Kun Mo

High Temperature Aging and Corrosion Study on Alloy 617 and Alloy 230

The very high temperature gas-cooled reactor (VHTR), with dual capacities of highly efficient electricity generation and thermochemical production of hydrogen, is considered as one of the most promising Gen-IV nuclear systems. The primary candidate materials for construction of the intermediate heat exchanger (IHX) for the VHTR are alloy 617 and alloy 230. To have a better understanding of the degradation process during high temperature long-term service and to provide practical data for the engineering design of the IHX, aging experiments were performed on alloy 617 and alloy 230 at 900°C and 1000°C. Mechanical properties (hardness and tensile strength) and microstructure were analyzed on post-aging samples after different aging periods (up to 3000 h). Both alloys attained increased hardness during the early stages of aging and dramatically soften after extended aging times. Microstructural analysis including transmission electron microscopy, scanning electron microscopy, energy dispersive X-ray spectroscopy, and electron backscatter diffraction was carried out to investigate the microstructure evolution during aging. A carbide particle precipitation, growth, and maturing process was observed for both alloys, which corresponds to the changes of the materials’ mechanical properties. Few changes in grain boundary character distribution and grain size distribution were observed after aging. In addition, high temperature corrosion studies were performed at 900°C and 1000°C for both alloys. Alloy 230 exhibits much better corrosion resistance at elevated temperature compared with alloy 617.

Investigation of High-Energy Ion-Irradiated MA957 Using Synchrotron Radiation under In-Situ Tension

In this study, an MA957 oxide dispersion-strengthened (ODS) alloy was irradiated with high-energy ions in the Argonne Tandem Linac Accelerator System. Fe ions at an energy of 84 MeV bombarded MA957 tensile specimens, creating a damage region ~7.5 μm in depth; the peak damage (~40 dpa) was estimated to be at ~7 μm from the surface. Following the irradiation, in-situ high-energy X-ray diffraction measurements were performed at the Advanced Photon Source in order to study the dynamic deformation behavior of the specimens after ion irradiation damage. In-situ X-ray measurements taken during tensile testing of the ion-irradiated MA957 revealed a difference in loading behavior between the irradiated and un-irradiated regions of the specimen. At equivalent applied stresses, lower lattice strains were found in the radiation-damaged region than those in the un-irradiated region. This might be associated with a higher level of Type II stresses as a result of radiation hardening. The study has demonstrated the feasibility of combining high-energy ion radiation and high-energy synchrotron X-ray diffraction to study materials’ radiation damage in a dynamic manner.

Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions need to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions and therefore evaluate the qualification of U3Si2 as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U3Si2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U3Si2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

Strain-Rate Sensitivity Analysis for Ni Alloys to be Used in Very High Temperature Reactors

Both Alloy 617 and Alloy 230 have been considered the most promising structural materials for the Very High Temperature Reactor (VHTR). In this study, mechanical properties of both alloys were examined by performing tensile tests at three different strain rates and at temperatures up to 1000°C. This range covers time-dependent (plasticity) to time-independent (creep) deformations. Strain-rate sensitivity analysis for each alloy was conducted in order to approximate the long-term flow stresses. The strain-rate sensitivities for the 0.2% flow stress were found to be temperature independent (m [approximate] 0) at temperatures ranging from room temperature to 700°C due to dynamic strain aging. At elevated temperatures (800°C to 1000°C), the strain-rate sensitivity significantly increased (m > 0.1). Compared to Alloy 617, Alloy 230 displayed higher strain-rate sensitivities at high temperatures. This leads to lower estimated long-term flow stresses. Results of this analysis were used to evaluate the current American Society of Mechanical Engineers (ASME) allowable design limits for each alloy. The study showed that the allowable design stresses in the ASME Boiler and Pressure Vessel Code for either alloy do not provide adequate long-term degradation estimation. Nevertheless, rupture stresses for Alloy 617, developed in the ASME code case N-47-28, can generally satisfy the safety margins at 800°C and 1000°C estimated in the study following the strain-rate sensitivity analysis. Furthermore, additional material development studies might be required, since the design parameters for rupture stresses are constrained such that the current VHTR conceptual designs cannot satisfy the material limits.

FASTGRASS implementation in BISON and Fission gas behavior characterization in UO 2 and connection to validating MARMOT

This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL). Two major accomplishments in FY 15 are summarized in this report: (1) implementation of the FASTGRASS module in the BISON code; and (2) a Xe implantation experiment for large-grained UO2. Both BISON AND MARMOT codes have been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. To contribute to the development of the Moose-Bison-Marmot (MBM) code suite, we have implemented the FASTGRASS fission gas model as a module in the BISON code. Based on rate theory formulations, the coupled FASTGRASS module in BISON is capable of modeling LWR oxide fuel fission gas behavior and fission gas release. In addition, we conducted a Xe implantation experiment at the Argonne Tandem Linac Accelerator System (ATLAS) in order to produce the needed UO2 samples with desired bubble morphology. With these samples, further experiments to study the fission gas diffusivity are planned to provide validation data for the Fission Gas Release Model in MARMOT codes.

Synchrotron radiation study on alloy 617 and alloy 230 for VHTR application

High-energy synchrotron radiation has proven to be a powerful technique for investigating fundamental deformation processes for various materials, particularly metals and alloys. In this study, high-energy synchrotron X-ray diffraction (XRD) was used to evaluate Alloy 617 and Alloy 230, both of which are top candidate structural materials for the Very-High-Temperature Reactor (VHTR). Uniaxial tensile experiments using in-situ high-energy X-ray exposure showed the substantial advantages of this synchrotron technique. First, the small volume fractions of carbides, e.g. ∼6% of M6 C in Alloy 230, which are difficult to observe using lab-based X-ray machines or neutron scattering facilities, were successfully examined using high-energy X-ray diffraction. Second, the loading processes of the austenitic matrix and carbides were separately studied by analyzing their respective lattice strain evolutions. In the present study, the focus was placed on Alloy 230. Although the Bragg reflections from the γ matrix behave differently, the lattice strain measured from these reflections responds linearly to external applied stress. In contrast, the lattice strain evolution for carbides is more complicated. During the transition from the elastic to the plastic regime, carbide particles experience a dramatic loading process, and their internal stress rapidly reaches the maximum value that can be withstood. The internal stress for the particles then decreases slowly with increasing applied stress. This indicates a continued particle fracture process during plastic deformations of the γ matrix. The study showed that high-energy synchrotron X-ray radiation, as a non-destructive technique for in-situ measurement, can be applied to ongoing material research for nuclear applications.Copyright

Microstructural evolution of Alloy 617 and Alloy 230 following high temperature aging

Alloy 617 and Alloy 230 are solid-solution strengthened nickel based superalloys, which have been considered two of the most promising structural materials for the Very-High-Temperature Reactor (VHTR). In order to have a better understanding of the degradation process of the materials in the VHTR, long-term aging experiments have been carried out to investigate the dynamic process of microstructure evolution at 900 and 1000°C for Alloy 617 and Alloy 230. The microstructural evolution process in different aging periods (up to 3000 hours) was analyzed by Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM) and Electron Backscatter Diffraction (EBSD). A diffusion-controlled precipitation and coarsening of carbide particles (mainly M23 C6 and M6 C) for both alloys was observed. The corresponding characteristics of the precipitates, i.e. type, size and coherence, were analyzed. The coarsening rate of the intergranular precipitates in Alloy 617 was found to be much faster compared to Alloy 230’s. The inhomogeneous precipitation process in the transverse plane of Alloy 617 was observed, which may be attributed to the alignment of the inclusion particles induced by the hot rolling. Hardness and tensile tests were carried out to investigate the aging impacts on materials’ strength. Both alloys obtained increased hardness and strength during early stages of aging and softened after elongated time. The results of mechanical tests were in a good agreement with the microstructure evolution process.Copyright