SEARCH · Engineering Papers

Results for “irradiation effect”

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.

SourceResearch connectionFrom yearThrough year

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

At least 289 records · Page 16

All sources NASA NTRS DOE OSTI

DOE OSTI2020

AGC-4 Experiment Irradiation Monitoring Data Qualification Final Report

The Graphite Technology Development Program ran a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. This report focuses on the fourth experiment, Advanced Graphite Creep 4 (AGC 4). The Advanced Reactor Development (ARD) Technology Development Office (TDO) Program for research and development activities require documentation of qualified monitoring data to design and license the first high-temperature reactor nuclear plant. Qualified data meets the requirements for use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements and provide no useable information. Trend data may not meet all requirements but still provide some useable information. Use of Trend data requires assessment of how any deficiencies affect a particular use of the data. AGC-4 began with Advanced Test Reactor (ATR) Cycle 157D on May 30, 2015. After irradiating the graphite for two cycles, the capsule was removed from the reactor after ATR Cycle 158A, which ended on January 2, 2016, due to interference with another experiment. Irradiation was resumed with Cycle 162A on October 7, 2017 after the interfering experiment was removed from the reactor. AGC-4 irradiation completed by the end of Cycle 166B on January 10, 2020. Between Cycles 162A and 166B, AGC-4 capsule was removed from the reactor core during two Powered Axial Locator Mechanism (PALM) Cycles 163A and 165A to avoid overheating. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. A total of 9,256,184 out of all 11,167,213 temperature records (or 82.9%) are Qualified for use by the ARD Program and 1,911,029 missing values are Failed. Argon, helium, and total gas flow data were within expected ranges, except only eight out-of-range values occurred during outages. A total of 21,240,627 out of all 22,288,5103 gas flow rates (or 92.8%) are Qualified for use by the ARD Program and 1,644,476 are Failed records mostly due to missing values. Discharge gas line moisture values were consistently low during full ATR power, except for 162B, when moisture content was increasing to more than 200 ppmv by the end of this cycle. During outages, moisture content reached as high as 700 ppmv. 1,089,941 out of a total of 1,107,331 moisture values (or 98.4%) are within the measurement range of the instrument and are Qualified for use by the ARD Program and 17,390 missing moisture values are Failed. Graphite creep specimens were subjected to one of three loads: 393, 491, or 589 lbf. For a brief period during Cycle 157D between 12:19 on June 2, 2015, and 08:23 on June 11, 2015, the load cells were wired incorrectly, resulting in missing stack load data. Missing stack loads were estimated from measured ram pressures using regression equations developed from the existing data from Cycle 157D. Estimated stack loads during this period are considered to be an accurate representation of actual load applied to the stacks. These loads deviate slightly from the planned loads. This deviation does not prevent the data from being Qualified for use but must be taken into account when analyzing the effect of load on creep. 6,095,728 out of a total of 6,403,985 moisture values (or 95.2%) are within the measurement range of the instrument and are Qualified for use by the ARD Program and 308,257 missing load values are Failed. Stack displacement increased consistently throughout the eight cycles, with total displacement reached highest value of 2.4 in by the end of irradiation. During ATR outages, a set of pneumatic rams raised the stacks of graphite creep specimens to ensure the specimens were not stuck within the test train. This stack raising was performed seven times throughout irradiation. All stacks were raised successfully each time. 4,744,974 out of a total of 6,094,513 displacement values (or 77.9%) are within the measurement range of the instrument and are Qualified for use by the ARD Program and 1,349,539 displacement values are Failed mostly due to missing values. Analyses were conducted on correlations between TCs to look for trends and step changes that might indicate instrument degradation or failure. Correlation analysis was used to identify instances when TCs form short circuits, referred to as virtual junctions, which result in TCs reporting temperatures from some location in the capsule other than the location where they were intended to read. No evidence of virtual junctions was found. Analyses were also conducted on control charts of temperature differences of between two TCs, which are expected to behave consistently throughout the entire irradiation period. Upward or downward trend over time indicates at least one TC in the pair was drifted. Examining control charts for all 66 possible pairs out of twelve TCs installed in the AGC-4 capsules reveal no clear drift failures occurred, except unstable behavior of two TCs in Zone 3, TC-7 and TC-8, over irradiation time. In conclus

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

A mechanism-based quantitative multi-scale framework for investigating irradiation hardening of tungsten at low temperature

We present the development and validation of a mechanism-based multi-scale modeling framework to quantitatively link the irradiation defect evolution kinetics at the microscopic scale to the macroscopic yield strength and flow stress evolution of tungsten irradiated at low temperature (0.08 < T/T m < 0.18, where T m is the melting point and is equal to 3673 K). The mechanism-based strength model, proposed as the superposition of thermal softening and modified dispersed barrier hardening, is developed to understand the underlying strengthening mechanisms. The thermal softening exponent as defined in the Johnson-Cook model is obtained by fitting unirradiated yield strength at different temperatures. A set of irradiation-induced defect kinetics equations with the thermodynamics parameters derived from the atomic calculations reported in the literature is used to determine the densities and sizes of defect clusters at the meso-scale, and the predicted irradiation damage characteristics are in reasonable quantitative agreement with experimental data from literature. The effects of irradiation condition (temperature and irradiation dose) and test temperature on yield strength are quantitatively predicted and compared with experimental measurements from the literature. Here, the predicted irradiation hardening and strain hardening are compared with experimental data as model validation.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Radiation-Induced Changes in Single Crystal Calcite and Dolomite: Mineral Analogues of Light Water Reactor, Nuclear Power Plant Concrete Aggregates

Two analogues of aggregates (calcite and dolomite) found in nuclear power plant (NPP) concrete have been neutron irradiated to fluences up to 2 × 10 20 n/cm 2 at 52 °C and 4 × 10 19 n/cm 2 at 95 °C (E > 0.1 MeV). X-ray diffraction studies showed that these aging conditions affected crystallography of both calcite and dolomite, with both showing a growth in their unit cells compared to that of the unirradiated samples. Only dolomite samples showed a continuous increase of its lattice parameters with the increase in irradiation. In calcites, the lattice growth with irradiation dissipated after a fluence of 5 × 10 18 n/cm 2 . Radiation-induced disordering of the crystal structure was observed in both materials, while a healing effect of irradiation temperature was mainly observed in calcites. Dislocation densities calculated by using XRD data also supported these observations. Insights made into the unit cells of both minerals by using Fourier synthesis showed that the Mg sites are more prone to the neutron irradiation than the Ca sites, resulting in larger radiation effects and higher disruption in the crystal structure of dolomite than that of calcite. Furthermore, other than the polycrystalline characteristics observed by using TEM, these minerals did not show complete amorphization or significant structural decomposition at the studied irradiation fluences.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Isotope-Specific Analysis of Neutron-Irradiated Lithium Aluminate Ceramics by Nuclear Magnetic Resonance Spectroscopy

The effects of neutron irradiation on lithium aluminate targets and the permeation of tritium into surrounding components are primary concerns in the development of materials for use in tritium production. The high specific activity and volatility of the tritium in activated samples greatly increase the challenges of experimental analysis. This paper reports the determination of the chemical forms and quantities of H-3 and Li-6 in irradiated and non-irradiated LiAlO2 by nuclear magnetic resonance (NMR) spectroscopy, and demonstrates that measurements can be performed in a non-destructive, safe manner to obtain isotope-selective, quantitative information on highly hazardous materials.

Tritium, solid state nuclear magnetic resonance sp↗

DOE OSTI2021

Effects of HZ-PAN Sorbent Irradiation on Kr Adsorption

Idaho National Laboratory (INL) has developed and studied an engineered form hydrogen mordenite in polyacrylonitrile matrix (HZ-PAN) for krypton (Kr) adsorption. Adsorption studies to date have successfully separated and captured Kr from carrier gas streams, but those studies have used non-radioactive gases for testing. However, if utilized in a used nuclear fuel (UNF) reprocessing facility, this sorbent will be exposed to the potential deleterious effects of radiation. This report explores Kr adsorption behavior on HZ PAN after the sorbent was exposed to gamma radiation. A batch of HZ-PAN was divided into five segments, a control segment that was not exposed to radiation, and the other four segments that were exposed to progressively increased levels of radiation dose, up to 5 MGy. Samples were then analyzed to determine total surface area as well as their capacity to capture Kr. This study reveals HZ PAN retains its initial surface area and Kr-capture capacity after radiation exposure, indicating viability for service in radioactive gases.

42 - ENGINEERING↗

DOE OSTI2022

Stress Profile in Coating Layers of TRISO Fuel Particles in Contact with One Another

This work presents a discussion on a series of finite element analyses that assess stress evolution in the coating layers of tristructural isotropic (TRISO) particles in contact with each other while embedded in a matrix. The initial simulations were of applied uniaxial pressure versus matrix elastic modulus. These simulations predicted increasing stress in the silicon carbide coating layers of the TRISO particles with decreasing matrix elastic modulus. The second set of simulations focused on the effects of heating and cooling and the associated dimensional change on the state of stress in the coating layers. The general finding was that there was no significant difference below the coating layer’s deposition temperature. Although, above the deposition temperature, the contacting particles had higher stress compared with those that were separated. The third set of simulations focused on the effects of irradiation, specifically, creep, dimensional change, and swelling. An interface debonding model was introduced since these potential effects have a significant bearing on predicted stresses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Effect of Hydrogen and Helium on Irradiation Damage Accumulation in Nickel

We report our results on the study of the effects of hydrogen and helium on irradiation damage accumulation in nickel, aiming to understand the TEM observations made by Edwards et al. regarding the Ni-coatings protecting the getter tube in a TPBAR. We hypothesized that differences in the fluxes of 3 H and/or He between the coatings might explain these microstructural variations. However, our simulation results show that during the initial stages of irradiation, the presence of H or He enhances void nucleation by stabilizing small vacancy clusters against thermal dissociation and does not enhance void growth. Nevertheless, we think that helium’s higher diffusivity, stronger binding with vacancy clusters, and ability to displace a nearby lattice atom—either to create a new vacancy or to enlarge an existing vacancy cluster—could result in a larger average vacancy cluster size. Therefore, it is likely that void growth will occur post-nucleation phase as more He accumulates at higher doses than those simulated in this study.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Helium effects on the surface and subsurface evolutions in single-crystalline tungsten

Tungsten (W) has been perceived as one of the most promising plasma facing materials (PFMs) for future fusion reactors. In the past decade, its behavior under irradiation and helium (He) plasma interaction has been extensively studied. However, some key knowledge gaps still exist, such as the influence of crystallographic orientation on the surface and subsurface evolutions. In this work, we focus on the He ion-beam irradiation damage effects in mirror-polished single-crystalline W samples with three different surface planes of {100}, {110} and {111}. Irradiation was performed at room temperature using 40 keV He+ to a fluence of 1 × 10 16 /cm 2 , followed by thermal desorption spectroscopy (TDS) up to ~1920 K. The microstructures of He-irradiated W before and after TDS heat treatment were characterized by scanning and transmission electron microscopy. Subsurface He bubbles were imaged in all irradiated samples, but newly formed <111>-oriented surface grains and surface blisters were only observed in W {100} and {110} starting orientations. These results reveal that radiation damage, He thermal desorption, and surface/subsurface evolution are all strongly dependent on crystallographic orientation. Underlying physical mechanisms are discussed based on ion channeling effects, He-vacancy interactions, and surface diffusion. In conclusion, these findings provide new insights into He effects in W.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Modeling and simulation of oxygen transport in high burnup LWR fuel

We have developed a formulation for oxygen transport in uranium dioxide nuclear fuel that accounts for the effects of irradiation. The overall simulation combines the evolving isotopic composition, thermochemistry, and oxygen transport in irradiated fuel. The driving forces for oxygen transport are computed from local thermodynamic equilibrium calculations and account for the effects of temperature gradients and composition, including fission products. The proposed method provides a mechanism for including complex thermodynamic models of nuclear fuel in modeling of mass redistribution, and alleviates difficulties associated with the common thermodiffusion formulation. Here, the transport model has been implemented within the nuclear fuel performance code BISON utilizing the thermochemistry code Thermochimica, with burnup calculations provided by the ORIGEN isotopic transmutation code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Spectral Effects in Albedo and Rearside Irradiance Measurment for Bifacial Performance Estimation: Preprint

Albedo data are essential for accurate prediction of bifacial PV module performance. However, spectral response mismatch between PV modules and irradiance sensors used in albedometers can limit the accuracy of performance predictions. In order to provide quantitative assessment of this effect, we investigate via simulation the differences in spectrally responsive albedo measured with thermopile pyranometers and crystallinesilicon PV reference cells in comparison to a representative crystalline-silicon bifacial PV module for nine different representative ground surface materials. Calculations are performed using simulated solar spectra together with catalogued spectral reflectivity data distributed with the SMARTS simulation software. For the specific materials considered, the results show that albedo measurement using thermopile pyranometers could over or under-estimate the ground-reflected radiation usable by a bifacial PV module by up to 10%, versus only approximately 4% total range of variation for a PV reference cell.

Albedo↗

DOE OSTI2021

Review of Graphite Properties Relevant to Micro-Reactor Design

Microreactors are advanced reactors that are designed to be small, scalable, and portable with up to 20 MW power. These reactors could operate independently, or as part of a microgrid, without need of a frequent refueling. Micro-reactors are designed to be deployable to provide electric power in remote locations, therefore providing greater flexibility. Several micro-reactors are under development in the US. These reactors are based on advanced gas-cooled or heat-pipe concepts. While Graphite can be found in various applications, some high-temperature reactors plan to employ Graphite in their core components due to its role as a moderator and its ability to maintain its structural integrity when subjected to elevated temperatures and high neutron fluence. Although there is significant experience with use of Graphite in nuclear applications, there are still unknowns that should be addressed. Graphite properties depend on the graphite type and affected by the fabrication procedures. Material properties are functions of the temperature and irradiation dose. The fundamental properties also affected by the oxidation environment. Therefore, in-depth understanding of graphite properties, including the effect of irradiation, oxidation and other influencing parameters are required to evaluate the design accurately. This draft aims to provide a comprehensive review of material properties for Graphite relevant to micro reactor design. For this work, two distinct grades of graphite were considered, namely IG-110 and Mersen-2114. Physical, thermal, and mechanical properties for these grades were comparatively evaluated. The properties that are discussed in this draft aim to serve as a basis for design and safety analysis.

42 ENGINEERING↗

DOE OSTI2023

Compatibility of ethylene vinyl acetate (EVA)/ethylene vinyl alcohol (EVOH)/EVA films with gamma, electron-beam, and X-ray irradiation

Many polymer-based medical devices are sterilized by gamma irradiation. To reduce the use of cobalt-60 gamma-ray sources, transition from gamma ray to alternative irradiation technologies was proposed, namely electron beam (e-beam) and X-ray. A major impediment for such a transition is the knowledge gap in material compatibility with the different radiation sources. In this study, multi-layer films consisting of ethylene vinyl acetate (EVA) and ethylene vinyl alcohol (EVOH) components were irradiated to target doses of 30, 45, and 60 kGy by gamma-ray, e-beam, and X-ray sources. Effects of irradiation were evaluated on 12 material properties, and statistical comparisons between gamma irradiation and alternative technologies were conducted using the two one-sided t-test (or “equivalence test”) and classic t-test. Melting temperature and UV absorbance below 300 nm showed dose dependencies, while other investigated properties such as discoloration and mechanical durability did not change with dose up to 60 kGy. Based on these results, there is no material compatibility issue associated with the transition from gamma to e-beam or to X-ray as source of sterilization radiation of the studied multi-layer film.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

The effects of high energy deuteron ion beam irradiation on the tensile behavior of HT-9

Ion beam implantations are widely performed to understand the effects of irradiation-induced displacement damage on nuclear structural materials. However, the volume of material that can be investigated is often limited by the maximum energies of accelerator facilitates, leading to a limitation in the thickness of samples whose mechanical properties can be evaluated. The Lawrence Berkeley National Laboratory’s 88-Inch Cyclotron offers a wide range of ions and energies, allowing for material ion implantations at larger scales than typical. Four HT-9 SS-J-geometry tensile specimens were polished and then irradiated with deuterons at the 88-Inch Cyclotron to doses of approximately 0.2 dpa prior to small scale tensile testing. Here, the results from this study show irradiation hardening characterized by the tensile test results and black dot irradiation defects. Additionally, a comprehensive look at low temperature irradiations of high-Cr F/M steels is presented and our results show agreement with the available data.

36 MATERIALS SCIENCE↗

DOE OSTI2026

HDG-1 Experiment Irradiation Monitoring Data Qualification Final Report

SUMMARY The U.S. Department of Energy (DOE) Advanced Reactor Technologies (ART) Graphite Research and Development (GRD) Program is conducting a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. This report documents the qualification of irradiation monitoring data for the fifth experiment, High Dose Graphite-1 (HDG-1). Qualified monitoring data are required by the ART program to support the design and licensing of the first high-temperature reactor (HTR) nuclear plant. Data are classified as Qualified if they meet the usage requirements described in the experiment planning and quality assurance (QA) documents, Failed if they do not meet those requirements and provide no usable information, or Trend if they do not fully meet all requirements but still provide useful information subject to an assessment of how any deficiencies may affect a particular use of the data. HDG-1 irradiation began with Advanced Test Reactor (ATR) Cycle 168B on August 24, 2020, and concluded after Cycle 173C on January 27, 2025. The HDG-1 capsule was removed from the reactor core twice—during core internal change (CIC) Cycle 170A and powered axial locator mechanism (PALM) Cycle 172A—to prevent overheating of the graphite specimens during high-power PALM cycles. The capsule was therefore irradiated during a total of seven normal ATR cycles: 168B, 169A, 171A, 171B, 173A, 173B, and 173C. Irradiation monitoring data evaluated in this report include thermocouple (TC) temperature, gas flow rate, gas moisture, gas pressure, specimen load, and graphite stack displacement. Temperature. A total of 14,508,065 TC temperature records were captured. Of these, 13,901,785 (95.8%) are Qualified and 606,280 (4.2%) are Failed. The principal source of failed temperature data was the instrument failure of TC-9 (Zone 2) on June 24, 2024, and TC-10 (Zone 1) on July 5, 2024, near the end of Cycle 173A, which resulted in 595,554 Failed readings. An additional 379 missing values and 10,347 slightly negative values from TC-13 during ATR outages are also Failed. Neither TC-9 nor TC-10 was used as a temperature-control TC, and their failures did not compromise capsule condition monitoring. Correlation analysis of all 13 TCs found no evidence of virtual junction formation. Control chart analysis revealed clear downward drift of approximately 80°C for TC-6 (Zone 3) relative to other stable TCs, and possible downward drift of approximately 60°C for TC-13 relative to the Zone 5 control TC (TC-1), though TC-13 remained consistent with the Zone 2 control TC (TC-12). Gas flow. A total of 20,088,090 gas flow rate records were captured. Of these, 19,941,463 (99.3%) are Qualified and 146,627 (0.7%) are Failed due to missing values. All argon, helium, and total gas flow data were within expected ranges throughout the irradiation. Gas moisture. A total of 1,116,005 outlet gas moisture values were captured. Of these, 1,101,421 (98.7%) are Qualified and 14,584 (1.3%) are Failed, comprising 14,556 out-of-range values and 28 missing values. The out-of-range moisture values exceeded 22,000 ppmv for approximately 1 week at the beginning of Cycle 173A, when accumulated moisture evaporated after the capsule was retrieved from water storage during PALM Cycle 172A and reinserted into the east flux trap. Moisture levels returned to below 25 ppmv for the remaining three cycles, and the transient high-moisture event did not affect the integrity of specimen irradiation. Gas pressure. A total of 7,812,035 gas pressure values were captured. Of these, 6,642,048 (85.0%) are Qualified and 1,169,987 (15.0%) outlet pressure values are Failed, comprising 718,537 zero outlet pressure values due to sensor failure from Cycle 168B through Cycle 171B, 54,550 missing values, and 396,900 too-low outlet pressure values, ranging from 1.1 to 1.6 psia after sensor replacement during Cycle 173A. Load. A total of 6,696,030 load values were captured. Of these, 6,694,580 (99.98%) are Qualified and 1,450 (0.02%) are Failed due to missing values. Applied loads to the six specimen stacks were stable throughout the irradiation. Stack displacement. A total of 6,696,030 displacement values were captured. Of these, 5,713,297 (85.32%) are Qualified and 3,781 (0.06%) are Failed due to missing values. Stack displacement increased consistently throughout the irradiation, reaching approximately 3.08 in. for Channels 5 and 6 by the end of irradiation. 978,952 (14.62%) substantially elevated displacements observed for Channel 6 beginning in Cycle 171A and for Channel 5 beginning in Cycle 173A are assigned Trend status. Raising pressure. A total of 1,115,999 raising pressure values were captured. Of these, 1,115,430 (99.95%) are Qualified and 569 (0.05%) are Failed due to missing values. Ram pressure. A total of 6,696,030 ram pressure values were captured. Of these, 6,692,249 (99.95%) are Qualified and 3,484 (0.05%) are Failed due to missing values. Stack raising was perf

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Impact of post-irradiation annealing on mechanical performance of irradiated 718 alloy

Here, the effect of post-irradiation annealing on solution-annealed 718 alloy was investigated using advanced mechanical testing, fractography, scanning electron microscopy, and transmission electron microscopy. Specimens were extracted from a proton beam window operated at the Spallation Neutron Source, irradiated with 940 MeV protons to a maximum dose of approximately 9.7 displacements per atom (dpa) at a calculated temperature not exceeding 110 °C while in service. Helium and hydrogen concentrations reached about 1700 and 6900 atomic parts per million (appm), respectively. Despite irradiation and high tensile strength (yield stress over 1 GPa), the material exhibited significant ductility. Annealing at 500 °C, 700 °C, and 900 °C for 30 min resulted in an appreciable decrease in yield strength and an increase in ductility for annealing treatments at 500 °C and 900 °C relative to the strength and ductility of the as-irradiated material. The presence of helium and hydrogen led to cavity formation and cleavage-like brittle features on fractured surfaces; however, high-magnification imaging revealed the presence of small-scale ductile dimples, indicating that the fracture mechanism remained mixed. The annealed specimens retained total elongation levels of 14–33 %, and there was no sudden drop in ductility after heat treatments. The ductility level in the irradiated and annealed material is notable despite the presence of helium and hydrogen.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Yttrium Hydride Post Irradiation Examination Plan

The overarching goal of the Department of Energy Office of Nuclear Energy (DOE-NE) microreactor program is to develop technologies for the deployment of civilian microreactors by stakeholders1. Microreactors are expressed as advanced transportable nuclear reactors operating at low power (<20MWth) but high temperatures (>600°C), as well as plug-and-play and inherently safe designs. One prerequisite of a microreactor is the compactness, so that a truck can transport the reactor under safe conditions with the current road infrastructure1,2. The compactness of these reactors likely can be attainable by use of solid components for the essentials of the nuclear core, such as fuel, core heat removal components, reflectors, and moderators. Among these essentials, where fuel enrichment must remain < 20% to meet High Assay Low Enrichment Uranium criteria (HALEU), the largest contribution to the compactness is offered by use of solid moderators which benefit from light atomic weight elements, such as hydrogen, carbon, and beryllium2. Among these, hydrogen-bearing materials, such as metal hydrides, are superior to other options from the lowest critical mass standpoint. Noting that, factors other than critical mass should be considered for a specific reactor design. Yttrium- or zirconium-based metal hydrides have been down-selected due to their neutronic performance. In addition to the neutronic perspective, maintaining hydrogen within the metal hydride is important at the high operating temperatures proposed by advanced reactors. Yttrium hydride (YHx) is, therefore, a proposed moderator material that offers better hydrogen retention at higher operating temperatures than zirconium hydrides due to higher retention and thermal stability of hydrogen in the metal3. The irradiated materials properties of metal hydrides, in this case YHx, must be assessed for the qualification of these moderators. Material testing and inspection processes must illustrate that the effects of dimensional and property changes on thermophysical and mechanical properties do not cause any significant changes on the microreactor safety, and the moderating power is maintained within design limits. Thus, the effect of irradiation on the thermophysical and mechanical properties must be determined. This post-irradiation examination (PIE) plan specifically aims to determine these properties for YHx following Advanced Test Reactor (ATR) irradiation.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2022

PLN-6268 Yttrium Hydride Post Irradiation Examination Plan

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2022

University of Missouri Research Reactor (MURR) LEU Fuel Fluid-Structure Interaction Analysis

The University of Missouri Research Reactor (MURR ® ) is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, that are actively collaborating with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M 3 ) Reactor Conversion Program to convert from highly enriched uranium (HEU, ≥ 20 wt.% U-235) to low-enriched uranium (LEU, < 20 wt.% U-235) fuel. A new type of LEU fuel with very high density, based on an alloy of uranium and 10 weight percent molybdenum (U-10Mo), is expected to allow the conversion to LEU of USHPRR that have been found unable to be converted with previously qualified uranium silicide-aluminum (U 3 Si 2 -Al) dispersion fuel. MURR has been working with the USHPRR Reactor Conversion (RC) Pillar at Argonne National Laboratory to perform fuel element design and fuel cycle performance analyses, steady-state thermal hydraulics safety analyses, and accident safety analyses in preparation for the conversion of MURR and to support a preliminary Safety Analysis Report (SAR) for conversion to LEU fuel. In this work, Fluid-structure interaction (FSI) analysis at the fuel element level, as compared to the fuel plate level of the previous work. is performed which models all components of the MURR LEU fuel element, including fuel plates and the supporting structures e.g., side plates, end fittings, and combs. Therefore, the effect of supporting structures on the coolant flow distribution, the fuel plate deflection, and the resulting coolant channel gap reduction can be evaluated. In addition to the nominal element geometry and flow rate, the tolerances in the geometry dimensions of coolant channel and plate thickness, the effect of a comb on plate deflection, and the uncertainty of the flow rate per element are considered in this work. The effect of comb on plate displacement is quantified through two bounding cases: the case assuming a perfect bond between the comb and plates and the case neglecting the comb effect. Note that in this analysis, the FSI has been decoupled from the other structural effects caused by irradiation (e.g., swelling and irradiation creep). Assessment of combined effects is planned for a later stage of this project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗