Engineering topics

Glazoff, Michael

Publications and source records attributed to Glazoff, Michael.

6 records

DOE OSTI2024

Characterization of Precipitation-Strengthening Heat-Resistant Austenitic Stainless Steels for Life-Prediction Modeling

In this study, the role of minor alloying additions in 347H stainless steels (UNS34709, ASTM A240/240M) on creep-rupture properties at 650-750°C and microstructure evolution during isothermal exposure at 750°C has been investigated, aiming to provide the experimental dataset as boundary conditions of physics-based modeling for material/component life prediction. Four different 347H heats containing various amounts of boron and nitrogen additions were prepared and evaluated. The combined additions of B and N are found to stabilize the strengthening secondary M23C6 carbides and retarding the transition from M23C6 to sigma phase precipitates during thermal exposure. The observed kinetics of microstructure evolution reasonably explains the improvement of creep-rupture properties of 347H stainless steels with the B and N additions.

Yamamoto, Yukinori↗

DOE OSTI2019

Characterization of Creep-Fatigue Crack Growth in Alloy 709 and Prediction of Service Lives in Nuclear Reactor Components (Final Report)

This final report summarizes the activities and results that were obtained by the research team working on the project “Characterization of Creep-Fatigue Crack Growth in Alloy 709 and Prediction of Service Lives in Nuclear Reactor Components”, DOE-NEUP Project 15-8623. This project was mainly concerned with the creep-fatigue crack growth characterization of austenitic stainless-steel Fe-20Cr-25Ni (Alloy 709). For this purpose, an array of experimental and computational activities was performed. The main goal of these activities was to describe and predict the crack growth behavior in Alloy 709 under different conditions of applied loading and temperature. Alloy 709 specimens were tested to describe both the uniaxial creep behavior and creep-fatigue crack growth rates at temperatures between 550°C and 700°C. Microstructure and fracture surface analysis was performed using several experimental techniques, and the defining features of crack growth propagation were identified and described. Thermodynamic and diffusion simulations were performed to understand the microstructure evolution of Alloy 709 at elevated temperatures, and to establish an aging schedule relevant for in-service conditions over extended periods. Experimental testing was performed on both as-received and aged Alloy 709 specimens. Two computational methods were developed to predict crack growth rates and remaining lives in specimens loaded under either fatigue of creep-fatigue. The first computational technique was the finite element method, while the second was strip-yield modeling. Both methods relied on the prediction of crack growth rates using the concept of plasticity-induced crack closure. Crack-tip opening loads were computed and used to perform crack growth rate predictions. The computational predictions matched well with the experimental results. The strip-yield model was shown to be an excellent predictive tool not only for crack growth in austenitic SS Alloy 709, but also in 316 SS, martensitic Cr-Mo steels, Ni-based alloys and Al-based alloys. The main conclusion of this study is that Alloy 709 exhibits excellent creep crack growth resistance at temperatures between 550°C and 700°C. In fact, the crack growth resistance of this alloy appears to be higher than that of 316 SS in similar testing conditions. Moreover, when creep loading with hold (dwell) times of 60s or 600s were intercalated every fatigue cycle, the crack growth rates did not change significantly. The conclusion is that the tests performed revealed that the crack growth is insignificant during the creep portion of a loading cycle, as compared with the crack growth during the fatigue portion of the cycle. In general, Alloy 709 shows excellent resistance to creep deformations and creep crack growth up to 700°C.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Effect of heat treatment on microstructure and hardness of Grade 91 steel

The modified 9Cr-1Mo steel (Grade 91) is a material of choice in fossil-fuel-fired power plants with increased efficiency, service life, and reduction in emission of greenhouse gases. It is also considered a prospective material for the Next Generation Nuclear Power Plant for application in reactor pressure vessels at temperatures up to 650°C. In this paper, heat treatment of the modified 9Cr-1Mo steel was studied by normalizing and tempering the steel at various temperatures and times, with the ultimate goal of improving its creep resistance and optimizing material hardness. The microstructural evolution of the heat treated steels was correlated with the differential scanning calorimetric results. Optical microscopy, scanning and transmission electron microscopy in conjunction with microhardness profiles and calorimetric plots were used to understand the evolution of microstructure including precipitate structures in modified 9Cr-1Mo steel and relate it to the mechanical behavior of the steel. Thermo-CalcTM calculations were used to support experimental work and provide guidance in terms of the precipitate stability and microstructural evolution. Furthermore, the carbon isopleth and temperature dependencies of the volume fraction of different precipitates were constructed. The predicted and experimentally observed results were found to be in good agreement.

20 FOSSIL-FUELED POWER PLANTS↗

DOE OSTI2014

Advanced heat exchanger development for molten salts

This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, which show corrosion resistance to molten salt at nominal operating temperatures up to 700°C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet material in Hastelloy N were corrosion tested in?58 mol% KF and 42 mol% ZrF4 at 650, 700, and 850°C for 200, 500, and 1,000 hours. Corrosion rates found were similar between welded and nonwelded materials, typically <10 mils per year. For materials of construction, nickel and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of contaminant type and alloy composition with respect to chromium and carbon to better define the optimal chromium and carbon composition, independent of galvanic or differential solubility effects. Also presented is the division of the nuclear reactor and high temperature components per ASME standards, along with design requirements for a subcritical Rankine power cycle heat exchanger that has to overcome pressure difference of about 17 MPa.

42 ENGINEERING↗

DOE OSTI2014

Draft of M2 Report on Integration of the Hybrid Hydride Model into INL's MBM Framework for Review

This report documents the development, demonstration and validation of a mesoscale, microstructural evolution model for simulation of zirconium hydride {delta}-ZrH{sub 1.5} precipitation in the cladding of used nuclear fuels that may occur during long-term dry storage. While the Zr-based claddings are manufactured free of any hydrogen, they absorb hydrogen during service, in the reactor by a process commonly termed ‘hydrogen pick-up’. The precipitation and growth of zirconium hydrides during dry storage is one of the most likely fuel rod integrity failure mechanisms either by embrittlement or delayed hydride cracking of the cladding. While the phenomenon is well documented and identified as a potential key failure mechanism during long-term dry storage (NUREG/CR-7116), the ability to actually predict the formation of hydrides is poor. The model being documented in this work is a computational capability for the prediction of hydride formation in different claddings of used nuclear fuels. This work supports the Used Fuel Disposition Research and Development Campaign in assessing the structural engineering performance of the cladding during and after long-term dry storage. This document demonstrates a basic hydride precipitation model that is built on a recently developed hybrid Potts-phase field model that combines elements of Potts-Monte Carlo and the phase-field models. The model capabilities are demonstrated along with the incorporation of the starting microstructure, thermodynamics of the Zr-H system and the hydride formation mechanism.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

MorphoHawk

An overview of Idaho National Laboratory's MorphoHawk software.

97 MATHEMATICS AND COMPUTING↗