Late News Poster Session: Nuclear Materials
Program Organizers: TMS Administration

Tuesday 5:30 PM
March 21, 2023
Room: Exhibit Hall G
Location: SDCC

O-17: Ceramic Crucibles for Nuclear Materials Pyroprocessing: Thomas Dalger1; Ludovic Deliere1; Sophie Le Gallet2; 1CEA; 2ICB, Universit de Bourgogne
    Ceramic crucibles are widely used in pyrochemical processing; it offers good chemical and thermal resistance under severe conditions. However, chlorine salts can corrode ceramic materials, leading to brittleness or impregnation of the crucible by the salt. When pyroprocessing nuclear materials, the latter can result as contamination of the crucible. A study was carried out to understand the phenomenon of ceramics impregnation by molten salts. Ceramic samples were contacted with salts and the surface state of the samples was characterized afterwards. It showed a preferential dissolution route through grain boundaries, and thus the formation of a pathway for the salt to penetrate. Chemical analysis performed on the salt revealed the presence of elements composing the ceramic in the salts. In parallel, an innovative route to prepare ceramic materials is studied, involving 3D printing. It allows the manufacturing of ceramic pieces with complex geometry, which are difficult to make with conventional methods.

O-18: Change of Regulatory's Cladding Model and Its Effects on Steady State Fuel Performance: Yong sik Yang1; Hyun-Gil Kim1; Ju Yeop Park2; 1Korea Atomic Energy Research Institute; 2Korea Institute of Nuclear Safety
     In Korea, it is known that the commercial license of advanced cladding, which contains Nb and has increased corrosion-resistance, will be applied soon. KAERI is modifying a regulatory fuel performance code to be used for the safety and design evaluation by KINS(Korea Institute of Nuclear Safety) during the HANA cladding licensing review. The new code, based on FRAPCON code of NRC/PNNL, can consider the characteristics of HANA cladding, and is particularly interested in phenomena such as corrosion, creep, and irradiation growth which have a great influence on fuel performance. From the results of the preliminary analysis, various effects(fuel centerline temperature reduction, FGR and rod internal pressure decrease, etc) caused by surface oxide thickness reduction were expected.This paper summarizes the preliminary performance evaluation results of nuclear fuel with HANA cladding conducted by using the newly developed code.

O-19: Comparison of Solid-state Structures, Magnetic Susceptibilities and Electronic Properties of UTc3 and URu3: Josephine Libero-Cruzado1; Frederic Poineau1; Daniel Koury1; 1UNLV
    UTc3 and URu3 intermetallic alloys were prepared via vacuum arc-melting. After annealing for 10+ days in vacuum-sealed quartz tubes, cross-sections of the alloys were studied using scanning electron microscopy (SEM) and energy-dispersive x-ray spectroscopy (EDS) to characterize the growth of microstructures and analyze the elemental composition of the alloys. Structural analysis was performed using powder x-ray diffraction (PXRD) on the polished cross section surface. The PXRD patterns were used to experimentally solve the samples’ solid-state structures. The resulting structural information was used to compare the solid-state structure of UTc3 to the known Pm-3m structure of the URu3 alloy. The magnetic susceptibility and the resistivity/conductivity of the samples were investigated using a Physical Properties Measurement System (PPMS). This study investigates the efficacy of ruthenium as a surrogate for technetium in metallic phases. It also provides valuable information regarding the structure and properties of uranium-technetium and uranium-ruthenium alloys.

O-20: High Temperature in-SEM Nanoindentation of TRISO SiC Coatings: Alexander Leide1; Eric Hintsala2; Dong Liu3; 1United Kingdom Atomic Energy Authority; 2Bruker Corporation; 3University of Bristol
    Measuring local properties of TRISO fuel particles at their operating temperature is a notoriously difficult challenge which must be overcome to fully understand the performance of this fuel and the operation of high temperature reactors. In-SEM nanoindentation has been performed using a Bruker PI-89 Picoindenter to measure the hardness and elastic modulus of the SiC layer of TRISO particles at room temperature and 1000 C, continuously measuring properties as a function of contact depth. Post-indentation analysis shows different deformation mechanisms, with plastic deformation at 1000 C and minimal fracture, which correlates to the significantly lower hardness. Oxidation of pyrolytic carbon coating layers occurred above 800 C, which was investigated using plasma-FIB cross-sections, which also revealed the structure of the interfaces between SiC and inner and outer pyrolytic carbon. The combination of high temperature micromechanical property measurement and microstructural analysis is a powerful tool for evaluating advanced nuclear materials.

O-21: Influence of Machining Parameters on Stress Corrosion Cracking Susceptibility of 08CH18N10T Austenitic Steel in Primary Water Environment: Marek Kudlac1; Peter Brziak2; Vladimir Magula2; Katarina Bartova1; Maria Domankova1; Alena Kosinova1; 1Slovak University of Technology; 2Welding Research Institute
    Austenitic stainless steel 08CH18N10T (AISI 304) is frequently used for primary circuit devices manufacturing in VVER nuclear power plants. However, these devices are starving to stress corrosion cracking (SCC) during long term operation in primary circuit water environment. Thermal fields distribution and microstructural features along with the primary water chemistry plays the key role in SCC initiation and propagation. The machining technology is the less visible SCC initiation contributor; indeed, it is more dangerous. In contribution, several machining parameters are employed to find out the influence of surface roughness, surface deformation zone depth and subsurface stresses magnitude and orientation on SCC initiation and propagation. Recirculating corrosion loop working at 270C and 14 MPa, slow strain rate corrosion testing along with X RAY crystallography and scanning electron microscopy were used to determine all SCC initiation aspects.

O-22: Microstructural and Micromechanical Analysis of Steels After Neutron Irradiation: Brandon Bohanon1; Assel Aitkaliyeva1; 1University of Florida
    With implantation of next generation reactor designs and the operation of some US light water reactors extending beyond 60 years, the integrity of materials used in permanent structures after exposure to high fluences need to be assessed. In response to this challenge, this work focuses on neutron irradiated steels that were irradiated to high fluences/displacement damage in pressurized water reactor and advanced test reactor. This contribution will cover both microstructure and mechanical properties of neutron irradiated steels. The microstructure of the specimens was characterized in transmission electron microscope (TEM) and obtained data includes information on dislocation loops, precipitates, and voids. In addition to TEM-based characterization, the mechanical properties of these neutron irradiated steels were accessed using micro-tensile tests using a PI88 PicoIndenter.

O-23: Oxidation Studies of UN/UB2: Megan Pritchard1; Joel Turner1; Timothy Abram1; 1The University of Manchester
     Accident tolerant fuels (ATF) have the potential to enhance safety and lower costs in nuclear power stations by removing the need for safety systems and potentially providing inherent safety. It may be possible to offset increased cladding costs by creating a fuel material that contains more uranium per unit volume, reducing the required enrichment or extending fuel life. Currently there are challenges facing promising fuel materials for light water reactors (LWR) due to their reaction with high temperature steam.This project will explore the effect of light element concentration within UN on its steam reaction. A composite UN fuel also containing UB2 will be manufactured - UB2 appears to improve the hydrolysis behaviour of UN when co-sintered. Characterisation of the composite pellets will be performed alongside steam testing, to understand the effects of UB2 on UN sintering behaviour and oxidation performance.

O-24: Radiation Enhanced Diffusion Along Fast Pathways in Model Oxides: Kayla Yano1; Tiffany Kaspar1; Aaron Kohnert2; Hyosim Kim2; Yongqiang Wang2; Blas Uberuaga2; Daniel Schreiber1; 1Pacific Northwest National Laboratory; 2Los Alamos National Laboratory
    Corrosive degradation of structural alloys in reactors is mitigated by thin protective oxide layers that separate the metal from a corrosive media. However, irradiation produces a population of non-equilibrium point defects that fundamentally alter transport through the irradiated oxide matrix. Two dimensional defects, such as grain boundaries that terminate at the surface, serve as fast diffusion pathways for oxide ingress. The extent to which these pathways enhance transport under irradiation is yet broadly unexplored. In this work embedded isotopic tracers (18O and 57Fe) are used to monitor and quantify atomic transport in irradiated polycrystalline model oxide systems with atom probe tomography. In combination, a chemical rate-theory model provides insight on fundamental transport mechanisms and rates. In this Late News Poster session, results will be delivered on a Fe2O3-Cr2O3 bi-layer system, where previous studies have shown orders of magnitude increases in anion and cation diffusivity in argon irradiated bulk oxides.

O-25: Semi-empirical Modeling of Irradiation Induced Dimensional Change of Nuclear Graphites: Steven Johns1; William Windes1; Anne Campbell2; 1Idaho National Lab; 2Oakridge National Lab
    Nuclear graphite has been used as a moderator material in nuclear reactor designs dating back to the first reactor to reach criticality, Chicago Pile 1 in 1942. In addition, it is anticipated to be used in the conceptual Generation four (GenIV) Molten-salt reactors (MSRs), and the High-temperature gas cooled reactors (HTRs). The macroscopic dimensional change observed in irradiated nuclear graphite is a property change of significant importance. In large part, volumetric change provides valuable insight to the in-service lifetime of graphite components used in nuclear reactors. The dimensional change behavior varies amongst each grade of nuclear graphite due to processing techniques and the resulting microstructure. This work assumes a universal activation energy and an Arrhenius model to predict irradiation dimensional change behavior for all nuclear graphites.

O-26: Study of of Thermal Oxidation to Helium Implantation in 316L Stainless Steel: Minsung Hong1; Angelica Lopez2; Mehdi Balooch1; Yujun Xie1; Ho Lun Chan3; Elena Romanovskia3; John R. Scully3; Djamel Kauomi2; Peter Hosemann1; 1UC Berkeley; 2North Carolina State University; 3University of Virginia
    The effect of thermal oxide layer on He implanted 316L SS was studied to evaluate experimentally how thermal oxidation affects the diffusion and distribution of He in the material. In the case of thermal oxidation of a He implanted sample, as the increasing the oxidation time, the max swelling height increases logarithmically as a function of time and finally saturates at the all samples except for the lowest dose of implanted He. Concerning TEM results, two void regions are identified. Similar to the calculation, the total irradiated depth was around 250 nm and the large void region was formed around 100-150 nm depth. On the other hand, the small void region was observed immediately under oxide layer from the thermal oxidation. In contrast, there were no voids in the altered zone near the metal/oxide interface in the non-thermal oxidized / He implanted sample.

O-27: Synchrotron XRD Hydride Phase Mapping In Zicracloy-2 Cladding: Aaron Colldeweih1; Malgorzata Makowska1; Johannes Bertsch1; 1PSI
    Synchrotron micro-beam X-ray diffraction (XRD) has been used to map the crystallographic phases throughout the cross-section of Zircaloy-2 cladding where a radial delayed hydride crack (DHC) propagated. The sample cross-sections were prepared with a focused ion beam (FIB) to mill a 10-35 m thick lamella. The DHC cracks were propagated with a three-point bending test at 100C, 150C, and 320C in unirradiated Zircaloy-2 nuclear fuel cladding, which contained about 201 wppm H. The spatial resolution provides clues to the critically sized hydride specifically at low temperatures. Tests on cladding with an inner liner have shown that a vast majority of hydrogen precipitates within the inner liner directly at the liner-substrate interface. It has been found that the δ- and ɣ-hydride phases constitute the DHC-responsible hydrides at the tip of arrested DHC cracks and that the volume of material occupied by hydrides is notably larger at high temperatures.

O-28: The Response of Silicon Carbide Composites to He Ion Implantation and Ramifications for Use as a Fusion Reactor Structural Material: Max Rigby-Bell1; Alex Leide1; Slava Kuksenko1; Chris Smith1; Gyula Zilahi1; Louise Gale2; Tony Razzell2; James Wade-Zhu1; David Bowden1; 1UKAEA; 2Rolls-Royce plc
     Silicon carbide fibre–reinforced silicon carbide matrix (SiCf/SiC) composites are desirable in nuclear applications due to their low density, high temperature strength, and tolerance to energetic radiation. We have assessed the tolerance of three industry-grades of composite to high energy He⁺ irradiation, up to 10,000 appm implanted He at 700 C, to determine their suitability for blanket structural applications in fusion. All grades display evidence of crystallographic evolution following irradiation, in the form of phase localised stress, intragranular strain, and lattice swelling. This is attributed to a combination of vacancy production and subsequent He bubble formation. Despite this, no major degradation or microcracking of the composite was observed.During post-irradiation in-situ TEM annealing up to 1300 C, the fibres of the SiCf/SiC composites remain stable without detectable bubbles. Meanwhile, He bubbles in the fibre coating and matrix grains grow, resulting in further crystallographic swelling, with Si phase grains degrading significantly.

O-29: U(AlxSi1-x)3 Surface Composition and Its Interactions with Water Vapor at the Temperature Range of 300-800 K: Shai Cohen1; Maayan Matmor1; Genadi Rafailov1; Moshe Vaknin1; Oshrat Appel1; Noah Shamir2; Shimon Zalkind1; 1Nuclear Research Centre-Negev; 2Ben-Gurion University of the Negev
    The interest in U-Al-Si system stems from its nuclear applications and scientific properties. Al-based alloys containing silicon are used as cladding for uranium fuels and diffusion can occur between the cladding and the fuel. Therefore, study of the U-Al cladding interface is important to determine the fuel rods reliability. In addition, U-Al alloys can be used for research reactor fuel systems. In this study, U(AlxSi1-x)3 surface composition (x=0.57) and its interactions with H2O, at elevated temperatures, were studied, utilizing AES, XPS and DRS. It was found that heating the alloy in UHV causes aluminum segregation to the surface, forming a thin Al layer (~0.7 nm) at 700 K. Exposure of the surface to H2O at 300-500 K results in mild oxidation of U and Al components and adsorption of hydroxyls. Above 600 K, only the aluminum segregated overlayer is oxidized, forming a passivation layer that inhibits further alloy oxidation.