Gyeong-Geun Lee

INVESTIGATION ON MATERIAL DEGRADATION OF ALLOY 617 IN HIGH TEMPERATURE IMPURE HELIUM COOLANT

The corrosion of materials exposed to high temperature helium in a very high temperature reactor is caused by interaction with the impurities in the helium. This interaction then induces high temperature mechanical deterioration. By considering the effect of the impurity concentration on material corrosion, a long-term coolant chemistry guideline can be determined for the range of impurity concentration at which the material is stable for a long time. In this work, surface reactions were investigated by analyzing the thermodynamics and the experimental results for Alloy 617 exposed to controlled impure helium at 950℃. Moreover, the surfaces were examined for the Alloy 617 crept in air and in uncontrolled helium, which was explained by possible surface reactions.

MULTISCALE MODELING OF RADIATION EFFECTS ON MATERIALS: PRESSURE VESSEL EMBRITTLEMENT

Radiation effects on materials are inherently multiscale phenomena in view of the fact that various processes spanning a broad range of time and length scales are involved. A multiscale modeling approach to embrittlement of pressure vessel steels is presented here. The approach includes an investigation of the mechanisms of defect accumulation, microstructure evolution and the corresponding effects on mechanical properties. An understanding of these phenomena is required to predict the behavior of structural materials under irradiation. We used molecular dynamics (MD) simulations at an atomic scale to study the evolution of high-energy displacement cascade reactions. The MD simulations yield quantitative information on primary damage. Using a database of displacement cascades generated by the MD simulations, we can estimate the accumulation of defects over diffusional length and time scales by applying kinetic Monte Carlo simulations. The evolution of the local microstructure under irradiation is responsible for changes in the physical and mechanical properties of materials. Mechanical property changes in irradiated materials are modeled by dislocation dynamics simulations, which simulate a collective motion of dislocations that interact with the defects. In this paper, we present a multiscale modeling methodology that describes reactor pressure vessel embrittlement in a light water reactor environment.

A phase-field modeling of void swelling in the Austenitic stainless steel

ABSTRACT Two-dimensional phase-field simulations of void swelling in the Austenitic stainless steel were performed for irradiated materials. A numerical model was established for void swelling with an implementation of the elasticity effect, and we examined the roles of the applied stress and grain boundary sink strength and Frenkel defect recombination in determining the void swelling rate. The obtained results were compared with the existing experimental observations.

Microstructural Investigation of Alloy 617 Creep-Ruptured in Pure Helium Environment at 950℃

The very high temperature gas reactor (VHTR) is one of the next generation nuclear reactors for its safety, long-term stability, and proliferation-resistance. The high operating temperature of over 800 enables various applications with high energy efficiency. Heat is transferred from the primary helium loop to the secondary helium loop through the intermediate heat exchanger (IHX). The IHX material requires creep resistance, oxidation resistance, and corrosion resistance in a helium environment at high operating temperatures. A Ni-based superalloy such as Alloy 617 is considered as a primary candidate material for the intermediate heat exchanger. In this study, the microstructures of Alloy 617 crept in pure helium and air environments at 950 were observed. The rupture time in helium was shorter than that in air under small applied stresses. As the exposure time increased, the thickness of outer oxide layer of the specimens clearly increased but delaminated after a long creep time. The depth of the carbide-depleted zone was rather high in the specimens under high applied stress. The reason was elucidated by the comparison between the ruptured region and grip region of the samples. It is considered that decarburization caused by minor gas impurities in a helium environment caused the reduction in creep rupture time.

RELATIONSHIP BETWEEN RADIATION INDUCTED YIELD STRENGTH INCREMENT AND CHARPY TRANSITION TEMPERATURE SHIFT IN REACTOR PRESSURE VESSEL STEELS OF KOREAN NUCLEAR POWER PLANTS

The decrease in the fracture toughness of ferritic steels in a reactor pressure vessel is an important factor in determining the lifetime of a nuclear power plant. A surveillance program has been in place in Korea since 1979 to assess the structural integrity of RPV steels. In this work, the surveillance data were collected and analyzed statistically in order to derive the empirical relationship between the embrittlement and strengthening of irradiated reactor pressure vessel steels. There was a linear relationship between the yield strength change and the transition temperature shift change at 41 J due to irradiation. The proportional coefficient was about 0.5℃/MPa in the base metals (plate/forgings). The upper shelf energy decrease ratio was non-linearly proportional to the yield strength change, and most of the data lay along the trend curve of the US results. The transition regime temperature interval, ΔTT, was less than the US data. The overall change from irradiation was very similar to the US results. It is expected that the results of this study will be applied to basic research on the multiscale modeling of the irradiation embrittlement of RPV materials in Korea.

Atom probe tomography analysis of radiation-induced solute clustering in austenite stainless steels

ABSTRACT Various types of defects are produced by the irradiation of energetic particles onto a structural material. The large number of mobile vacancies and self-interstitial atoms during irradiation induce defect fluxes and the diffusion of solute atoms in the matrix. The preferential interaction between the solute atoms and radiation-induced defects leads to the enrichment/depletion or clustering of the solutes at defect sinks. In the current work, atom probe tomography (APT) was used for the analysis of radiation-induced solute clustering in ion-irradiated austenite stainless steel 316. Quantitative analysis of the localised clustering of chemical elements was implemented and a parameter selection procedure was proposed. The number density and size distribution of Si clusters in APT specimens irradiated at various temperatures were examined. At high temperature, the number density of the clusters decreased and their size increased. The localized Si atoms in variously shaped defects were clearly identified. The APT method was demonstrated to be suitable for identifying defect structures and for the quantitative analysis of clustering in irradiated specimens.

Microstructure Observation of Inconel Alloy 617 Creep-Tested in He Environment at 950°C

The very high temperature gas reactor (VHTR) has been chosen by the Generation IV International Forum as one of the next-generation nuclear reactors. Due to the high operating temperatures of VHTR, Inconel alloy 617 is being considered as a primary candidate material for the intermediate heat exchanger (IHX) of the VHTR. In this study, the microstructures of creep specimens under various creep loads in a He environment were investigated. As the creep time increased, the thickness of Cr-oxide on the outer layer of the specimens clearly increased, and delaminated after a long creep time. Depths of the decarburized zones in the specimens increased slowly with creep time. However, precipitates at grain boundaries near the surface disappeared before the bulk diffusion of Cr in the matrix. It is considered that decarburization caused by minor gas impurities in He caused the reduction in creep rupture time.Copyright