Research

The mechanical properties of additively manufactured (AM) materials significantly depend on AM processing and post-build treatments due to microstructural changes. This research proposal aims to develop a microstructure-based physics model to predict the mechanical strength of laser powder bed fusion (LPBF) manufactured 316H stainless steel (SS) after post-build processing. This model will be informed by advanced characterization of the phase structure and composition, precipitates and defects size and number density, and mechanical property data. The models will be implemented in the LPBF 316H SS performance code to investigate the effects of post-build treatment and long-term aging on the mechanical properties of LPBF 316HSS. The project is aligned with industry needs and will benefit the Nuclear Regulatory Commission (NRC) by closing one of the critical knowledge gaps associated with the qualification of LPBF 316H SS for advanced non-light water reactors (ANLWRs).

Advanced radiation-resistant materials with high mechanical strength and corrosion resistance are essential for the advancement of sodium-cooled fast reactors (SFR) and molten salt reactors (MSR). However, the steels currently employed in these applications have limitations, such as inadequate irradiation performance, insufficient high temperature strength, and compatibility issues in corrosive environments. To address these challenges, the proposed research will focus on a comprehensive investigation of the neutron irradiation effects on the mechanical properties of a novel class of alloys known as high entropy alloys (HEAs). It is hypothesized that the complex nature of HEAs leads to a retardation of radiation damage accumulation, resulting in enhanced radiation damage tolerance. The mechanical deformation experiments, including tensile testing, microhardness measurements, and nanoindentation, on both pristine and neutron-irradiated HEA samples will be performed. These experiments will enable us to determine the strength, hardness, and Young’s modulus of the materials. By comparing the pristine and irradiated HEAs, the effects of neutron irradiation on material degradation will be revealed. A multimodal characterization approach spanning from the atomic scale to the mesoscale will be used to examine microstructural evolution under both irradiation and deformation conditions. This characterization will involve the identification of radiation-induced defects, assessment of chemical composition in the bulk and at interfaces, analysis of phase changes, and investigation of defects and grain structure changes induced by mechanical deformation. Additionally, in situ small-scale mechanical testing will be employed to evaluate the local mechanical properties around specific features such as grain and phase boundaries. The selection of samples for these tests will be based on microstructural characterization and results obtained from macroscale mechanical testing. Furthermore, advanced microstructural data will be utilized to validate mesoscale models that predict the deformation behavior at the structural mechanics level. A polycrystal microstructure will be reconstructed using experimental data, forming the basis for mesoscale simulations. The deformation model employed in these simulations will incorporate dislocation glide and climb mechanisms based on crystal plasticity theory, and it will be coupled with a phase-field model for crack growth. Once the simulations have been validated against experimental data, they will be utilized to predict the mechanical strength of HEAs. By conducting this in-depth study, we aim to gain valuable insights into the behavior of HEAs under neutron irradiation, elucidating the mechanisms behind their radiation resistance and evaluating their potential as superior materials for advanced reactors.

Stainless steel 316H (SS 316H) has garnered significant attention from the nuclear industry as a promising structural material for deployment in advanced reactors, particularly the Molten Chloride Fast Reactor (MCFR) and fluoride salt-cooled high-temperature reactor (FHR). One of the key advancements that has facilitated the exploration of this material is additive manufacturing (AM), which allows complex designs that were previously unattainable using traditional manufacturing methods. Recent advancements have been made in the development of in-pile molten salt corrosion capability for the PULSTAR reactor at NCSU, while creep/fatigue testing capability in molten salt environments has been achieved at UCI. These two distinctive capabilities offer valuable opportunities into the behavior of materials, such as SS 316H, under molten salt corrosion conditions within reactor environments. Our goal is to leverage a blend of innovative molten salt corrosion experiments and cutting-edge characterization techniques to advance our understanding of molten salt corrosion behavior in both commercial and AM SS 316H, particularly in radiation or stress environments. The corrosion behavior of SS 316H in reactor environments can be significantly influenced by microstructure variations resulting from different manufacturing processes. To explore this idea comprehensively, we have designed a multifaceted research approach consisting of the following key investigations: 1) To assess the impact of manufacturing on corrosion resistance, we will conduct electrochemical studies comparing two distinct types of SS 316H in MgCl 2 -NaCl eutectic salt. 2) We will subject both types of SS 316H to in-pile corrosion testing within the PULSTAR reactor at NCSU, allowing us to analyze the specific corrosion response under neutron irradiation. 3) To understand the effects of stress on corrosion behavior, we will conduct creep-fatigue experiments in controlled molten salt environments at UCI. The in-pile corrosion and creep-fatigue experiments present unique test conditions, enabling us to explore the corrosion resistance of SS 316H under neutron irradiation and stress, respectively. To comprehensively investigate microstructural evolution during corrosion in various environments, we will employ a multimodal characterization approach spanning from atomic to mesoscale levels. This thorough analysis will encompass dislocations, voids, cracks, precipitates, and theirinteractions with cellular boundaries and grain boundaries of SS 316H.

The goal of the current proposal is to enhance the scientific understanding of the mechanisms leading to high-burnup structure (HBS) formation in nuclear fuel and develop experimentally validated computational capabilities for predicting grain subdivision and associated fission products evolution. To improve the economics of commercial nuclear power production, utilities are seeking to increase the allowable burnup limit of UO2 fuel. However, exposure to extended burnup leads to the formation of a refined grain structure (i.e., HBS), which generates the risk of performance degradation and fragmentation of UO2 fuels. Therefore, understanding the HBS formation mechanisms and its impact on fuel performance is important.

Phase I of the EPRI/NCSU research program proposes to develop a multiaxial testing system (MTS) for testing alloys within molten salt flow conditions to develop needed data towards qualifying EPRI-selected alloys for the ASME Code nuclear application. The MTS will be coupled with a molten salt pump loop to be fabricated by Copenhagen Atomics (CA) for flowing molten salt with desired flow rates and temperatures. A fifth-generation CA molten salt pump loop will enable customized material and component testing by connecting to external plumbing hookups. In addition, with a CA loop, specimens and components can be corroded to different degrees for subsequent performance evaluation testing of pre-corroded specimens. The proposed novel system will investigate the degradation mechanisms of selected alloys under fatigue, creep, and creep-fatigue loading conditions in air and molten halide salt environments. The proposed Phase I project will develop a novel and essential testing system for addressing the MSR material technology gaps. After successfully developing the MTS, systematic tests of selected alloys will be performed in the future phases of the EPRI/NCSU research program needed to develop ASME Code cases for Section III Division 5 for EPRI-selected alloys. Upon developing the proposed MTS, strain- and stress-controlled cyclic tests in the temperature range of MSR will be performed in Phase II. Both in situ MS testing on virgin specimens and ex-situ SEM/TEM and in situ SEM testing of pre-corroded and/or pretested specimens will be performed to probe the degradation mechanisms. The test data will enable the development of physics-based models to reduce future test cases. We anticipate that developing a systematics test database with the proposed testing system will accelerate the development of the Sec III Div. 5 ASME Code Cases for selected alloys. Rapid deployment of the CA loop with customized external ports will enable new opportunities for MSR-relevant testing, including plumbing components (e.g., valves and pumps), novel sensors, and salt processing. Additionally, the operational experience gained during Phase I with the pumped loop will inform the design criteria for the Large-Scale Molten Salt Flow Loop Test Facility, which is the overarching goal of the EPRI/NCSU collaborative project. During Phase I, initial design studies will be conducted to identify requirements and constraints for the Large-Scale Facility.

We propose to study the role of grain boundaries (GBs) on mass transport in extreme environments by combining well-designed corrosion and irradiation experiments, multimodal characterization and multimodal computation. The goal of this project is to obtain a fundamental understanding of the unit mechanisms governing defect interaction at GBs as a function of GB character and GB chemistry in extreme radiation/corrosion/stress environments. General studies of corrosion and irradiation will necessarily have a large parameter space (ion type, ion energy, corrosion environment and material composition). In order to reduce this problem to something trackable, focus will be primarily limited to single energy proton irradiation, single component molten salt corrosion on pure nickel (Ni) and binary nickel-chromium (Ni-Cr) alloys with various GB characteristics. The research goal will be achieved by making a progressive approach: 1) understanding the mass transport mediated by GBs in pure corrosion and pure irradiation environments, 2) understanding mass transport mediated by GBs in corrosion + stress and corrosion + irradiation environments 3) understanding mass transport mediated by GBs in corrosion + irradiation + stress environments.