and test reactors. Passive temperature monitors, such as Silicon Carbide (SiC) and melt wires, may be included in such tests as an independent technique of detecting peak temperatures experienced during irradiation. In less expensive static (drop-in) capsule tests, which have no leads attached for real-time data transmission, melt wires, and SiC temperature monitors (TMs) are essentially the only possibility for peak temperature indication. A melt wire involves placing materials (wires) of a known composition and melting temperature in a test. An inventory is maintained at Material Science Laboratory (MSL) for melt wires ranging in temperatures from 30°C to 1500°C. Unfortunately, melt wires are limited in that it can only detect whether a single temperature is or is not exceeded (melt wire melted or not). SiC TMs, which can also be used to detect peak irradiation temperatures, are advantageous because a single monitor can allow to determine the peak temperature reached within a relatively broad range (100 – 1200°C) resulting in accuracies within ±20°C. Irradiation temperature is determined by measuring a property change after isochronal annealing or during a continuously monitored annealing process using specialized equipment at MSL. Recent research has produced a passive monitor known as sublime temperature monitor. This passive sensor has the capability of recording temperature gradients and pinpointing exactly where a temperature is located along that gradient. Long measurement lengths are achieved with very high accuracy in the location of desired temperature measurements (±2 mm over a 1 m span); however, this sensor has not been deployed in a nuclear reactor. This article will focus only on passive temperature sensors currently being researched and implemented under the Advanced Sensors and Instrumentation (ASI) program at Idaho National Laboratory (INL).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High flux boiling heat transfer enhancement using triangle shaped vertical walls in two-phase microchannel heat exchangers

Two-phase heat exchangers allow heat loads to be dissipated with lower fluid flow rates without inducing thermal gradients, and high surface area microchannels further reduce the thermal resistance of these heat exchangers. However, there are many practical limits that prevent widespread use of two-phase microchannel heat exchangers. Many authors have looked to address these practical limits by treating the channel floors or creating artificial nucleation sites. Few works look to enhance heat transfer in the walls of channels with hydraulic diameters below 200 μm. This work is aimed at increasing the efficacy of high aspect ratio silicon microchannels that use R134a as the working fluid. The 125 parallel channels are 40 μm wide, 200 μm deep, and 2 mm long and contain a 15 μm wide, 150 μm long restriction at each inlet. The thermal resistance and dryout characteristics of traditional, plain-walled channels are compared to channels with 2 area enhancement patterns each of which increases the area available for heat transfer by 41%. Furthermore, a resistance to dryout is seen in both area-enhanced test sections, and a 15.5% reduction in thermal resistance is shown for the area-enhanced part with a 6 μm triangular pitch. Only a 7.8% performance increase is seen in the test section with a 3 μm pitch. Both area enhancement patterns induce an average 27% increase in pressure drop for all tests. Heat transfer characteristics are solved for using a finite element analysis (FEA) model. By examining heat transfer coefficients and local heat fluxes, it is postulated that the area enhancement reduces heat transfer coefficients and the performance increase is caused by increasing the number of active nucleation sites.

42 ENGINEERING↗
DOE OSTI2023

ADDITIVE MANUFACTURING AND REGULATORY TESTING OF CANISTERS FOR SPENT NUCLEAR FUEL MANAGEMENT

The back end of the fuel cycle focuses on the interim storage, transportation, and final disposition of the spent nuclear fuel from nuclear reactors. Commercial light water nuclear power stations across the United States operate with fuel assemblies being irradiated for up to 6 years (planned) in the reactor pressure vessel. After their planned irradiation, the fuel assemblies are moved to a spent fuel pool within the facility complex. After the spent nuclear fuel is removed from the fuel pool, it is typically inserted into a welded metal canister that can be transferred between overpacks for storage, transportation and possibly disposal. Most dry storage systems being used by industry today use dual purpose canisters (DPCs), designed for use in storage and transportation overpacks, but not specifically designed for disposal. Triple purpose-canisters, designed for disposal in addition to storage and transportation, have also been researched. Traditional manufacturing methods for spent fuel canisters involve fusion welding along the length or circumference of the canister which results in high tensile residual stresses in the joint weld zone (WZ) and heat affected zone (HAZ). In this paper, spent fuel canister designs were printed by wire arc additive manufacturing (AM) using the 316L SS welding wire to demonstrate: 1. Feasibility of spent nuclear fuel canister fabrication using this advanced manufacturing method, 2. Dynamic response of the additive manufacture canister design when subjected to the federally mandated Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) physical tests for Type B packages. This paper will focus on the canister printing design and structural tests. The AM 3D printed design, regulatory testing, and post-test evaluation of the canister tested to the 10 CFR 71.71 and 10 CFR 71.73 requirements will be presented. One AM canister design was subjected to the penetration, free drop, and puncture test. Before and after the dynamic structural tests, the AM canister design was scanned with a handheld scanner to capture a 3D CAD geometry to compare to the 3D printed canister design in the deformed shape. The scanned geometry was sectioned in areas with deformation and the cross-section profile was measured to determine accurate and repeated results of the deformed shape of the AM canister design.

Martinez, Oscar↗
DOE OSTI2023

Uncertainty Quantification for Component Modeling Using the Discrete-Direct Approach

Threaded fastener behavior can be an important aspect of complex component and system behavior, but there is no one-size-fits-all finite element analysis technique. Proper modeling of threaded fastener joints requires careful consideration of many details, from test setup and data acquisition to constitutive modeling and uncertainty quantification approaches. This report details analysis of a “mini-radax” bolted-joint exemplar where a Discrete-Direct uncertainty quantification approach is employed to evaluate margin of the component. The mini-radax geometry is tested to failure on a drop table, and single-coupon tests of individual fasteners serve as foundational data for the analysis. Analysis predictions complement the test data well and provide additional context for engineering decision-making.

42 ENGINEERING↗
DOE OSTI2023

Recent Radioactive Material Package Testing Experiences at Oak Ridge National Laboratory

Oak Ridge National Laboratory (ORNL) was a pioneer in the testing of radioactive material shipping packages and has been performing such tests for over 50 years. Currently, except for Type B thermal tests, testing is performed at the purpose-built Packaging Research Facility that is located at the National Transportation Research Center - a U.S. Department of Energy User Facility. Packages used for the transportation of Type B quantities of radioactive materials must be capable of meeting normal conditions of transportation (NCT) requirements and hypothetical accident conditions (HAC) requirements as defined in Title 10 Code of Federal Regulations (CFR) Part 71. Evaluation of package designs to these criteria typically involves a combination of analysis and physical testing. Required physical tests include free drop, crush, puncture, penetration, compression, vibration, water spray, water immersion, and thermal. The Oak Ridge National Laboratory (ORNL) Packaging Evaluation Facility (PEF) at the National Transportation Research Center (NTRC) provides an inclusive testing capability for a wide variety of packages needed for shipment of radioactive materials. Since PATRAM 2019, ORNL has performed several sets of tests including testing of two designs of the Defense Program Package (DPP-3 and DPP-1) and several Special Form Capsules. This paper provides an overview of the testing performed as well as a summary of advancements made to the ORNL package testing program during that time.

Martinez, Oscar↗
DOE OSTI2023

Qualification and Commissioning of Helium Flow Loop Experiment for Blanket Design Measurements

Sufficient cooling of plasma-facing materials remains an outstanding challenge in the design of fusion reactor blankets in commercial power demonstration plants. Due to its chemical inertness and low neutron interaction cross section, pressurized helium is a candidate coolant fluid for such systems; however, helium has a small thermal mass compared to liquid coolants, potentially reducing heat removal performance. To address this need, a number of heat transfer enhancements have been proposed to improve the cooling efficiency of such components, thereby decreasing pumping power needs and improving overall plant efficiency. Toward this end, a helium flow loop experiment (HFLE) has been designed and commissioned to test advanced passive heat transfer enhancements in unit-cell test sections, providing necessary data for model validation and subsequent system design. The HFLE is designed to provide flow of pressurized (up to 4 MPa) helium at flow rates up to 80 g/s, enabling heat transfer and pressure drop measurements in test pieces at Reynolds numbers in excess of 180 000. To explore the effects of novel and complex heat transfer enhancements, test sections are produced via additive manufacturing, providing geometries not typically obtainable by conventional machining. Here in this work, we present results from HFLE commissioning and the initial thermal-hydraulic tests of an additively manufactured rifled-rib test section. Results are compared to smooth pipe correlations, and plans are described for future HFLE measurements. These preliminary experiments indicate the utility of the HFLE for heat transfer enhancement testing and simulation validation activities.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

CONSTRAINT-INDEPENDENT CONSTANT CTOA DETERMINATION FOR DUCTILE STABLE CRACK GROWTH

Crack tip opening angle (CTOA) has been used as a reliable fracture toughness parameter for decades to characterize stable ductile crack growth for thin-walled aerospace structures in the low-constraint conditions. Recently, the CTOA parameter was also applied to the pipeline industry, and a CTOA test standard ASTM E3039 was thus developed for testing a critical constant CTOA. Research showed that the constant CTOA can reasonably describe fracture toughness required to arrest a dynamic crack propagation for a modern gas pipeline. However, the CTOA fracture criterion requires constraint-independent CTOA toughness against stable ductile crack growth. ASTM E3039 recommends a drop weight tearing test (DWTT) specimen for CTOA testing. Since a shallow crack is used, DWTT measured CTOA may depend on constraint level at the crack tip. To understand if it is the case, this paper evaluates the critical CTOA for a set of fracture toughness tests on single edge notched bend (SENB) specimens with shallow and deep cracks based on four CTOA estimation models. In which, the Ln(P)-LLD linear fit model is similar to that used by ASTM E3039 in the CTOA calculation. Fracture test data for X80 pipeline steel and HY80 structural steel are considered in the CTOA evaluation. The results show that the four CTOA models can determine a crack size-independent constant CTOA over stable ductile crack growth for the SENB specimens. As a result, CTOA determined by ASTM E3039 is constraintindependent and transferable to use for an actual crack propagating in a gas pipeline.

Zhu, Xian-Kui↗
DOE OSTI2021

X-ray Imaging-Based Void Fraction Measurement in Saturated Flow Boiling Experiments with Seawater Coolant

Recent studies have shown that the presence of dissolved salts in water can exhibit peculiar flow boiling and two-phase flow regimes. Two-phase flow and convective flow boiling are typically characterized with the help of void fraction measurements. To quantitatively improve our understanding of two-phase flow and boiling phenomenon with seawater coolant, void fraction data are needed, which can not be obtained from optical imaging. In this paper, we present experimental void fraction measurements of saturated flow boiling of tap water and seawater using X-ray radiography. X-rays with a maximum energy level of 40 KeV were used for imaging the exit region of the heated test section. At lower heat flux levels, the two phase flow in seawater was bubbly and homogeneous in nature, resulting in higher void fractions as compared to tap water. With an increase in heat flux, the flow regime was similar to slug flow, and void fraction measurements approached similarity with tap water. The predicted pressure drop using the measured void faction shows good agreement with the measured total pressure drop across the test section, demonstrating the validity of the measurement process.

Eckels, Seth↗
DOE OSTI2021

Structural dynamics modeling of spent nuclear fuel during hypothetical package drop events

The response of spent nuclear fuel (SNF) to hypothetical package drop events is of particular interest in the scope of spent fuel storage and transportation because of the mechanical shock encountered in such scenarios. Previous testing and modeling by the U.S. Department of Energy has demonstrated that the shock and vibration environment of normal shipping and handling conditions (excluding package drop events) is relatively benign and does not challenge the integrity of spent nuclear fuel. Cask drop events are worth considering because SNF packages are required to withstand free drops onto unyielding surfaces as part of their licensing basis. The acceleration experienced during drop events can be orders of magnitude higher, and thus more advanced models are needed to encompass potential nonlinear behavior of the fuel, such as spacer grid buckling and rod-to-rod impact. This work describes a number of finite element models developed to calculate the response of spent nuclear fuel to various hypothetical drop events that have been validated by package and fuel assembly drop tests conducted in the last decade. Sensitivity of the model response to factors such as package drop orientation, secondary impacts, and irradiated material properties as well as their potential impacts to fuel cladding integrity, was also investigated. Cask drops are not expected as a regular occurrence during SNF transportation, but this work helps raise the understanding of SNF mechanical loads to the point of consistency with the package design requirements.

Kadooka, Kevin↗
DOE OSTI2022

Measurement of Convective Heat Transfer Coefficients with Supercritical CO2 in Novel Additively Manufactured Helically Patterned Pin Fin Tubes Using the Wilson Plot Technique

This paper describes the measurement of convective heat transfer coefficients and friction factors for sCO2 flowing in pin-fin patterned pipes in the Heat Exchange and Experimental Testing (HEET) facility at the US DoE’s National Energy Technology Laboratory (NETL) in Morgantown, WV. The measurement procedures in the HEET rig were validated by conducting benchmark tests with smooth stainless-steel tube and comparing the results with published correlations for Nusselt number (Nu) and friction factor. Over typical Reynolds number range in sCO2, the measured Nu and friction factors were within 7% of classical correlations for smooth tube flow.The candidate pin fin patterned pipes were additively manufactured (AM) at the Oak Ridge National Laboratory. The pins were circular or elliptical in cross-section. Pin length to diameter aspect ratios were 1.33 and 2, while the pin diameter to tube diameter ratio was 0.188 and 0.125. Tests were performed for ReD varying from 6.9×104 to 2.2×105 and at conditions equivalent to the low pressure outlet (8.69 MPa, 361 K) of the low temperature recuperator (LTR) in an indirect sCO2 power cycle. The Wilson plot technique was utilized to measure the bulk heat transfer coefficients.For the better performing design (tube A, pin length to tube diameter ratio: 1.33, pin diameter to tube diameter ratio: 0.188), the local heat transfer coefficient increased by 112% relative to the Dittus-Boelter correlation at the LTR low pressure outlet. This corresponded to a 282% increase in the product of the heat transfer coefficient and the surface area. Large pressure drops across the test articles were observed.

Searle, Matthew↗
DOE OSTI2020

Heat Transfer Coefficients of Additively Manufactured Tubes with Internal Pin Fins for Supercritical Carbon Dioxide Cycle Recuperators

This paper describes the measurement of convective heat transfer coefficients and friction factors for sCO 2 flowing in additively manufactured tubes with internal pin fins at the US DOE’s National Energy Technology Laboratory in Morgantown, WV. The measurement procedures were validated by conducting benchmark tests with smooth stainless-steel tube and comparing the results with published correlations for Nusselt number (Nu) and friction factor. Over Reynolds numbers (ReD) ranging from 5 × 10 4 to 2.5 × 10 5 , measured Nu was within 5% of the Dittus-Boelter correlation and measured friction factors were within 5% of the McAdams correlation for smooth tube flow. The candidate pin fin patterned pipes were additively manufactured (AM) at the Oak Ridge National Laboratory. The pins were circular or elliptical in cross-section and were printed at a 30⁰ angle relative to the inner wall (to meet AM constraints). The pin arrangement was helical to promote enhanced heat transfer due to swirl flow. Pin length to diameter aspect ratio was 1.33, 2, and 8, while the pin diameter to tube diameter ratio was 0.188, 0.125, and 0.063. Tests were performed for ReD varying from 6.9 × 10 4 to 2.2 × 10 5 and at conditions equivalent to the low pressure (LP) outlet (8.69 MPa, 361 K) and the high pressure (HP) inlet (20.7 MPa, 350 K) of the low temperature recuperator (LTR) in an indirect sCO 2 cycle. The Wilson plot technique was utilized to measure the bulk heat transfer coefficients. For the best performing design (tube A, pin length to tube diameter ratio: 1.33, pin diameter to tube diameter ratio: 0.19), the local heat transfer coefficient increased by 136% relative to the Dittus-Boelter correlation at the LTR low pressure outlet and 194% at the LTR high pressure inlet. These correspond to a 282% and a 271% increase in the product of the heat transfer coefficient and surface area (adjusted for fin efficiency) product, respectively. Large pressure drops across the test articles were observed. For Tube Design A, the average friction factor, across the range of ReD considered, was significantly larger than the McAdams correlation at both the LTR LP outlet and the LTR HP inlet. A thermal performance factor was utilized to express the ratio of material required to build a finned heat exchanger relative to a finless heat exchanger with the same heat duty and pumping power. Tube Design A was estimated to decrease the required heat exchanger material by 13%.

Searle, Matthew↗
DOE OSTI2021

Production of desensitized, ultrafine PETN powder

Pentaerythritol tetranitrate (PETN) is a widely studied high explosive (HE), most commonly used in detonator applications. In this work, we use a spray drying technique to manufacture a new “nano-powder” morphology of PETN. The spray dried PETN is several orders of magnitude smaller in particle size than traditionally prepared PETN powders. The spray dried PETN has a mostly spherical and smooth morphology, while traditional crystalline PETN typically has sharp edges and crystal facets. Small-scale sensitivity tests including drop-weight impact, friction, and electrostatic discharge (ESD) indicate the spray dried PETN is less sensitive than traditionally used forms of PETN powders. Furthermore, we also observed no changes in chemical properties (melt temperature and onset of decomposition) after spray drying, and the material remains in the tetragonal, crystalline phase.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

The Explosive Hazards of Icosahedral Carboranes and Dodecaborate with Common Oxidizers and Salts

Carboranes find use in a wide variety of applications ranging from specialty catalysts to medicinal uses. Certain functionalized carboranes are described as explosive materials, although their nonfunctionalized precursors are regarded as inherently stable and not explosive. Herein, we report new findings on the sensitization of nonfunctionalized 12-vertex ortho-, meta-, and para-carboranes as well as dodecaborate in physical mixtures with several common inorganic salts. These mixtures were evaluated for explosive properties with drop hammer impact tests, friction tests, and differential scanning calorimetry thermal gravimetric analysis (DSC-TGA). Results indicate explosive sensitization to mechanical stimuli for the nonfunctionalized carboranes tested with a wide variety of inorganic salts, especially oxidizers. Sensitization of these compounds has not previously been reported, and raising awareness to the broader scientific community on this matter is warranted, especially as several of the mixtures evaluated are more sensitive to mechanical stimuli than primary explosives such as lead azide or pentaerythritol tetranitrate.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Can a shock-induced phonon up-pumping model relate to impact sensitivity of molecular crystals, polymorphs and cocrystals?

Impact sensitivity engineering of high-energy molecular crystals requires accurate predictive models. For this purpose, the promising multi-phonon based approach is selected, assessing a bit more its strengths and weaknesses. Presently used with high-quality phonon calculations of 22 molecular crystals, using a physics-based criterion to determine the phonon bath extent, the resulting intrinsic shock sensitivity index (SSI) is compared to the most common marker of impact sensitivity, h 50 , as determined from drop-weight impact tests. Selecting a data subset from experiments performed under very similar conditions (2.5 kg hammer with grit and 30–40 mg samples), the model can predict h 50 values for mono-molecular crystals with very good accuracy, including the ability to discriminate the polymorphs of HMX and CL20. This very good agreement validates an initial indirect up-pumping mechanism occurring under these conditions, where the doorway modes also interact with the phonon bath. However, the phonon bath criterion for mono-molecular crystals does not transfer well to cocrystals. Owing to the vibrational coupling of the co-molecules, it seems a broader phonon bath should be considered. Additionally recalling experimental uncertainty and various experimental factors affecting h 50 values for a given compounds, we recommend that the density of the sample, granularity and morphology be systematically considered and reported along with measurements, which will in turn allow for more systematic data and predictive capabilities for sensitivity models.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Development and modeling for a small-scale, rapidly heated high explosives initiation time (HEIT) experiment

Most small-scale assessments of explosive sensitivity, including the popular drop-weight impact test, convolute thermal and mechanical phenomena to the extent that it has been extremely challenging to decipher how an explosive ignites and propagates reactions. For instance, an impact generates heat through a range of dissipation mechanisms, which can in turn, depending on the reaction rates of the explosive, lead to chemical decomposition. To deconvolute the various contributions to the sub-shock initiation and propagation of explosive reactions, we describe the development and modeling of the High Explosives Initiation Time (HEIT) test - a new, small-scale, high-throughput experiment designed to rapidly heat milligram quantities of energetic materials confined within small diameter steel needles. Specifically, we have modeled and designed a 250 joule pulsed power system capable of rapidly delivering electrical current to the needles, resulting in rapid heat delivery to the sample. In conclusion, the energy deposition rate into the sample is controlled by different transmission line topologies. Modeling in COMSOL is performed to understand the energy required to heat up the explosive sample.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2020

Multichannel Deconvolution of Vibrational Shock Signals: An Inverse Filtering Approach

When transporting critical systems of national security interest, out-of-the ordinary, impulsive events that can potentially be undetected and affect overall system performance are of great concern. Impulsive events that can occur are essentially pulse-like, transient signals of short duration that evolve from various phenomena. Here the event can be created by either the dropping of a test object, the system, subjecting it to a compact high-energy blow or being struck unintentionally during transit resulting in potential damage. The intensity and location of the strike can cause an inoperability condition that is unacceptable in a national security environment. Therefore, it is essential to detect, classify and localize damage of any test object subjected to an impulsive-event. This effort was targeted to evaluate the vibrational response of test objects that are subjected to “transport” shocks and roadway vibrations during shipping and handling. Any potential damage that could be inflicted during transportation must not only be detected, but also be evaluated to determine the operational readiness of a test object before and after transport. This event is a critical task that must be addressed as part of the Lawrence Livermore National Laboratory (LLNL) national security mission. The estimation of excitation signals from noisy data is termed the deconvolution problem in the signal processing literature. The deconvolution problem is based on recovering the input excitation signal from a system characterized by its impulse response sequence. Using this model of the system, an “inverse” representation or filter is developed to remove the system from the measured data and recover the input. Deconvolution techniques have existed for a long-time; however, transient deconvolution presents a few uncommon problems, since the signal has a finite-length time duration resulting in a limited amount of data containing information about the excitation process. The transient is wideband in the frequency domain relative to any measurement sensor implying that the smaller bandwidth sensor system “filters” the excitation eliminating some of its essential information for recovery. This fact, coupled with the filtering effect of the test object itself makes this excitation recovery (deconvolution) problem a challenge for signal processing.

42 ENGINEERING↗
DOE OSTI2020

Improvements to CTF Closure Models for Modelling of Two-Phase Flow

This report documents the efforts to improve the CTF prediction of void fraction and two-phase pressure drop by improving two-phase closure models. Previous validation activities have revealed that CTF tends to overpredict void fraction and two-phase pressure drop. In response to this, two approaches were taken to improve CTF’s predictive capabilities. Based on findings that the interfacial drag and subcooled boiling models significantly impact void prediction, alternative closure models for these physical effects were found and implemented into the code. An extensive assessment was performed by using the existing and newly added two-phase experimental data, which showed that void prediction, wall temperature, and two-phase pressure-drop results are improved by using the newly implemented models. A second approach for improving CTF modeling accuracy involved the use of a Bayesian calibration process that uses CTF validation data to optimize selected modeling coefficients and closure model multipliers to achieve a more accurate prediction of experimental results. A similar assessment was performed with calibrated models that improved the void and pressure-drop prediction for test cases that were both included and not included in the calibration dataset. Results of this study will be used to change the models used in CTF to achieve more accurate BWR analyses moving forward. This study also identified opportunities for improving additional closure models and other opportunities to use calibration techniques to improve CTF.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Drop Analysis of a Department of Energy Standard Canister Containing Fort Saint Vrain SNF

DOE manages over 300 types of SNF, many of which are located at the INL site. Managing this large variety of SNF for storage, transportation, and disposal poses a challenge to DOE. The Idaho Cleanup Project and INL are collaborating on the DOE SNF Road-Ready Demonstration (“Road-Ready Demonstration”), which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF for “road-ready dry storage.” Road-ready dry storage (RRDS) is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in on-site storage in anticipation of later transport and disposition. The forward-looking goal of the Road-Ready Demonstration is to establish the foundation for a large-scale RRDS program at the INL site. One critical aspect of RRDS is the ability to certify the DOE Standard Canister and its associated transportation package in accordance with 10 CFR 71 for offsite transportation. Depending on the SNF type and transportation strategy, DOE Standard Canisters may be required to maintain structural integrity under hypothetical accident scenarios (e.g., drop events). The DOE Standard Canisters have been tested and analyzed under various SNF loading configurations and accident drop events in support of the Idaho Spent Fuel Facility and other DOE programs; however, no analysis has yet been completed in support of the recently initiated Road-Ready Demonstration. This paper presents preliminary results from a finite element analysis of the Ø45.7 cm × 4.6 m (Ø18 in. × 15 ft) DOE Standard Canister under the 9 m drop at 80 degrees off-vertical drop scenario considered in previous INL tests and analyses. It considers the Fort St. Vrain spent nuclear fuel loading configuration proposed for the Road-Ready Demonstration, uses updated material properties, and applies the strain-based acceptance criteria established in ASME Boiler and Pressure Vessel Code’s Section III, Division 3 rules for storage and transportation spent nuclear fuel containments. This updated analysis is compared to previous DOE Standard Canister drop analyses. Preliminary results from the updated analysis show that certain regions of the containment exceed the allowable limits during the accidental drop event. However, these regions are limited to components performing a non-structural function. While further work on this analysis will be pursued, this analysis serves as the foundation for formal calculations used to support applicable certification efforts of the RRDS system at INL.

42 ENGINEERING↗