Lizhen Tan

Materials Challenges for Generation IV Nuclear Energy Systems

Abstract The U.S. Department of Energy is sponsoring the Generation IV Initiative in the United States for the purposes of developing future-generation nuclear energy systems. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. Critical to the development of Generation IV concepts is successful development and deployment of materials that operate successfully in the aggressive operating environments envisioned in the Generation IV concepts. This paper summarizes the Generation IV operating environments and describes materials challenges and potential solutions, including crosscutting solutions applicable to multiple Generation IV concepts.

Altering Corrosion Response via Grain Boundary Engineering

Grain boundary engineering (GBE) was employed to improve the oxide exfoliation resistance and mitigate oxide growth by optimizing the grain boundary character distribution. Studies were performed on alloys of Incoloy 800H and Inconel 617. Alloys 800H and 617 were selected due to their potential applications for the Generation IV nuclear power systems. The effect of GBE on the corrosion response was evaluated using supercritical water exposure tests and cyclic oxidation tests. The microstructure of the tested samples was analyzed by means of optical microscopy, scanning electron microscopy, energy dispersive X-ray spectroscopy, electron backscatter diffraction, and gravimetry. The effects of thermal expansion mismatch and Cr volatilization on the corrosion response are discussed.

Radiation Resistance of Advanced Ferritic-Martensitic Steel HCM12A

HCM12A is an advanced 12 Cr ferritic-martensitic steel designed for higher temperature operation than could be achieved using earlier generation steels such as HT9. HCM12A is one of the advanced alloys under consideration for application in core components in Generation IV nuclear energy systems, and is of particular interest to the supercritical water reactor, sodium fast reactor, and lead fast reactor designs. The radiation resistance of HCM12A has not previously been studied. This work provides information on the hardening and microstructural changes in HCM12A after irradiation using 2.0 MeV protons at 400°C to 10 dpa and 5 MeV Ni-ions at 500°C to 50 dpa. Following irradiation, changes in hardness were measured using Vickers hardness indentation, changes in microstructure and phase stability were studied using transmission electron microscopy, and changes in microchemistry were measured using scanning Auger microscopy. The hardness increases by roughly 70 % and saturates by roughly 5 dpa. The changes to the microstructure contributing to this hardness increase are primarily due to the formation of precipitate phases, with some contribution from changes in dislocation density. Chromium is enriched at grain boundaries prior to irradiation, likely due to grain boundary carbides, and increases further during the irradiation.

Experimental and Simulation Insight on the Transport of Silver Fission Product in SiC

Understanding of the fission product transport in TRISO fuel particles is fundamental to improving the safety and performance of high temperature gas cooled reactors. Previous experiments showing silver release from TRISO fuel have focused on release measurements and not direct observation of the fission product transport. The possible diffusion of Ag via a grain boundary diffusion mechanism is being examined. By characterizing the SiC grain boundary structure according the coincidence site lattice scheme and detecting diffusion along specific grain boundaries, insight into the relationship between SiC microstructure and Ag release may be obtained. In addition computer modeling is being used to investigate the diffusion of silver through SiC. We employ a multi-scale approach based on ab initio techniques, molecular dynamics, and continuum rate equations in order to establish relationships between complex microstructures and diffusion rates. Initial work has begun on transport through bulk SiC and on building realistic models of grain boundaries in SiC.Copyright

Tensile and toughness assessment of the procured advanced alloys

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance by 2024 in light water reactor (LWR)-relevant environments

Computational Thermodynamics for Interpreting Oxidation of Structural Materials in Supercritical Water

The Supercritical water-cooled reactor (SCWR) is one of the advanced nuclear reactors being developed to meet the soaring energy demand. The corrosion resistance of structural materials used in the SCWR becomes one of the major concerns as the operation conditions are raised up to ~600°C and ~25 MPa as compared to pressurized water reactors (PWRs) at ~315°C and ~15.5 MPa. Oxidation has been observed as the major corrosion behavior. To mitigate the oxidation corrosion, stabilities of metals and oxides need to be understood with respect to environmental temperature and oxygen partial pressure. Computational thermodynamics provides a practical approach to assess phase stabilities of such multicomponent multi-variable systems. In this study, calculated phase stability diagrams of alloys and corresponding oxides were used to guide the interpretation of oxidation behavior of SCW-exposed structural materials. Examples include ferritic-martensitic steel, austenitic steels and Ni-base alloy, e.g., HCM12A (Fe-11Cr), D9 (Fe-15Cr-15Ni), 800H (Fe-21Cr-32Ni), and 690 (Ni-30Cr-10Fe). Calculated results are in good overall consistence with the experimental data./

Influence of Alloy Microstructure on Oxide Growth in HCM12A in Supercritical Water

HCM12A is a ferritic-martensitic steel alloy envisioned for cladding and structural material in the Generation IV Supercritical Water Reactor (SCWR). This alloy was oxidized in 600oC supercritical water for 2, 4 and 6 weeks, and the oxide layers formed were analyzed using microbeam synchrotron radiation and electron microscopy. The oxide layers show a three-layer structure with an Fe 3O4 outer layer, an inner layer containing a mixture of Fe 3O4 and FeCr 2O4 and a diffusion layer containing FeCr 2O4 and Cr 2O3 precipitates along ferrite lath boundaries. The base metal microstructure has a strong influence on the advancement of the oxide layers, due to the segregation at the lath boundaries of chromium rich particles, which are oxidized preferentially.