Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “transient behavior”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 469 records · Page 26

In-situ X-ray and visual observation of foam morphology and behavior at the batch-melt interface during melting of simulated waste glass

We report to attain a basic understanding of the primary foam structure and behavior, which affects the heat and mass transfer and the efficiency of the glass melting process, we investigated the primary foam layer under the glass batch floating on molten glass. The primary foam affects mass transfer during batch melting, in turn affecting the melting process. The recently performed direct in-situ three-dimensional X-ray computed tomography of the batch melting in a laboratory-scale melter vessel allowed us to visualize the features of the reacting batch layer and the foam that develops at its bottom, though with an insufficient resolution of images. In this study, we obtained better temporal and spatial resolution using the two-dimensional X-ray radiography and visual observation of the structure and behavior of transient primary foam as it formed and decayed. As soon as the batch was charged onto the melt surface, foam bubbles began to evolve, grow, and coalesce, forming a primary foam layer, 5-10 mm thick, within tens of seconds. This foam layer was sustained by ongoing gas evolving reactions counterbalanced by bubble coalescence into cavities that moved sideways and escaped to the atmosphere. Eventually, the entire remaining batch turned into foam that gradually decayed at the melt surface. The decay rate agreed with literature observations of surface foam produced by secondary foaming.

36 MATERIALS SCIENCE↗

Insight Into the High-Potential Branch of the Alternate Nad+-Dependent NADPH:Ferredoxin Oxidoreductase II (NfnII) from Pyrococcus Furiosus. Evidence for the Gating Step in Electron Bifurcation

We have investigated the rapid-reaction kinetics of the NAD+-dependent NADPH:ferredoxin oxidoreductase II (NfnII) from Pyrococcus furiosus, permitting a comparison with recent work done with the paralog NfnI from the same organism. The half-potentials of the electron-bifurcating L-FAD are highly crossed in both NfnI, meaning the potential for the quinone/semiquinone couple is significantly lower than that for the semiquinone/hydroquinone couple so that the semiquinone oxidation state is thermodynamically unstable. The same appears to be the case with NfnII on the basis of its similar behavior in transient absorption spectroscopy experiments and the absence of any evidence for FAD*- accumulation in the course of reductive titrations (which would be manifested as a transient increase in absorbance at ~380 nm)1. Reductive titrations with both the one-electron donor sodium dithionite and the obligate two-electron donor NADPH demonstrate that, in contrast to NfnI, little FADH* accumulates in the high-potential pathway of NfnII in the course of reduction, as reflected in the absence of a transient increase in absorbance in the 500-600 nm region. This indicates that the half-potentials of the S-FAD are crossed in NfnII by a minimum of 120 mV. Rapid-reaction experiments mixing oxidized NfnII with NADPH also show no evidence of S-FADH* accumulation. Furthermore, the enzyme is only partially reduced at the end of the reaction with NADPH, indicating that there is little electron transfer into the high-potential pathway of NfnII. When the reaction is carried out in the presence of the one-electron carrier ferredoxin, only approximately one equivalent of ferredoxin becomes reduced, in contrast to the maximum of three equivalents seen with NfnI. This is consistent with only a single electron bifurcation event taking place under these conditions with NfnII, with only a single high-potential electron passing into the [2Fe-2S] cluster of the high-potential pathway of NfnII but not progressing further to the S-FAD. This accounts for the observation that NfnII has greatly reduced bifurcating activity compared to NfnI. The lack of electron transfer into the S-FAD of NfnII due to its crossed half-potentials prevents successive bifurcation activity and indicates that electron transfer within the high-potential branch gates bifurcation activity in NfnI.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Effect of reactor radiation on the thermal conductivity of TREAT fuel

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Metallic fuel transient fuel-cladding interface liquefaction model assessment platform enabled by integrating BISON with databases

A novel platform has been developed within the BISON fuel performance code to assess models of fuel-cladding interface liquefaction for sodium-cooled fast reactor (SFR) metallic fuels. Here, this platform is crucial because liquefaction at the fuel-cladding interface significantly impacts fuel performance and may compromise fuel pin integrity during transient events. To ensure accurate predictions, the platform integrates data collected during the Integral Fast Reactor (IFR) program, now archived in metallic fuel databases. This integration supports verification and validation (V&V) of the models in BISON. Leveraging the extensive US experience with metallic fuel liquefaction and the collections of preserved legacy data, the platform serves as a powerful tool for evaluating existing models and advancing the development of new ones.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Validating the BISON fuel performance code to integral LWR experiments

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Li-Metal Anode in Dilute Electrolyte LiFSI/TMP: Electrochemical Stability Using Ab Initio Molecular Dynamics

Ab initio molecular dynamics simulations were performed for Li + conducting electrolytes based on trimethyl phosphates (TMP) and lithium bis(fluorosulfonyl)imide (Li + FSI – ) salt in contact with a Li-metal electrode. We focused on the transient-state behavior at the electrolyte, interfacial electrolyte–Li-metal electrode, and lithium reference electrode–electrolyte–Li-metal electrode to study dynamics and activation energy barriers of the Li + ion, electrochemical and thermal stability of the interface electrode-electrolyte, and potential behavior of the Li-metal electrode, respectively. Our results show that in the most stable state, Li + ions are tetrahedrally coordinated to three TMPs and one FSI – . The inner solvation shell of a Li-ion is composed of three TMP and one FSI – in one contact ion-pair and four TMPs in a solvent-separated ion-pair. On the other hand, Li-ions transport through electrolyte cages takes place when they are coordinated with three or less molecules that could be a combination of TMPs and FSI – . The decomposition pathway of the LiFSI salt when in direct contact with the Li-metal anode starts with defluorination of FSI – , rapidly losing F– to the lithium surface, forming LiF species. The remaining FSO 2 NSO 2 –2 with the addition of 2e – from the Li-metal decomposes into SO 2 –2 and NFSO 2 –2 . SO 2 –2 deposits on the Li-surface and decomposes into Li 2 O and Li 2 S. The remaining NFSO 2 –2 defluorinates, losing F – ion to the lithium surface, resulting in LiF and the remaining NSO 2 –1 deposits on the lithium surface and decomposes in the following picoseconds, forming several binary compounds such as Li 3 N, Li 2 S and Li 2 O. In contrast, when the salt is solvated by the TMP molecules, avoiding a direct contact with the Li metal electrode, only one defluorination occurs, decomposing the FSI – into FSO 2 NSO 2 –2 and a F – . The two anions remain stable as they are solvated by the TMP molecules. Furthermore, we also analyzed the open circuit potential energy (OCPE) of the Li-metal electrode during the SEI formation. OCPE is calculated from the average local potential profile difference within the Li-metal electrode and a pristine Li-crystal reference electrode (LRE). When no SEI is formed, the Li-metal electrode has an average OCPE of +0.36 eV vs LRE. Due to the formation of a SEI, the Li-metal electrode has an average OCPE between -0.07 and -0.21 eV vs LRE. The OCPE of the Li-metal electrode decreases by ~0.42 eV when a SEI is formed.

25 ENERGY STORAGE↗

Pathways Following Electron Injection: Medium Effects and Cross-Surface Electron Transfer in a Ruthenium-Based, Chromophore–Catalyst Assembly on TiO 2

Interfacial dynamics following photoexcitation of the water oxidation assembly [((PO 3 H 2 ) 2 bpy) 2 Ru II (bpy-bimpy)Ru II (tpy)(OH 2 )] 4+ , -[Ru a II –Ru b II –OH 2 ] 4+ , on nanocrystalline TiO 2 electrodes, starting from either -[Ru a II –Ru b II –OH 2 ] 4+ or -[Ru a II –Ru b III –OH 2 ] 5+ , have been investigated. Transient absorption measurements for TiO 2 –[Ru a II –Ru b II –OH 2 ] 4+ in 0.1 M HPF 6 or neat trifluoroethanol reveal that electron injection occurs with high efficiency but that hole transfer to the catalyst, which occurs on the electrochemical time scale, is inhibited by local environmental effects. Back electron transfer occurs to the oxidized chromophore on the microsecond time scale. Photoexcitation of the once-oxidized assembly, TiO 2 –[Ru a II –Ru b III –OH 2 ] 5+ , in a variety of media, generates -[Ru a III –Ru b III –OH 2 ] 6+ . The injected electron randomly migrates through the surface oxide structure reducing an unreacted -[Ru a II –Ru b III –OH 2 ] 5+ assembly to -[Ru a II –Ru b II –OH 2 ] 4+ . In a parallel reaction, -[Ru a III –Ru b III –OH 2 ] 6+ formed by electron injection undergoes proton loss giving -[Ru a II –Ru b IV $=$O] 4+ with possible conversion to -[Ru a II –Ru b II –OH 2 ] 4+ by an electrolyte-mediated reaction. In the following slow step, re-equilibration on the surface occurs either by reaction with added Fe III/II or by cross-surface electron transfer between spatially separated -[Ru a II –Ru b IV $=$O] 4+ and -[Ru a II –Ru b II –OH 2 ] 4+ assemblies to give -[Ru a II –Ru b III –OH 2 ] 5+ with a half-time of t 1/2 ~ 68 μs. Finally, these results and analyses show that the transient surface behavior of the assembly and cross-surface reactions play important roles in producing and storing redox equivalents on the surface that are used for water oxidation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Contactless Determination of Electrical Conductivity of One-Dimensional Nanomaterials by Solution-Based Electro-orientation Spectroscopy

For nanowires of the same composition, and even fabricated within the same batch, often exhibit electrical conductivities that can vary by orders of magnitude. Unfortunately, existing electrical characterization methods are time-consuming, making the statistical survey of highly variable samples essentially impractical. Here, we demonstrate a contactless, solution-based method to efficiently measure the electrical conductivity of 1D nanomaterials based on their transient alignment behavior in ac electric fields of different frequencies. In comparison with direct transport measurements by probe-based scanning tunneling microscopy shows that electro-orientation spectroscopy can quantitatively measure nanowire conductivity over a 5-order-of-magnitude range, 10–5–1 Ω–1 m–1 (corresponding to resistivities in the range 102–107 Ω·cm). With this method, we statistically characterize the conductivity of a variety of nanowires and find significant variability in silicon nanowires grown by metal-assisted chemical etching from the same wafer. We also find that the active carrier concentration of n-type silicon nanowires is greatly reduced by surface traps and that surface passivation increases the effective conductivity by an order of magnitude. Moreover, this simple method makes electrical characterization of insulating and semiconducting 1D nanomaterials far more efficient and accessible to more researchers than current approaches. Electro-orientation spectroscopy also has the potential to be integrated with other solution-based methods for the high-throughput sorting and manipulation of 1D nanomaterials for postgrowth device assembly.

36 MATERIALS SCIENCE↗

Time-dependent SOLPS-ITER simulations of the tokamak plasma boundary for model predictive control using SINDy *

Abstract Time-dependent SOLPS-ITER simulations have been used to identify reduced models with the sparse identification of nonlinear dynamics (SINDy) method and develop model-predictive control of the boundary plasma state using main ion gas puff actuation. A series of gas actuation sequences are input into SOLPS-ITER to produce a dynamic response in upstream and divertor plasma quantities. The SINDy method is applied to identify reduced linear and nonlinear models for the electron density at the outboard midplane n e , s e p O M P and the electron temperature at the outer divertor T e , s e p d i v . Note that T e , s e p d i v is not necessarily the peak value of T e along the divertor. The identified reduced models are interpretable by construction (i.e. not black box), and have the form of coupled ordinary differential equations. Despite significant noise in T e , s e p d i v , the reduced models can be used to predict the response over a range of actuation levels to a maximum deviation of 0.5% in n e , s e p O M P and 5%–10% in T e , s e p d i v for the cases considered. Model retraining using time history data triggered by a preset error threshold is also demonstrated. A model predictive control strategy for nonlinear models is developed and used to perform feedback control of a SOLPS-ITER simulation to produce a setpoint trajectory in n e , s e p O M P using the integrated plasma simulator framework. The developed techniques are general and can be applied to time-dependent data from other boundary simulations or experimental data. Ongoing work is extending the approach to model identification and control for divertor detachment, which will present transient nonlinear behavior from impurity seeding, including realistic latency and synthetic diagnostic signals derived from the full SOLPS-ITER output.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Imaging coherent phonons and precursor dynamics in LaFeAsO with 4D ultrafast electron microscopy

In this work, we used 4D ultrafast electron microscopy (UEM) to directly image femtosecond photoinduced structural dynamics in single-crystal LaFeAsO at initial temperatures of 300 and 100 K, above and below the known structural and magnetic phase-transition temperatures, respectively. With nanometer-picosecond resolution, we resolved an initial (precursor) sigmoidlike response arising from photothermal expansion and lattice reorientation that precedes the onset of propagating coherent acoustic phonons (CAPs). In the specific regions probed, the precursor response at 100 K is shorter than at 300 K ($t$ 0.5;100 K = 11.3ps vs $t$ 0.5;300 K = 17.8ps), and the CAP oscillation frequency is lowered with cooling (ƒ CAP;100 K = 12 GHz vs ƒ CAP;300 K = 21 GHz), correlated to known lattice softening due to the structural phase change. The transient CAP behaviors at 300 K are dispersive, displaying an exponentially decaying phase velocity over the first nanosecond. Further, the CAP symmetry at 300 K matches a first-order antisymmetric shear mode (A 1 ), while at 100 K it is best matched by a mostly nondispersive zero-order symmetric mode (S 0 ). These findings illustrate the sensitivity of UEM imaging to spatially heterogeneous dynamics in the Fe-pnictide materials and more broadly in other quantum materials.

36 MATERIALS SCIENCE↗

Aviation Mission Profile Performance Evaluation in a High Specific Power Surface Permanent Magnet Machine with Additively Manufactured Coils and Integrated Heat Pipes

The purpose of this paper is to accurately evaluate the performance and overall efficiency of an electrical machine driving a propeller (as a part of a hybrid or electric propulsion system) during the flight mission profile using electromagnetic-thermal simulations. The study focuses on a 250 kW, 5000 rpm, 1.25 kHz, 36/30 slot/pole surface permanent magnet (SPM) machine with additively manufactured (AM) hollow conductors and integrated heat pipes (HP). The machine drive parameters (frequency, and current) are tuned to maximize the efficiency for a given flight cycle, while providing the highest specific power (SP) during takeoff. Afterward using Multiphysics simulations, the transient thermal behavior of the machine is studied and temperature profile for different parts of the machine is extracted. Finally, steady-state temperatures during cruising are estimated and overall efficiency is re-evaluated.

additive manufacturing↗

Denudation, solute export, landscape evolution modeling, and geographic information system data for the East River watershed, Colorado, USA (2020-2024)

This data package contains geographic information system (GIS) layers and tabular datasets associated with the study of lithologic controls on denudation, solute export, carbon-scaling relationships, and transient landscape evolution in the East River watershed near Crested Butte, Colorado, USA. The package includes GIS layers used to produce the Figure 2 map, including drainage, hillshade, lithology, sample locations, and basin polygons, together with comma-separated value (CSV) tables and matching CSV data dictionaries. One group of tables reports sample-level and catchment-level information for river-sediment samples analyzed for in situ-produced cosmogenic beryllium-10 (10Be), including sample names, outlet elevations, geographic coordinates, upstream drainage area, rock-type classes, production-rate scaling scheme, analyzed nuclide, catchment-averaged denudation rates, and associated lower and upper analytical uncertainties. Sample and catchment attributes provide the basis for comparing denudation rates across intrusive, shale, sedimentary, and mixed-lithology settings. A second group of tables reports supporting information for landscape-evolution modeling and the mapped geologic framework of the study area. Included files list parameter values and definitions for the two-phase landscape-evolution simulations, summarize full-domain model erosion fluxes and topographic metrics for different simulation configurations, provide a fixed-area carbon-model scaling table, and summarize mapped geologic units within the East River study domain, including geologic code, formation name, lithologic description, mapped area, and lithologic class grouping. Model outputs and geologic summaries support interpretation of transient landscape behavior and its relation to the mapped distribution of shale, intrusive, sedimentary, and surficial units. A third group of tables reports hydrologic and hydrochemical information used to quantify dissolved export from the watershed. Included files provide site-level values for drainage area, mean annual solute export, standard error of annual export, area-normalized solute yield, and equivalent weathering rate for five East River monitoring sites, along with metadata describing the number, sampling cadence, and date range of discharge records and partial and full total dissolved solids observations used in the solute-yield analyses. The package also contains a supplementary daily ion-load time series with daily mean discharge, discharge observation counts, dissolved concentrations, and daily loads for calcium, magnesium, sodium, potassium, chloride, sulfate, nitrate, fluoride, dissolved silica, charge-balance bicarbonate, and total dissolved solids. The package contains GIS files, comma-separated value files (.csv), CSV data dictionaries, a file-level metadata table, a package-tree text file, and a readme text file.

10Be↗

High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Integration of Human Reliability Analysis Models into the Simulation-Based Framework for the Risk-Informed Safety Margin Characterization Toolkit

This report presents an application of a computation-based human reliability analysis (HRA) framework called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER). HUNTER has been developed not as a standalone HRA method but rather as framework that ties together different HRA methods to model dynamic risk of human activities as part of an overall probabilistic risk assessment (PRA). While we have adopted particular methods to build an initial model, the HUNTER framework is meant to be intrinsically flexible to new pieces that achieve particular modeling goals. In the present report, the HUNTER implementation has the following goals: • Integration with a high fidelity thermal-hydraulic model capable of modeling nuclear power plant behaviors and transients • Consideration of a PRA context • Incorporation of a solid psychological basis for operator performance • Demonstration of a functional dynamic model of a plant upset condition and appropriate operator response This report outlines these efforts and presents the case study of a station blackout scenario to demonstrate the various modules developed to date under the HUNTER research umbrella.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Integration of a Microturbine Power Conversion Unit in MAGNET

The Microreactor Agile Non-nuclear Experimental Testbed (MAGNET) facility is being constructed at Idaho National Laboratory’s (INL’s) Energy Systems Laboratory to assist with the development, demonstration, and validation of microreactor components and systems. MAGNET will accommodate various types of test articles representative of different microreactor designs. A power conversion unit (PCU) simulator loop was proposed to use the rejected heat from the test article and mimic the thermal hydraulic conditions (e.g., temperature, pressure and flow rate) that would be experienced if an actual Brayton cycle PCU was attached. However, due to the high-estimated cost of the PCU simulator and the availability of a turbogenerator used for previous testing at Sandia National Laboratories, the PCU simulator design was abandoned in favor of integrating an actual PCU into MAGNET. A commercial Capstone C30 was modified by Sandia to use external electrical heating, rather than combustion heating, to provide a power output from 1 to ~30 kWe. The C30 has a single-stage centrifugal combustor, a static radial heat exchanger, and a radial inflow turbine arranged on a single shaft with a high-speed alternator. Since the operating temperature and energy transfer per unit working fluid mass are comparable, the MAGNET PCU will provide researchers with the ability to evaluate the test article heat transfer under representative operating conditions with the transient system behavior associated with a closed Brayton cycle PCU. This report examines the integration of a modified commercial Capstone C30 turbogenerator into MAGNET, outlines a proposed set of tests, and describes future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

High Dose and Dose Rate 60 Co γ-Irradiation of High-Density Polyethylene

We have irradiated high-density polyethylene ribbons and cylinders to ~200 kGy at a dose rate of ~5 kGy day -1 , using a 60 Co source. The radiolytically-generated gas consists almost entirely of H 2 . Based on the time- and dose-dependence of the outgassing, we have determined the radiation- chemical yield for the evolved H 2 , as well as the diffusion coefficient of H 2 in HDPE. We find that the evolved gas yield G evolved depends on the interplay of sample geometry, experimental timescale, and the diffusion coefficient of H 2 in HDPE. When these factors facilitate completion of diffusion prior to gas quantification, such as for sub-mm-thickness ribbons, we measure a time- and dose-invariant value of G evolved = 0.0076 cm 3 std. H 2 J -1 that is in agreement with literature reports. When sample geometry and experiment timescale do not permit completion of diffusion, we measure a time- and dose-dependent value of G evolved . We have developed a model for the transient outgassing behavior of the cylinders, and fitted it to the experimental data. Based on these fits, we estimate a room-temperature diffusivity of D H 2 /HDPE ≈ 1.67 ± 0.12 x 10 -6 cm 2 s -1 for H 2 in HDPE, in good agreement with literature values. At present, the effects of dose rate and γ-photon energy on HDPE gas yield are not known and require further investigation.

36 MATERIALS SCIENCE↗

FY23 Post-LOCA Characterization of Irradiated Fuel (NSUF Award 17-12985)

This report summarizes the sample preparation and data collection on fuel pins following integral LOCA testing in the Severe Accident Test Station at Oak Ridge National Laboratory. The test sample cross-sections show post-transient fragmentation behavior of UO 2 commercially irradiated above 67 MWd/kgU. Ring compression testing and hydrogen measurements were collected on the cladding segments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 was a protected loss of flow test where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R was an unprotected loss of flow test where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for SHRT-17, SHRT-45R, BOP-301, and BOP-302R there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗