Testing Accident Tolerant Fuels in Reactivity Initiated Accident Conditions -- PIE Results
Initial PIE results of the SETH series of transient irradiation experiments performed in TREAT.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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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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X501 Experiment Simulation and PIE Data Comparison
Fuel performance simulations have been developed for X501 MA bearing fuel in BISON. The results of these simulations are compared with PIE results to assess the applicability of metallic fuel models in BISON to MA bearing metallic fuel.
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PIE-Enabled Study of Aqueous Corrosion & Zr Hydriding in Cr-Coated Cladding: M3GV-23PN0101132
Chromium-coated zirconium alloy cladding is under investigation as an accident-tolerant fuel (ATF) concept to extend performance to higher burnups via improved oxidation resistance and reduced hydrogen pickup. However, hydrothermal corrosion behavior and hydrogen transport mechanisms governing in-service hydriding of these coatings remain poorly understood. In a collaborative study between Pacific Northwest National Laboratory (PNNL), Idaho National Laboratory (INL), and the University of Huddersfield, cold spray (CS) and physical vapor deposition (PVD) Cr-coated Optimized ZIRLO™ cladding samples were characterized after exposure to PWR-simulated water chemistry under both in-core (neutron irradiation) and out-of-core (aqueous-only) conditions at the MIT Research Reactor. Multi-scale characterization was performed independently at PNNL and INL to evaluate coating integrity, microstructural evolution, Cr/Zr interface chemistry, and hydride formation. Both laboratories observed a consistent divergence in hydriding behavior: out-of-core CS Cr-coated cladding exhibited elevated hydride concentrations exceeding those of uncoated cladding, whereas in-core CS samples showed substantially suppressed hydriding. In-core specimens also displayed irradiation-specific features, including nanoscale voids within the Cr coating and radiation-induced segregation clusters in the Zr substrate. CS coatings retained an interdiffusion-free Cr/Zr bond with no intermetallic layer, whereas PVD coatings exhibited inferior quality with visible cracks and pores. An automated image-based hydride quantification method systematically overestimated bulk hydrogen content relative to inert gas fusion measurements, underscoring the need for standardized sample preparation and reference standards. Interpretation of hydride nucleation and growth is complicated by the open inner-diameter of the specimens, which provides additional hydrogen pathways. The results motivate further foundational studies to support predictive models of hydrothermal corrosion and hydrogen transport in unirradiated and irradiated Cr-coated cladding.
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Accelerated fission rate irradiation design, pre-irradiation characterization, and adaptation of conventional PIE methods for U-10Mo and U-17Mo
Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 10 20 and 1.5 × 10 21 fissions per cubic centimeter. Swelling measurement of the specimen sizes studied here required development and assessment of new methods for volume determination before and after irradiation. Laser profilometry and X-ray computation tomography (XCT) were used to provide preirradiation characterization of samples to determine the error and applicability of each to determine swelling following irradiation. These outcomes are contextualized through use of BISON simulations performed to assess the predicted expansion of U-Mo fuels subjected to the irradiation conditions of this work. Use of existing BISON fuel performance models predicted a maximum of 7% swelling under the irradiation conditions of this study. Pre-irradiation characterization revealed the as-cast U-Mo fuel samples were uniformly large-grained fully cubic U crystals with small U-C/N bearing precipitates and pores distributed throughout. Samples were found to contain a bulk porosity between .4 and 3% because of the casting process. Local porosity in areas far from large, interconnected pores was found by Slice-and-View to be under .2%. Nanometer-sized precipitates rich in C and N were identified in all samples, likely because of impurities during the fabrication process. Dendritic bands were also observed throughout the samples. These bands were characterized by variable Mo content that deviated from the overall Mo content by 2–3 wt %. No other microstructural features were correlated to these bands. Mechanical properties were found to be slightly strengthened compared to literature reports of bulk U-Mo fuels due to the nano-scale precipitates throughout the sample.
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AGR-5/6/7 PIE Overview 2024
This is a status update and description of major post-irradiation examination activities under the DOE's AGR TRISO fuel program. This presentation will be given at the annual program review hosted by INL the week of July 15.
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LaSalle Leaking Pin PIE Proposal
Determine the most probable root cause of the LaSalle fuel rod failure, distinguish primary failure mechanisms from post-failure degradation, and assess whether the condition is isolated or indicative of a broader issue affecting similar rods or assemblies. Results will support decisions related to core operation, fuel management, and corrective actions.
LaSalle↗
Accident Tolerant and High Burnup Hotcell PIE at ORNL
Explore the source record for details and available documents.
Olsen, Luke↗
Irradiation and PIE of alloys X-750 and XM-19 (EPRI Phase III)
The Nuclear Science User Facilities (NSUF) based at Idaho National Laboratory (INL), along with the Electric Power Research Institute (EPRI), formed an agreement to test representative alloys used as reactor structural materials as a pilot program to establish guidelines for future NSUF research programs. This report contains results from the portion of this program established as Phase III (of three phases), entailing irradiation and post-irradiation examination of select alloys typical of boiling water reactor (BWR) internal structural materials. Phases I and II are the subject of separate reports and represent baseline material test results and irradiation experiment design, respectively. The intent of this Phase III research program is to determine properties for the materials of interest after being irradiated at the Advanced Test Reactor (ATR) to three different target fast (E>1MeV) fluences: 5.0 x 10¹⁹ n/cm², 2.0 x 10²⁰ n/cm², and 1.0 x 10²¹ n/cm². These correspond to irradiation damage levels (displacements per atom [dpa]) of approximately 0.08, 0.30, and 1.4 dpa, which represent comparable levels to (a) a previous study which looked at X-750 irradiated to ~1 x 10¹⁹ n/cm², comparable to the lowest fluence; (b) approximately a medium level of fluence for BWR components; and (c) extended life (60 – 80 years) for BWR components. The materials chosen for this research are the nickel-based alloy X-750 and austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co., and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research Center (GE-GRC), and parts were provided to INL for use in this pilot project.
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Irradiation and PIE of alloys X-750 and XM-19 (EPRI Phase III)
The Nuclear Science User Facilities (NSUF) based at Idaho National Laboratory (INL), along with the Electric Power Research Institute (EPRI), formed an agreement to test representative alloys used as reactor structural materials as a pilot program to establish guidelines for future NSUF research programs. This report contains results from the portion of this program established as Phase III (of three phases), entailing irradiation and post-irradiation examination of select alloys typical of boiling water reactor (BWR) internal structural materials. Phases I and II are the subject of separate reports and represent baseline material test results and irradiation experiment design, respectively. The intent of this Phase III research program is to determine properties for the materials of interest after being irradiated at the Advanced Test Reactor (ATR) to three different target fast (E>1MeV) fluences: 5.0 x 10¹⁹ n/cm², 2.0 x 10²⁰ n/cm², and 1.0 x 10²¹ n/cm². These correspond to irradiation damage levels (displacements per atom [dpa]) of approximately 0.08, 0.30, and 1.4 dpa, which represent comparable levels to (a) a previous study which looked at X-750 irradiated to ~1 x 10¹⁹ n/cm², comparable to the lowest fluence; (b) approximately a medium level of fluence for BWR components; and (c) extended life (60 – 80 years) for BWR components. The materials chosen for this research are the nickel-based alloy X-750 and austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co., and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research Center (GE-GRC), and parts were provided to INL for use in this pilot project.
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Status Report on Characterization of High Burnup Fuel with Advanced Nondestructive Pulsed Neutron PIE
Characterizing irradiated or spent nuclear fuels with pulsed neutron techniques provides microstructural data such as phase fractions as well as crystallographic data, e.g. lattice parameters, from diffraction analysis. Diffraction characterization is complemented by spatially resolved mapping of isotope densities from energy-resolved neutron imaging, in particular neutron absorption resonance imaging, and overall bulk isotope assay with better sensitivity for minority isotopes from neutron absorption resonance spectroscopy without spatial resolution. Furthermore, after characterization at ambient condition, heating of irradiated or spent fuel will allow to characterize differences of e.g. lattice thermal expansion or phase transition temperature and kinetics compared to fresh fuel as well as enable the study of disappearance of irradiation defects. This data enables benchmarking of predictions of properties of irradiated fuels for which otherwise experimental data is sparse. The effort described here strives to characterize a section cut from a high-burnup fuel. Volumes smaller than entire fuel pellets or rodlets as proposed here, e.g. sections cut from a fuel pellet, to pave the way to characterize entire pellets or rodlets in the future.
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PIE of Cycle 13 316SS Cladding Using Aberration Corrected STEM Image and Analys
Optical, SEM and TEM analysis confirmed the presence of moderate sensitization in the 20% cold worked 316SS after the iron aluminide coating was applied. The Cr depletion occurs on the grain boundaries between the closely spaced carbides, and at the interface between the carbides and matrix. Cr depletion varied from locally along the grain boundaries, in many cases showing levels down to 15 wt%, but a few isolated regions had depletion levels down to 12.5 wt%. Optical and SEM images taken from a weld heat affected zone revealed the recrystallization occurred after an end plug cap had been welded to the coated cladding. Irradiation did not affect the carbides in a noticeable manner, but radiation-induced segregation/depletion altered the Cr depletion profiles by removing the wide Cr depletion profile, and replacing it with a much deeper, more narrow profile. Cr levels were depleted to 8-10 wt%, Ni and Si were enriched to levels of 40 wt% and 12 wt% respectively. Substantial precipitation formed in the matrix, and on line dislocations in the matrix and on grain boundary dislocations. Voids were observed in the upper region of the TPBAR where the irradiation temperature exceeds 320C. The levels of Cr depletion and Ni/Si enrichment indicate a material in a state susceptible to irradiation assisted stress corrosion cracking.
316SS cladding↗
Destructive PIE and Safety Testing of Six AGR-5/6/7 Capsule 2 Compacts
This study evaluates fission product retention and particle failure mechanisms in AGR-5/6/7 Capsule 2 uranium carbide and uranium oxide (UCO) tristructural isotropic (TRISO) fuel under high-temperature gas reactor accident-relevant conditions using high-temperature safety tests and destructive postirradiation examination. Three Capsule 2 compacts were held isothermally at 1600°C for approximately 300 hours and one compact at 1800°C for approximately 300 hours; two additional compacts were examined in the as-irradiated state. Post-test deconsolidation–leach–burn–leach (DLBL) quantified nuclide inventories in matrix and particles. Individual particles were surveyed for radioisotope inventories, and microanalytical approaches resolved microstructural evolution and fission product distributions within the coating layers. At 1600°C, no krypton was detected above the minimum detectable limit, and cesium releases were far below a single particle equivalent, indicating the absence of full TRISO failure or SiC failures. Silver releases were limited and primarily reflected depleted postirradiation inventories, consistent with prior compact-level exams indicating substantial in-pile 110mAg loss. At 1800°C, cumulative 134Cs release of approximately 2.5 particle equivalents and delayed 85Kr totaling approximately 0.53 particle equivalents were consistent with one full TRISO failure and two SiC failures. Europium and strontium releases were roughly one order of magnitude higher than at 1600°C and comparable to AGR-1/AGR-2 high-temperature tests, with sustained late-hold rates indicating diffusion through intact coatings coupled with matrix depletion. Overall, AGR-5/6/7 Capsule 2 UCO fuel demonstrated fission product retention during safety testing consistent with prior AGR campaigns, while distinctive in-pile 110mAg depletion and measurable 1600°C europium loss motivate targeted follow-on studies.
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Destructive PIE and Safety Testing of Six AGR-5/6/7 Capsule 2 Compacts
This study evaluates fission product retention and particle failure mechanisms in AGR-5/6/7 Capsule 2 uranium carbide and uranium oxide (UCO) tristructural isotropic (TRISO) fuel under high-temperature gas reactor accident-relevant conditions using high-temperature safety tests and destructive post-irradiation examination. Three Capsule 2 compacts were held isothermally at 1600°C for approximately 300 hours and one compact at 1800°C for approximately 300 hours; two additional compacts were examined in the as-irradiated state. Post-test deconsolidation–leach–burn–leach (DLBL) quantified nuclide inventories in matrix and particles. Individual particles were surveyed for radioisotope inventories, and microanalytical approaches resolved microstructural evolution and fission product distributions within the coating layers. At 1600°C, no krypton was detected above the minimum detectable limit, and cesium releases were far below a single particle equivalent, indicating the absence of full TRISO failure or SiC failures. Silver releases were limited and primarily reflected depleted post-irradiation inventories, consistent with prior compact-level exams indicating substantial in-pile 110m Ag loss. At 1800°C, cumulative 134 Cs release of approximately 2.5 particle equivalents and delayed 85 Kr totaling approximately 0.53 particle equivalents were consistent with one full TRISO failure and two SiC failures. Europium and strontium releases were roughly one order of magnitude higher than at 1600°C and comparable to AGR-1/AGR-2 high-temperature tests, with sustained late-hold rates indicating diffusion through intact coatings coupled with matrix depletion. Overall, AGR-5/6/7 Capsule 2 UCO fuel demonstrated fission product retention during safety testing consistent with prior AGR campaigns, while distinctive in-pile 110m Ag depletion and measurable 1600°C europium loss motivate targeted follow-on studies.
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Completion Report: Neutron Irradiation and PIE of ZrC Coated Surrogate Particle Fuel in the IN-Pile Steady-state Extreme Temperature (INSET) Testbed (#4603)
Explore the source record for details and available documents.
Wilson, Brandon [ORNL] (ORCID:0000000320651647)↗
Telescopes on the Moon or pie in the sky?
The question is examined of whether it makes sense to believe that there will one day be an interferometric array of telescopes on the Moon. The question is really one of national commitment to a lunar base, since it is not likely that a scientific undertaking of this magnitude would occur in the absence of permanent human presence on the Moon. A discussion is also given of the real possibility if the exploration of space should be a multinational effort.
Pilcher, Carl B.↗
PIE Nacelle Flow Analysis and TCA Inlet Flow Quality Assessment
This presentation includes three topics: (1) Analysis of isolated boattail drag; (2) Computation of Technology Concept Airplane (TCA)-installed nacelle effects on aerodynamic performance; and (3) Assessment of TCA inlet flow quality.
Shieh, C. F.↗
PIE: A Dynamic Failure-Based Technique
This paper presents a dynamic technique for statistically estimating three program characteristics that affect a program's computational behavior: (1) the probability that a particular section of a program is executed, (2) the probability that the particular section affects the data state, and (3) the probability that a data state produced by that section has an effect on program output. These three characteristics can be used to predict whether faults are likely to be uncovered by software testing. Index Terms: Software testing, data state, fault, failure, testability. 1 Introduction
Voas, Jeffrey M.↗
Post-Irradiation Examination (PIE) of Westinghouse-led ATF Concepts U3Si2 and U3Si2-UN. Rodlet Non-Destructive Examinations and Sectioning Diagram.
This report contains the results of non-destructive examinations of three ATF-1 rodlets, namely W01, WB2 and WB6. The rodlets are part of the series of irradiation tests led by Westinghouse within the Accident Tolerant Fuel (ATF) Phase 1 campaign. This report is intended to summarize the performed non-destructive examinations gathered so far on these rodlets and to propose a sectioning diagram before proceeding with destructive examinations.
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