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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

2022 Accomplishments Report Template

Chromium-doped UO 2 fuels is an industrially important Accident Tolerant Fuel (ATF) due to the ability to increase the average grain size relative to standard UO 2 . However, there is a lack of knowledge of the thermophysical properties of these doped fuels, which is necessary for accurately modeling the in-pile performance. The work accomplished this FY focused on developing large grain Cr-doped UO 2 pellets as a function of Cr content and subsequently measuring the thermophysical and mechanical properties. The microstructural variations were also investigated to evaluate chromium formation within the microstructure of the sintered pellets.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

High-Energy X-ray Diffraction Microscopy for Nuclear Forensics FY2022 Project Report

Morphological information on nuclear material has been identified using visible light and scanning electron microscopy. These identify qualitative differences in particle morphology. Three-dimensional imaging of materials through alternating scanning electron microscopy imaging and focused ion beam milling has also been used. Unfortunately, these techniques are time- and labor-intensive, with significant sample preparation required and lengthy analysis times. Further, the resulting 3D images are qualitative, require manual identification, and do not capture statistically-representative populations. High energy X-ray 3D imaging using a direct-beam or diffracted-beam (High-Energy Diffraction Microscopy) have been developed at the Advanced Photon Source and can produce quantitative information on grains (phase, location, etc.) and pores (size distribution, sphericity) in a material. These techniques require only minutes to characterize a sample volume and are non-destructive, thus suitable for a wide range of existing samples and for confirmatory analyses to be carried out using conventional microscopy techniques. In this first year of the project, all uranium oxide samples were synthesized and characterized using conventional analyses by the analytical chemistry laboratory. Conventional analysis methods included powder x-ray diffraction, scanning electron microscopy, impurity analysis via inductively coupled plasma mass spectrometry, and infrared spectroscopy. Impurity analysis shows a drop in boron content from UO 3 to the lowest U 3 O 8 calcination temperature, but otherwise no appreciable difference in any sample. Analysis of diffraction data shows a flip of peaks from UO 3 dominated for the 600 °C calcined sample to U 3 O 8 dominated at 700 °C and 800 °C. Analysis of scanning electron microscopy images shows that with increased calcination temperature the size distribution of particles seems to increase and broaden. Both of these last findings are in line with previously published data, though this work used significantly fewer particles to simply show similar trends instead of getting truly quantitative particle analysis. Infrared analysis similarly shows ingrowth of U 3 O 8 as calcination temperature is increased, along with depression of peaks associated with UO 3 and water. Samples were prepared for analysis at the Advanced Photon Source at beamline 1-ID. It is anticipated that analysis will occur in November of 2022. AI/ML techniques to de-noise data coming out of 1-ID during the analyses was also developed during this time using previously gathered data. Preliminary results using a self-supervision technique called Noise2Selfshow good de-noising of data. Once the uranium oxide samples are analyzed, real data will be used to test the de-noising and other AI/ML techniques that may be developed in the second year of the project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Implementation and testing of physics-based pulverization model in BISON

This report summarizes lower length scale computational research conducted to improve the pulverization criterion for high-burnup UO 2 fuel in the BISON fuel performance code. This research was sponsored by the NEAMS program during FY22. Efforts to improve the model primarily focus on calculating the current pressure of bubbles in the high-burnup structure (HBS) region, as well as the critical bubble pressure at which pulverization occurs in the HBS. The phase-field model for predicting initial bubble pressure in the HBS was improved by implementing a more realistic model for defect production, and by coupling the phase-field model for inter-granular bubble evolution with the spatially-resolved cluster dynamics code Xolotl to simulate intra-granular fission gas evolution. An evolution equation for current bubble pressure was added to BISON, an improvement over the previous model that assumed that bubble pressure was static following HBS formation. To improve calculations of critical bubble pressure for pulverization, 3-D phase-field fracture simulations were performed, and a function for critical bubble pressure was fit—based on the simulation results—to replace the previous function, which had been determined using 2-D simulations. The impacts of these modifications to the existing pulverization assessment cases will be reported in the forthcoming engineering scale milestone on high-burnup UO 2 pulverization. A poromechanics-based approach was used to include the effect of bubble over pressurization on the stress state of the pellet at the engineering scale, and the initial strategy for integration of the pulverization criterion with pellet-scale mechanical degradation using a smeared cracking model was developed. This report also describes the initial research that was conducted to inform a forthcoming transient fission gas release (tFGR) model in BISON. Focusing on the fission gas release caused by pulverization of the outer pellet rim, a model for the amount of fission gas release was developed as a function of fuel porosity, bubble size, and fragment size. Implementation of this model in BISON and its impact of engineering-scale predictions will also be described in the aforementioned engineering-scale milestone on high-burnup UO 2 pulverization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Scaled up Process Report – Apparatus and Model

Advanced voloxidation with NO 2 is a proposed process for used nuclear fuel head-end reprocessing scheme that converts UO 2 to higher oxides, and it also converts partitioning volatile fission products into the gas phase, thus facilitating fuel dissolution and actinide recovery. NO 2 voloxidation is being studied on small batches of UO 2 Simfuel, but the real test of process feasibility will be when it is scaled up to work with >100 g of irradiated material. This report discusses the aspects of scale-up that must be considered for NO 2 voloxidation, including development of a stirred reactor to promote agitation of the mixture during processing, online process monitoring, and automation controls. Brief details on parallel efforts are also provided in this report, including (a) demonstration of iodine release from Simfuel made by Spark Plasma Sintering, and (b) development of an order-of-magnitude scale-up to react 100 g of UO 2 Simfuel in a metal reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Dissolution Flowsheet for Non-Aluminum Spent Nuclear Fuel Campaign 1

As part of the Accelerated Basin De-inventory (ABD) program, H Canyon plans to dissolve non-aluminum spent nuclear fuel (NASNF) in the 6.3D electrolytic dissolver. NASNF Campaign 1 plans to electrolytically dissolve 68 bundles of fuel assemblies from the Carolinas-Virginia Tube Reactor (CVTR), Heavy Water Components Test Reactor (HWCTR), and Experimental Boiling Water Reactor (EBWR). The fuel assemblies are intact Zircaloy or stainless steel (SS) clad UO 2 rods, tubes, and plates. The H Canyon electrolytic dissolver previously dissolved a variety of UO 2 core fuel types in SS, Zircaloy, Nichrome, or Incoloy cladding from 1969 to 1980. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely dissolve NASNF Campaign 1 bundles in the H Canyon electrolytic dissolver. Bench-scale electrolytic dissolution tests were performed to demonstrate a flowsheet for NASNF Campaign 1 bundles. The outer bundles are composed of SS or Al alloy, Al 6061-T6, and contain intact Zircaloy or SS clad UO 2 fuel assemblies. The key objectives of these tests were to determine bounding dissolver chemistries and the sparge requirement to ensure H 2 concentration remain less than 60 vol % of the lower flammability limit (LFL) during dissolution. The impact of HNO 3 concentration and the addition of fluoride on the dissolution efficiency of Zircaloy, 304L SS, Al 6061-T6, and Inconel 625 were examined. While SS, Al, and Inconel 625 readily dissolve utilizing electrolytic dissolution, Zircaloy disintegrated anodically; the surface of Zircaloy oxidized and the oxide layer spalled off and settled at the bottom of the dissolver as an insoluble material. The black flakes were identified as ZrO 2 and 85% of the Zr processed was converted to black ZrO 2 flakes when Zr was anodically disintegrated in 9.5 M HNO 3 .

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Fabrication and Characterization of MiniFuel Specimen

Burnable absorbers (BAs) are used in UO 2 fuels to tune the reactivity in a light water reactor (LWR) particularly at the beginning of life of the fuel. BAs are elements that have high neutron absorption cross sections and because of this property can prevent power peaking. The addition of BAs reduces the reactivity at the beginning of life, and this in turn allows for the increase of enrichments from > 5 % 235 U up to ~ 10 %. This also enables the increase of the burnup limit to 75 GWd/MTU and the cycle length to 24 months. The BA are eventually consumed under irradiation, and their concentration is diminished at roughly 10-20 GWd/MTU. The objective of this report is to fabricate UO 2 fuels that have been doped with BAs (Gd 2 O 3 in this case) and subsequently irradiate them in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. The test specimen that are fabricated include 4 and 8 weight % Gd 2 O 3 doped UO 2 to investigate the influence of the BAs on the thermophysical properties and microstructure of the fresh and irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Bridging multimodal microscopy for advanced characterization on nuclear fuel using machine learning

Uranium dioxide (UO 2 ), widely used as driver fuel in light water reactors, experiences microstructure and property change by nuclear fission reactions. This paper bridges the characterization of fresh UO 2 fuel at different length scales, serving as a baseline for future post irradiation examination of irradiated UO 2 fuel. To characterize the microstructural change of nuclear fuel, modern approaches cover a wide range of length scales through different characterization techniques, such as mm scale for Synchrotron-based X-ray computed tomography (SXCT) and microscale for focused ion beam (FIB) and scanning electron microscopy (SEM). It is challenging to bridge the data and knowledge of the same sample in different length scales. This paper proposed a deep learning framework leveraging transfer learning to detect microstructural defects, trained from a sparse FIB, SEM, and SXCT images. The proposed model achieved superior performance in defect segmentation on multiscale microscopic data compared to four of the latest deep learning models.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Thermal Process Technology for Nuclear Applications - 20379

ANSTO's Synroc technology has been developed to provide a safe, secure matrix for the immobilization and final disposal of radioactive waste. Synroc technology will be used to manage radioactive wastes from the production of the radioisotope Molybdenum-99 (Mo-99). This paper shall outline various stages of the process development with specific reference to the thermal treatment technology of calcination. Calcination is a key step in the Synroc process [1-2].The rotary thermal processing system includes: an advanced heating element design for increased robustness and ease of remote operation and maintenance, an enhanced modular design of components for ease of remote maintenance in a hot cell and in compliance with hot cell radioactive environment requirements for safety, reliability and maintainability. In addition to thermal treatment of waste from nuclear medicine production, this technology provides solutions for a variety of nuclear materials processing applications including sintering UO{sub 2} pellets for reactor fuel rods, oxidation of UO{sub 2} pellets, swarf, and powder to U{sub 3}O{sub 8}, de-nitration of Uranyl nitrate and hydrofluorination of UO{sub 2} pellets. The paper will also discuss thermal processing solutions for a range of nuclear applications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Redox‐Dependent Luminescence in a Pentavalent Uranyl Complex

The chemistry of pentavalent uranyl is of high interest due to the ion's central role as an intermediate in nuclear separations. However, due to the instability of [UO 2 ] + , little is known about its electronic structure. The synthesis and characterization of a pentavalent uranyl complex supported by a pyridine dipyrrolide pincer‐type ligand are reported. The pentavalent species is isolated following reduction of the hexavalent uranyl congener, allowing for direct comparisons to be made between the two species. The [UO 2 ] 2+ analog is nonemissive, but upon reduction, luminescence is “switched on” in the pentavalent uranyl complex, affording emission from a high‐energy state. Quantum chemical calculations support the experimental findings and help provide insight into the electronic structure of the excited states of the reduced complex. This work provides a unique example of ligand‐based non‐Kasha emission from a molecular pentavalent uranyl complex.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Controlling Cation-Cation Interactions in Uranyl Coordination Dimers by Varying the Length of the Dicarboxylate Linker

The chemistry of linear uranyl(V/VI) dioxo cations, [O yl –U–O yl ] +/2+ , is dominated by coordination of uranium in the equatorial plane. Effects of this constraint were evaluated by experiment and theory for gas-phase mixed-valence U V/VI coordination dimers in which uranyl moieties are linked by alkyl dicarboxylates, [(UO 2 + )(UO 2 2+ )(OOC-(CH 2 ) n-2 -COO 2– ) 2 ] – (n = 3–12). Faster O 2 -addition to dimers with short linkers n = 3 and 4, vs. n ≥ 5, suggests a structural difference. Computed structures with the shortest linkers have bridging dicarboxylates and nearly parallel, non-interacting uranyls. Longer linkers, n = 5–7, accommodate uranyl orientations with distinct U V –U VI end-on cation-cation interactions (CCIs), whereby Lewis base O yl from U V coordinates to the acid U VI , denoted as U V O yl ···U VI . The dimer structure for n = 8 has a U V –U VI side-on diamond-shape CCI, with U V O yl ···U VI and U VI O yl ···U V interactions. Addition of O 2 to the n = 4 and 5 dimers yields [(UO22+)2(OOC-(CH2)n-2-COO2–)2(O2–)]–, with U V oxidized to U VI and O 2 reduced to O 2 – . Whereas O 2 can associate to and oxidize the exposed U V center for dimers with n = 3 and 4, the more crowded U V site in the CCI structures inhibits O 2 addition. Finally, the results demonstrate rational structural control of uranyl-uranyl bonding and reactivity in small coordination complexes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Interband Transitions and Critical Points of Single-Crystal Thoria Compared with Urania

The interband transitions of UO 2 are validated independently through cathode luminescence. A picture emerges consistent with density functional theory. While theory is generally consistent with experiment, it is evident from the comparison of UO 2 and ThO 2 that the choice of functional can significantly alter the bandgap and some details of the band structure, in particular at the conduction band minimum. Strictly ab initio predictions of the optical properties of the actinide compounds, based on density functional theory alone, continue to be somewhat elusive.

42 ENGINEERING↗
DOE OSTI2020

The electronic structure of the actinide oxides and their singly and doubly charged cations: A ligand field approach

Abstract Ligand field theory (LFT) calculations of energy levels were performed for the neutral actinide monooxides ( An O) and their singly and doubly ionized cations ( An O + and An O 2+ ) by treating the molecular electronic states as An m+ free‐ion energy levels (where An ∈ Th through Lr and m = 1, 2, 3, or 4) perturbed by the electric field of O 2− or O − . LFT parameters obtained from fits to the energy levels of ThO, ThO + , UO, and UO + were used to compute molecular energy levels for the lowest energy (maximum S c , maximum L c ) 5 f ‐core states of An 4+ , An 3+ , An 2+ , and An + for the majority of the An 4+ O 2− , An 3+ O − , An 3+ O 2− , An 2+ O − , An 2+ O 2− , and An + O − electronic configurations. Simple linear relationships enabled predictions of the dissociation energies for An O, An O + and An O 2+ (where An ∈ Bk through Lr) and ionization energies for An O and An O + (where An ∈ Bk through Lr), mainly based on recent accurate experimental data for the ionization energies of An atoms (where An ∈ Fm, Md, No, and Lr) and correlations with the energetics of the atoms and ions.

Kaledin, Leonid A.↗
DOE OSTI2024

Amorphous K–Co–Mo–S x Chalcogel: A Synergy of Surface Sorption and Ion-Exchange

Chalcogel represents a unique class of meso- to macroporous nanomaterials that offer applications in energy and environmental pursuits. We report the synthesis of an ion-exchangeable amorphous chalcogel using a nominal composition of K 2 CoMo 2 S 10 (KCMS) at room temperature is reported. Synchrotron X-ray pair distribution function (PDF), X-ray absorption near-edge structure (XANES), and extended X-ray absorption fine structure (EXAFS) reveal a plausible local structure of KCMS gel consisting of Mo 5+ 2 and Mo 4+ 3 clusters in the vicinity of di/polysulfides which are covalently linked by Co 2+ ions. The ionically bound K + ions remain in the percolating pores of the Co–Mo–S covalent network. XANES of Co K-edge shows multiple electronic transitions, including quadrupole (1s→3d), shakedown (1s→4p + MLCT), and dipole allowed 1s→4p transitions. Remarkably, despite a lack of regular channels as in some crystalline solids, the amorphous KCMS gel shows ion-exchange properties with UO 2 2+ ions. Additionally, it also presents surface sorption via [S∙∙∙∙UO 2 2+ ] covalent interactions. Overall, this study underscores the synthesis of quaternary chalcogels incorporating alkali metals and their potential to advance separation science for cations and oxo-cationic species by integrating a synergy of surface sorption and ion-exchange.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Application of HRGS for forensic characterization of uranium oxides, pure uranium metals and uranium alloys

A nondestructive iterative method for uranium-bearing material characterization with HRGS developed earlier in (J. appl. Rad. Isotopes, 166 (2020) 109433) is applied to determine matrix densities and uranium isotope masses of a variety of uranium materials, namely uranium ore, UO 2 and U 3 O 8 powders, fuel elements in the form of UO2 microspheres, uranium metal and uranium alloys. It is shown that U 3 O 8 powders with uranium mass fraction of about 84% can be distinguished from the powders of UO 2 with uranium mass fraction of about 87%; uranium products in the form of liquid or loose powder with matrix density of 0.5-2.0 g/cm 3 can be distinguished from uranium products in the form of compacted fuel elements with matrix density of 6.0-10.0 g/cm 3 and from pure metal uranium and uranium alloys with matrix density of 14.0- 19.0 g/cm 3 . In fuel microspheres based on UO2 the uranium mass fraction 88.02% measured by HRGS is consistent, within the measurement uncertainties, with the results of isotope dilution mass spectrometry 87.76±0.64 % and also is confirmed by X-ray diffraction technique. The uranium mass fraction of the uranium ore estimated as 0.08% by HRGS is consistent, within the measurement uncertainties, with the value 0.09±0.01% determined with WDXRF. Densities of two different uranium metal samples, estimated as 18.42 g/cm 3 and 19.33 g/cm 3 by HRGS are consistent with values 18.24±0.55 g/cm 3 and 18.86±0.59 g/cm 3 , respectively, obtained by the gas pycnometry technique.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Synthesis and spectroscopic characterization of actinyl(VI) tetrahalide coordination compounds containing 2, 2'-bipyridine

A series of three new actinide tetrahalide coordination compounds containing 2,2'-bipyridinium (H 2 bpy) have been synthesized by room temperature evaporation and characterized via single crystal X-ray diffraction, Raman spectroscopy, and infrared spectroscopy. The solid-state compounds (Np1, U1, and U2) contain the AnO 2 2+ cation coordinated to four halide (chloride or bromide) atoms to form the anionic tetrahalide molecular unit [AnO 2 X 4 ] 2– (X = Cl, Br) and are charge balanced by the 2,2'-bipyridinium cations that engage in electrostatic interactions with the actinyl complex. Np1 (N 2 C 10 H 10 )[NpO 2 Cl 4 ] contains 2,2'-bipyridine in the cis conformation where both the nitrogen atoms are directed towards the [NpO 2 Cl 4 ] 2– monomer. Within U1 (N 2 C 10 H 9 )[UO 2 Cl 4 ], the 2,2'-bipyridine is in a trans conformation, and found in the same orientation in the isomorphous structure U2 (N 2 C 10 H 9 )[UO 2 Br 4 ]. The compounds were further characterized by vibrational spectroscopy with specific interest in the actinyl cation. Raman spectroscopy provided information on the actinyl symmetric stretch (ν 1 ) whereas IR spectroscopy is used to determine the location of the asymmetric stretch (ν 3 ). Here, spectral similarities are observed between the two isomorphous structures, but slight variations across the series of structures are found to be associated with bipyridine bands in the various conformations. Additional analysis of these vibrational features in U1 and U2 provided information on the actinyl force constants, allowing insight into the relative uranyl bond strength that was compared to similar tetrahalide compounds.

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

The impact of feedstock size and composition on the hydrothermal growth of (U,Th)O 2

Bulk crystal growth of refractory oxides requires unique approaches, and the U x Th 1-x O 2 solid solution is no exception. Hydrothermal transport reactions of U 0.1 Th 0.9 O 2 onto ThO 2 seeds with feedstocks comprised of U 0.1 Th 0.9 O 2 or mixtures of UO 2 and ThO 2 with a nominal composition of U 0.1 Th 0.9 O 2 are investigated with µ-Raman spectroscopy and X-ray fluorescence. In each case, the trends in stoichiometry as a function of distance from the seed are analyzed and the deviations from the nominal stoichiometry discussed. When the feedstock is composed of mixed oxides, the particle surface area’s influence on the feedstock dissolution rate is the dominant factor and can produce U-rich stoichiometries as high as U 0.75 Th 0.25 O 2 . When composed solely of U 0.1 Th 0.9 O 2 , the total growth amount depends on the particle size, but the obtained stoichiometry is the product of differing solubilities. Although the U0.1Th0.9O 2 feedstock produces the most homogeneous growth, the ending growth of the smallest particle UO 2 /ThO 2 mixture yields the stoichiometry closest to U 0.1 Th 0.9 O 2 . How to obtain desired stoichiometries from mixtures by matching the particle size and solubilities is briefly discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Epsilon Metal: A Waste Form for Noble Metals from Used Nuclear Fuel

Epsilon metal (e-metal) is the metallic phase that forms as inclusions at the grain boundaries in the UO2 fuel during reactor operation. This metal is composed of Pd, Mo, Rh, Ru, and Tc. These metallic inclusions are insoluble in strong acid and remnants of these metallic inclusions have been found in the UO2 matrix that remains from the natural reactors in Gabon that were active 1.8 billion years ago, therefore e-metal should be an excellent waste form for the immobilization of the long-lived isotopes 107Pd (6.5 ? 106 a) and 99Tc (2.13 ? 105 a), with 99Tc being the isotope of interest for repository performance. Therefore, the chemical durability of this potential waste form is assessed in this study. Typically, corrosion rates for metallic materials are measured electrochemically because they are quick, inexpensive, and can reveal the mechanism by which a metal corrodes, at least initially. However, in a repository the waste form would be subjected to slowly flowing water without an applied electrical potential over long time periods. Therefore, the corrosion rates of e-metal specimens were measured with both electrochemical tests and the single-pass flow-through test (SPFT). Potentiodynamic and potentiostatic polarization results suggest that a thin passive film exists on the alloy surface, which seems to be responsible for its high corrosion resistance. Additionally, X-Ray photoelectron spectroscopic results suggest that Pd oxides are significantly enriched in the passive film Results from the SPFT show that the dissolution rates were weakly dependent on pH. Only Mo and Re were found in solution and were used for the calculation of the dissolution rates. In general, the electrochemically determined corrosion rates agree reasonably well with the initial dissolution rate measured with the SPFT test, but they are about one or two orders of magnitude higher than the steady state rates. The causes for this discrepancy are discussed.

epsilon-metal, nuclear waste form, corrosion, sing↗
DOE OSTI2020

Mechanical failure of fresh nuclear grade iron–chromium–aluminum (FeCrAl) cladding under simulated hot zero power reactivity initiated accident conditions

The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiated by a rapid increase in reactivity which causes an increase in fission rate and fuel temperature. One potential mode of fuel system failure during RIA is pellet cladding mechanical interaction (PCMI) due to the rapid thermal expansion of the fuel pellet. Simulated PCMI experiments were performed by rapidly pressurizing iron-chromium-aluminum (FeCrAl) cladding tube samples using a hydraulic modified burst test system, causing the specimens to burst under a biaxial loading path. Deformation and rupture of the specimens were tracked with a telecentric lens and high-speed camera system. Outer surface strains were calculated using digital image correlation (DIC) on speckle patterns painted on the specimens’ outer surfaces. Experiments were conducted at approximately 275 °C, representative of hot zero power RIA conditions. The failure hoop strain was DIC-calculated between approximately 1.8–3.4%, corresponding to quasi-static energy depositions of approximately 110–260 calories per gram UO 2 assuming initial pellet-cladding contact. These conditions are very similar to the proposed energy deposition limit of 150 calories per gram UO 2 for unirradiated zirconium-based cladding in US Nuclear Regulatory Commission (NRC) Draft Regulatory Guide 1327. A very small strain rate dependence was observed in the data, with the magnitude of the failure hoop strain decreasing slightly with increasing strain rate. This observed dependence may be relevant because the FeCrAl cladding strain rate will be approximately 20% higher than in zirconium-based cladding due to the harder neutron spectrum and resultant shorter neutron generation time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗