Results for “Examination” — Page 11

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 OSTI2023

Post irradiation examination of a uranium-zirconium hydride TRIGA fuel element

Low-enriched (LEU) U-ZrH fuel, with a 235 U content less than 20% of the total uranium, is being evaluated for possible use in different types of reactors, including space nuclear systems, light water reactors (LWRs) and micro-reactors. As a result, it is beneficial to better understand the macrostructural and microstructural changes that occur in this fuel during irradiation. This paper reports the results of the post irradiation examination of an LEU U-ZrH fuel element (30 wt.% U, <20% 235 U) using neutron radiography, precision gamma scanning, chemical analysis, optical metallography and scanning electron microscopy combined with energy dispersive spectroscopy and wavelength dispersive spectroscopy, where the fuel element was irradiated in a Training, Research, Isotope, General Atomics (TRIGA) reactor. Results of microstructural characterization indicated some dehydriding and cracking of the U-ZrH fuel occurred during irradiation; an axial and radial burnup gradient existed in the fuel during irradiation, as measured by gamma scanning and chemical analysis; negligible microstructural changes transpired during irradiation, based on comparison of irradiated and as-fabricated U-ZrH fuel microstructures; and, negligible, fission product-rich, phases could be resolved in a U-ZrH fuel that was irradiated to a calculated 20% depletion of 235 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRSDate not supplied

Preliminary Examination Process of Apollo Core 73002 - Insights and Lessons Learned From ANGSA for Future Sample Return Missions

Apollo Sample 73002 is part of a 2-foot long “drive tube” (73001/73002) of regolith that was collected from a landslide deposit near Lara Crater at the Apollo 17 site, Station 3. The double drive tube is believed to have penetrated a lunar landslide deposit that was transported from the slope of the South Massif into the TLV [1]. As part of the ANGSA (Apollo Next Generation Sample Analyses) initiative, preparing preliminary examination (PE) catalog of 73002 is a crucial first step for the early identification of material types such as rock fragments, and potential stratigraphy within the core. PE of Apollo core 73002 is distinct from science activities with the main goal to produces a sample catalog with a level of detail about sample characterization that is sufficient for the ANGSA PIs (and later on the lunar sample community) to select and request the samples to conduct their individual, scientific studies. Ultimately, the PE catalog of 73002 will help to establish a better understanding of the stratigraphy of the land slide deposit; the processes of the landslide including the trigger(s) and possibly number of landslide events, as well as the role of volatiles [1] and will aid in the careful preservation of the material for future studies [2].

Apollo↗

DOE OSTI2020

AGR 3/4 Compact 7 4 Examination Plan

This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 7 4 taken from the Advanced Gas Reactor (AGR) experiment, AGR 3/4. This work will be performed in accordance with the general objectives outlined in the AGR 3/4 PIE Plan and guidance in the ORNL PIE Statement of Work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

AGR 3/4 Compact 8 4 Examination Plan

This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 8 4 taken from the Advanced Gas Reactor (AGR) experiment, AGR 3/4. This work will be performed in accordance with the general objectives outlined in the AGR 3/4 PIE Plan and guidance in the ORNL PIE Statement of Work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Yttrium Hydride Post-Irradiation Examination Plan

Micro nuclear reactors are expected to use metal hydride moderators to ensure thier compact design for the flexible operational purposes. Department of Energy’s Microreactor program therefore aims to provide infrasture and qualification bed for the development of micro nuclear reactors. Because micro nuclear reactors must be compact to offer a flexible operational purposes, metal hydrides are expected to be used as neutron moderator materials due to their high hydrogen number density. Los Alamos National Laboratory (LANL) directed by the DOE to provide a qualified fabrication techniques of metals hydride, especially the yttrium hydride (YHx) which provides advantages over other metal hydride candiates. Qualification of YHx involves fabrication, property, and performance pillars, which must be met for the deployment of microreactors. In the support of the qualification process metal hydride, LANL has developed fabrication techniques based in direct hydriding and powder compacting. For both manufacturing techniques, fresh material properties of metal hydrides have been reported in the literature. To establish the performance limits and the qualification parameters, an irradiation campaign has been initated at Idaho National Laboratory (INL). The irradation campaign aims to investigate the critical irradiated YHx properties as a function of neutron fluence and temperaute. The first phase of the irradiation campaign includes the effect of the operating temperature under neutron irradiation which provides initial data on the metal hydrides and informs the manufacturing processes. So that, drop-in irradations in Advanced Test Reactor (ATR) has been designed. This document depicts the post-irradation examination (PIE) plan of the ATR irradiated YHx. Objective of the PIE plan is to serve as a qualification basis for YHx as a neutron moderator material by understanding the impacts irradiation has on the physical and thermophysical properties; to establish a link between fabrication and the irradiated YHx properties; to institute a PIE strategy for other solid-state neutron moderator candidates, and; to inform micro nuclear reactor fuel performance and safety codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

AGR-5/6/7 Post-Irradiation Examination Plan

This plan describes the PIE activities to be carried out on AGR-5/6/7 fuel and the non-fuel components of the AGR-5/6/7 irradiation test train. PIE of AGR-5/6/7 fuel and test train components will provide data on fuel performance over a range of irradiation conditions (e.g., burnup, neutron fluence, and temperature), test fuel under postulated accident conditions, and support development of fuel performance and fission product transport models. Examination of the non-fuel test train components will give information on TRISO fuel and compact performance and information on releases of radioactive fission products from the fuel compacts. PIE of the fuel compacts is used for many purposes including, but not limited to: updating the irradiation thermal calculations, measuring burnup, determining retention/release of fission products from the compacts and the TRISO particles, determining TRISO layer failure rates, determining fuel performance at elevated accident temperatures, determining fuel performance under oxidizing atmospheres, determining the distribution of fission products within the compact and fuel particles, and observing compact and particle morphology to explain mechanisms of TRISO layer failure and/or fission product release/retention.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

AGR-2 TRISO Fuel Post-Irradiation Examination Final Report

An array of post-irradiation exams and experiments were employed to assess the irradiation performance and high-temperature behavior of AGR-2 UCO TRISO fuel produced in an engineering-scale coater. Work was also conducted on UO2 TRISO fuel produced for comparison with the AGR UCO fuel. Emphases we replaced on enumerating the frequency of SiC layer failures (failure of the SiC layer with at least one PyC layer remaining intact) and TRISO failures (failure of all three dense layers) and evaluating the retention of key fission products within the fuel. Examinations of the fuel kernel and coating morphologies were used to understand the features typical of normal intact fuel, degraded or failed fuel, and mechanisms of degradation. Radiochemical analyses were used to determine their retention within the fuel. The results confirm excellent performance of the fuel that exceeds historical HTGR design requirements with significant margin.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

PLN-6268 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

Irradiated AGR 5/6/7 Compact 4-1-3 Examination Plan

This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 4-1-3 taken from the Advanced Gas Reactor (AGR) experiment, AGR 5/6/7. This work will be performed in accordance with the general objectives outlined in the AGR 5/6/7 PIE Plan and guidance in the ORNL PIE Statement of Work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Non-Destructive Postirradiation Examination of the AFC-4C Capsule/Rodlet 5 Experiment

The AFC-4C irradiation test was irradiated at the Idaho National Laboratory Advanced Test Reactor to investigate several different fast reactor fuels that could be used to facilitate ultra-high burnup applications in sodium fast reactors. Several different alloys, fuel geometries, bonding materials, and coating/barrier were tested in ferritic-martensitic HT-9 cladding. The AFC-4C capsule 5 was removed from the reactor at around 8.7 at %HM burnup to begin postirradiation examination. This report presents and discusses the non-destructive PIE results of the AFC-4C-capsule / rodlet 5, a U-10Zr solid fuel, sodium bonded with HT9 with internal Cr-coating, and its performance is evaluated against the historical fuel performance of previously irradiated fuel from literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Irradiated AGR 5/6/7 Compact 5-6-2 Examination Plan

This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 5-6-2 taken from the Advanced Gas Reactor (AGR) experiment, AGR 5/6/7. This work will be performed in accordance with the general objectives outlined in the AGR 5/6/7 PIE Plan and guidance in the ORNL PIE Statement of Work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Post Irradiation Examination Results of Irradiated Yttrium Hydride

Department of Energy’s (DOE’s) Microreactor program (MRP) aims to provide the fundamental data to enable the development of microreactors. As such, material property data of critical materials for microreactor technologies are researched. Because substoichiometric yttrium dihydride (YHx, where x<2) is considered as a potential solid neutron moderator, its material property data has been combined in the Advanced Moderator Material Handbook which includes thermodynamic and thermophysical properties of YHx with the exception of irradiated material’s properties due to limited PIE. To fill the knowledge gap for the irradiated YHx, specimens and irradiation capsules were prepared at Los Alamos National Laboratory (LANL). Specimens were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory’s (INL’s). post-irradiation examination (PIE) was performed at INL’s Materials and Fuels Complex (MFC). This report compiles the PIE results of irradiated YHx specimens through fiscal years 2022 and 2023 (FY22-23) . The PIE data will directly be incorporated into the newer version of the Advanced Moderator Material Handbook. The main takeaways include that (i) the geometrical stability and mechanical integrity of YHx was intact with couple exceptions after high-temperature irradiations (600-800°C), (ii) hydrogen content variation due to manufacturing or irradiation in YHx caused visible surface discoloration, that is also related to the microstructural changes, (iii) qualitative comparisons of PIE methods implied that H retention was significantly higher at 600°C as compared to 800°C, (iv) thermal properties included signatures correlated with the H loss or re-gain, (v) importance of manufacturing readiness and initial as-manufactured specimens history was emphasized, (vi) the needs of targeted irradiations focusing on temperature and time parameters and very targeted PIE were specified.

08 HYDROGEN↗

DOE OSTI2024

Resumption of SFR Overpower Testing and Post-Transient Examination of the THOR-C-2 Irradiation Test

Transient overpower testing for fuel safety research and development has resumed at Idaho National Laboratory. The THOR-C-2 commissioning test has been performed and post-transient examination has been performed revealing the intended creep rupture failure near the top of the fuel zone. This paves the way for future testing on previously irradiated fuels that will support fuel development and qualification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Report on the Incorporation of Post-Irradiation Examination Results of Yttrium Hydride for Advanced Moderator Handbook

Department of Energy’s (DOE’s) Microreactor program (MRP) aims to provide the fundamental data to enable the development of microreactors. As such, material property data of critical materials for microreactor technologies are researched. Because substoichiometric yttrium dihydride (YHx, where x<2) is considered as a potential solid neutron moderator, its material property data has been combined in the Advanced Moderator Material Handbook which includes thermodynamic and thermophysical properties of YHx with the exception of irradiated material’s properties due to limited PIE. To fill the knowledge gap for the irradiated YHx, specimens and irradiation capsules were prepared at Los Alamos National Laboratory (LANL). Specimens were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory’s (INL’s). post-irradiation examination (PIE) was performed at INL’s Materials and Fuels Complex (MFC). This report compiles the PIE results of irradiated YHx specimens through fiscal years 2022 and 2023 (FY22-23) . The PIE data will directly be incorporated into the newer version of the Advanced Moderator Material Handbook. The main takeaways include that (i) the geometrical stability and mechanical integrity of YHx was intact with couple exceptions after high-temperature irradiations (600-800°C), (ii) hydrogen content variation due to manufacturing or irradiation in YHx caused visible surface discoloration, that is also related to the microstructural changes, (iii) qualitative comparisons of PIE methods implied that H retention was significantly higher at 600°C as compared to 800°C, (iv) thermal properties included signatures correlated with the H loss or re-gain, (v) importance of manufacturing readiness and initial as-manufactured specimens history was emphasized, (vi) the needs of targeted irradiations focusing on temperature and time parameters and very targeted PIE were specified.

08 HYDROGEN↗

DOE OSTI2024

Resumption of SFR Overpower Testing and Post-Transient Examination of the THOR-C-2 Irradiation Test

Transient overpower testing for fuel safety research and development has resumed at Idaho National Laboratory. The THOR-C-2 commissioning test has been performed and post-transient examination has been performed revealing the intended failure near the top of the fuel zone. This paves the way for future testing on previously irradiated fuels that will support fuel development and qualification.

advanced reactors↗

DOE OSTI2025

What's on the menu? Examining native apex‐ and invasive meso‐predator diets to understand impacts on ecosystems

Abstract Understanding how carnivores impact ecological communities is essential for guiding effective management actions and conserving biodiversity. Quantifying predators' diets, including prey selectivity, allows for the assessment of the relative effects native and invasive predators may have on prey populations. In Australia, populations of a native, terrestrial apex predator, the dingo Canis dingo/C. familiaris , and introduced and invasive subordinate mesopredators, the European red fox Vulpes vulpes and feral cat Felis catus , co‐occur, but there is limited understanding of their relative impacts on native and invasive prey in different ecosystems. To assess the possible effects of dingoes, foxes and cats on prey, we examined their diet and prey selectivity across a ~10,000 km 2 semi‐arid mallee ecosystem. Using macroscopic scat analysis, we identified strong dietary niche separation. Larger‐bodied dingoes primarily consumed large marsupial herbivores, whereas foxes and cats primarily consumed smaller prey, including introduced and native rodents and birds. Foxes had the broadest diet, and the greatest dietary overlap with cats ( = 0.81), compared with dingoes ( = 0.50) or between dingoes and cats ( = 0.36). Livestock were identified in 2% of dingo and 7% of fox scats. Cats and foxes consumed more than 15 times the volume of small native mammals compared with dingoes, including threatened species such as fat‐tailed dunnarts Sminthopsis crassicaudata . Cats and foxes also selectively consumed small mammals relative to their estimated availability and consumed fewer large mammals. In contrast, dingoes consumed fewer birds and more echidnas relative to their availability. Our results suggest limited intraguild competition within this semi‐arid ecosystem, as dingoes are primarily exerting top‐down pressure on large herbivores, whereas invasive mesopredators are disproportionately impacting smaller prey, including threatened native mammals. Practical implication . Our findings suggest that ongoing conservation management of dingoes, red foxes and feral cats must consider the variation in diets, impacts on prey and ecological roles of these different predator species, and avoid indiscriminate lethal control methods. Quantifying actual, rather than assumed, impacts of predators on threatened native species, large herbivores and livestock is essential to achieve effective and integrated ecosystem management.

Mason, Rachel T. [School of Life and Environmental↗

DOE OSTI2022

Examining the peatland shrubification-evapotranspiration feedback following multi-decadal water table manipulation

Northern peatlands are globally important long-term sinks of carbon due to their predominantly saturated conditions. However, these ecosystems are expected to become drier with climate change, potentially leading to shrubification. As such, the response of the shrubification–evapotranspiration (ET) feedback may be of critical importance to future peatland energy, water and carbon dynamics. Here we examined the effect of multi-decadal peatland water table (WT) alteration at three adjacent sites with increasing depth to WT (WET, INTermediate, and DRY). In order to better understand the WT–shrubification–ET feedback, we measured peatland vegetation composition, microtopography and ET partitioning, where ET was measured at the ecosystem, microform, and leaf level using eddy covariance (EC), chambers and porometry, respectively. Averaged across microforms and WT treatments, there was a difference in the median measured leaf resistance (r leaf ) between plant functional types ranging from 213 s m -1 for erect dwarf shrubs, 325 s m -1 for graminoids/sedges, and 520 s m -1 for prostrate dwarf shrubs. Scaled based on LAI, the low r leaf of erect dwarf shrubs dominated hummocks, where sites with a higher proportion of hummocks had lower median canopy resistance (r v ) of 141, 133 and 130 s m -1 at the WET, INT and DRY sites respectively. Nevertheless, ET was highest at the WET site and similar between the INT and DRY sites, with greater evaporation from the moss surface at the WET site. Porometry and EC data along with a three-source model were used to independently assess the evaporative contribution from the moss surface, which ranged from 17% to 40%. For moderate and persistent changes in WT from land-use or climate change, our results suggest vegetation succession is minimal, but the microtopographic development and the concomitant differences in LAI for the various plant functional types is key to understanding changes in total ET and partitioning.

54 ENVIRONMENTAL SCIENCES↗