Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials: On-Demand Poster Session
Sponsored by: TMS Structural Materials Division, TMS: Mechanical Behavior of Materials Committee, TMS: Nuclear Materials Committee
Program Organizers: Dong Liu, University of Oxford; Peng Xu, Idaho National Laboratory; Simon Middleburgh, Bangor University; Christian Deck, General Atomics; Erofili Kardoulaki, Los Alamos National Laboratory; Robert Ritchie, University of California, Berkeley
Monday 8:00 AM
March 14, 2022
Room: Nuclear Materials
Location: On-Demand Poster Hall
Microstructural and Mechanical Properties of Hot Deformation Behavior of Zr-4 Alloy
: Gaurav Singh1; Raviraj Verma1; Vishnu Narayanan KI2; Umesh Arora2; R Jayaganthan1; 1Indian Institute of Technology Madras; 2NFC Hyderabad
Zirconium alloy shows a good combination of strength and ductility, low neutron absorption cross-section, and excellent corrosion and oxidation resistance. The hot deformation behaviour of Zr-4 was studied in the temperatures (650oC and 750oC) with strain rates (0.01 and 1 s-1). It was observed that the flow stress of Zr alloy is high at a high strain rate (1 s-1) and low temperature (650oC). The true stress-strain curve of the alloy has shown the serrated oscillation characteristics. Electron backscattered diffraction (EBSD) was used to characterize y evolution of microstructural and texture features in hot deformed samples. Microhardness of measurements were performed in hot deformed samples was measured and it was correlated with the microstructural characteristics such as grain size and texture.
Investigation of Degradation Mechanism of Accident Tolerant Fuel (ATF) Coated Cladding Concepts during Interim Storage and Transportation of Used Nuclear Fuels: Evan W1; Hwasung Yeom1; Tyler Dabney1; Kumar Sridharan1; Andrew Nelson2; Tim Graening2; 1University of Wisconsin Madison; 2Oak Ridge National Laboratory
Thin coatings for zirconium-alloy light water reactor (LWR) cladding are being explored as near-term accident tolerant fuel (ATF) concepts. However, it is not clear how these ATF coated claddings will perform in the back end of the fuel cycle. For conventional Zr-alloys, delayed hydride cracking (DHC) is a concern in the used fuel interim storage and transportation stages. It is of interest to see how coating influences hydriding and consequently the DHC behavior of the underlying Zr-alloy. In coated cladding design, the effect of the formation of intermetallic compounds at the interface between the coating and the Zr-alloy on mechanical behavior must also be evaluated. The study investigates these effects using Cr coated cladding concepts that are at the forefront in industry. Mechanical properties have been evaluated using nanoindentation, ring compression tests, and ring tensile tests, coupled with microstructural analysis using scanning electron microscopy.
Formation of UN in U-Mo Systems by Mechanical Alloying: James Zillinger1; Nathan Jerred1; Adrian Wagner1; Samrat Choudhury2; Indrajit Charit2; 1Idaho National Laboratory; 2University of Idaho
U-Mo alloys show promise as a nuclear fuel system due to their high thermal conductivity and high fuel loading capability. However, U-Mo systems suffer from irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interaction. To address these shortcomings, this work investigated the formation of UN particles within a U-10Mo matrix to improve its irradiation performance. To promote the formation of UN, U-10Mo powders were milled under a high purity N2 atmosphere. Subsequent analysis has shown nitrogen successfully diffused into the alloy. Milling times, intensities (rpm), and media sizes were varied to observe their effects on resulting material characteristics. Mechanically alloyed powders were thermally treated to induce the formation of a UN phase. Powder and feedstock alloy samples were analyzed via light-elemental analysis, dissolution chemistry, SEM, XRD, and TEM to provide a clear understanding of how every step of the material processing affected fuel behavior and microstructure.
Impact Fretting Wear Behavior of Cr-alloy Coating Layer for Accident-tolerant Fuel cladding: Youngho Lee1; Dong-Jun Park1; Yang-Il Jung1; Sung-Chan Yoo1; Hyun_Gil Kim1; 1Korea Atomic Energy Research Institute
Cr-alloy coated fuel cladding was proposed to improve resistance to corrosion in high temperature steam, which was coated with various coating methods on the conventional Zr-based fuel cladding. When a certain gap is opened between the fuel cladding and the spacer grid due to spring relaxation and irradiation during the normal operations, fretting wear mode can be changed from positive contact force to impacting wear. In this study, impact fretting wear characteristics of Cr-alloy coating layer were evaluated in various impact force, velocity and displacement in room temperature water. Fractured layer and trapped wear debris in the grooves were predominantly observed in the Zr-based cladding, whereas flat-like deformed layer and no crack formation was observed in the Cr-alloy coating layer. It is concluded that the Cr-alloy coating layer showed improved fretting wear resistance without any significant degradation under small impacting force below 3 N.
Comparison of Neutron Irradiation Effects in PM-HIP and Cast Grade 91 Steel: Sri Sowmya Panuganti1; Yu Lu2; Sheng Cheng3; Megha Dubey3; Yangyang Zhao1; Caleb Clement1; Donna Guillen4; David Gandy5; Janelle Wharry1; 1Purdue University; 2Boise State University ; 3Boise State University; 4Idaho National Laboratory; 5Electric Power Research Institute
The purpose of this research is to compare irradiation effects in Grade 91 steel fabricated by casting versus powder metallurgy with hot isostatic pressing (PM-HIP). Grade 91 is a model ferritic steel proposed for advanced nuclear systems. PM-HIP components are of interest for nuclear applications because they are produced near-net shape, at lower cost, and have more uniform microstructures. But PM-HIP steels have not yet been thoroughly explored under irradiation. In this study, both PM-HIP and cast Grade 91 are irradiated in the Advanced Test Reactor to 1.0 and 3.0 dpa at 400oC. Tensile testing and nanoindentation reveal comparable irradiation hardening between PM-HIP and cast specimens, although the PM-HIP develops a yield-point phenomenon after irradiation. TEM and APT are used to characterize irradiation-induced dislocation loops and nanoclusters. Heterogeneous loop distributions and their role on source hardening will be discussed with respect to the yield-point phenomenon in the PM-HIP alloy.