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Leonard, Keith

Publications and source records attributed to Leonard, Keith.

10 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

A microscopic and crystallographic study of proton irradiated alloy 718

Solution annealing and age hardening are important processes for achieving good engineering and chemical properties of alloy 718. The composition of alloy 718 also plays an important role as it can affect both mechanical (γ”- and γ’-phase formation) and irradiation behaviors. Therefore, in this study five different sets of alloy 718 samples with two different chemical compositions were fabricated using different processing conditions and irradiated up to 4 dpa. Based on the effects of irradiation on the microstructural data of γ”- and γ’-phases, irradiation-induced dislocation statistics, and crystallographic data, it can be inferred that the samples processed at high solution annealing temperature (1093°C) behave better under irradiation compared to the samples processed at low solution annealing temperatures (945 and 1065°C).

36 MATERIALS SCIENCE↗
DOE OSTI2019

FRICTION STIR WELDING AND PRELIMINARY CHARACTERIZATION OF IRRADIATED 304 STAINLESS STEEL

The mitigation of helium induced cracking in the heat affected zone (HAZ), a transition metallurgical zone between the weld zone and base metal, during repair welding is a great challenge in nuclear industry. Successful traditional fusion welding repairs are limited to metals with a maximum of a couple of atomic parts per million (appm) helium, and structural materials helium levels in operating nuclear power plants are generally exceed a couple of appm after years of operations. Therefore, fusion welding is very limited in nuclear power plants structural materials repairing. Friction stir welding (FSW) is a solid-state joining technology that reduces the drivers (temperature and tensile residual stress) for helium-induced cracking. This paper will detail initial procedural development of FSW weld trials on irradiated 304L stainless steel (304L SS) coupons utilizing a unique welding facility located at one of Oak Ridge National Laboratory’s hot cell facilities. The successful early results of FSW of an irradiated 304L SS coupon containing high helium are discussed. Helium induced cracking was not observed by scanning electron microscopy in the friction stir weld zone and the metallurgical zones between the weld zone and base metal, i.e. thermal mechanical affected zone (TMAZ) and HAZ. Characterization of the weld, TMAZ and HAZ regions are detailed in this paper.

Tang, Wei↗
DOE OSTI2018

ADVANCED WELDING TECHNOLOGY DEVELOPMENT FOR NUCLEAR REACTOR REPAIR

Welding is widely used for repair, maintenance, and upgrade of nuclear reactor components. As a critical technology for supporting the extension of nuclear power plant service lifetimes beyond 60 years, there has been an industry need in further developing welding technology for highly irradiated materials. During welding of irradiated materials, Helium, which is a transmutation byproduct from Boron and Nickel contained in the structural alloys, can coalesce into bubbles along grain boundaries in the material under driving forces of high temperature and thermal tensile stress. This leads to embrittlement and potential intergranular cracking in the Heat Affected Zone (HAZ) of the weld. In this paper, irradiated 304 stainless steel coupons containing 5, 10 and 20 ppm Boron prior to radiation were successfully welded by advanced laser beam welding (LBW) and friction stir welding (FSW). Facilities and welding machine design, welding processes development, cold material welded joints evaluation, process safety control and documentation, and irradiated material welding are discussed. Both laser and friction stir welded coupons of the irradiated material exhibited high welding quality and surface finish. No Helium induced welding defects were observed on the weld surfaces or adjacent base metals.

Tang, Wei↗
DOE OSTI2018

Characterization of different forms of Zr-2.5Nb samples before and after neutron irradiation

Deleterious effects of hydride precipitation on the mechanical properties of Zr-based alloy systems have been studied for many years. Most of these studies have been conducted at high-temperature irradiation conditions that are compatible with those in commercial reactor systems, such as light water and pressurized water reactors. Evaluations of hydride effects on the mechanical properties of welded Zr-2.5Nb are sparse among these reported studies. Research on the mechanical properties of Zr-based alloys neutron irradiated at low temperatures (<100 °C) is also very rare in the literature. Since some companies are interested in using Zr-based alloys as structural materials and in other parts such as pressure tubes in small reactor systems, low-temperature irradiation studies of these materials are important. The current work presents an evaluation of the tensile properties of Zr-2.5Nb welded using tungsten inert gas and electron beam welding as a function of hydrogen charging and neutron irradiation at low temperatures (60–100 °C).

36 MATERIALS SCIENCE↗
DOE OSTI2017

Plastic Deformation Processes Accompanying Stress Corrosion Crack Propagation in Irradiated Austenitic Steels

During stress corrosion crack propagation, stress values in the crack tip vicinity often exceed material yield stress limit. Plastic deformation processes may accompany and influence cracking. Here, stress corrosion crack propagation and deformation mechanisms were investigated using EBSD analysis. The investigated material was 304L Ti-enriched austenitic stainless steel irradiated to 10.4 dpa at 320°C in the BOR-60 fast reactor. Crack growth tests were conducted in a simulated Normal Water Chemistry (NWC) environment in the temperature range 288–320 °C using compact tension specimens. By analyzing crack trajectory and grain structure in the crack vicinity, it was established that grain orientation with respect to the acting stress direction was not a key factor controlling crack propagation. No crystallographic orientation susceptible to cracking was identified. Also, EBSD analysis revealed strong inhomogeneity in plastic strain distribution along the crack path. Most crack-adjacent grains remained virtually strain-free whereas few grains experienced strong plastic strain. These areas were presumed to be “plastic bridges” or “ductile ligaments.”

Gussev, Maxim N↗
DOE OSTI2017

Characterization of alloy 718 subjected to different thermomechanical treatments

Chemical phase and microstructural investigations of alloy 718 solution-annealed and age-hardened were performed in this study. We focused on the effects of solution annealing temperature, aging temperature and holding time, and the amount of intermediate cold work on the alloy. We also studied the formation of secondary phases such as γ’-phase, γ”-phase, and δ-phase, grain sizes, and any deformations of the microstructure with respect to the processed conditions. Statistics such as size and number densities of these precipitates with respect to the processing conditions were evaluated and a discussion on optimum conditions in obtaining finer and higher density of γ’- and γ”-phase precipitates is also presented.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Mechanical and Microstructural Characterization of Low-Temperature Neutron Irradiated Zircaloy-4

Zr-based alloys, namely Zircaloy-2, Zircaloy-4, ZIRLO, or M5, have been used in light water reactors (LWRs) as structural materials (cladding, pressure tubes, and etc.). Under irradiation conditions, especially at harsh pressurized water reactor (PWR) conditions, these Zr alloys can undergo embrittlement due to corrosion, hydriding, and irradiation induced stresses. This ultimately leads to the severe degradation of Zr-alloys and increase the probability of fuel failure. While various research efforts have been conducted in developing better Zr-based alloys to withstand these harsh environments, there is a considerable amount of Zr-alloy characteristics, such as change in mechanical properties under irradiation conditions and its relationship to the materials microstructure, that need to be studied. This present study has focused on investigation of mechanical and microstructural properties of Zircaloy-4 under low temperature (80 - 100 deg. C) neutron irradiation conditions. Some of the samples studied were also hydrided to evaluate influence of hydride precipitates in the material. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2015

Light Water Reactor Sustainability Program Integrated Program Plan

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 60%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to grow by about 24% from 2013 to 2040 . At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license, for a total of 60 years of operation (the oldest commercial plants in the United States reached their 40th anniversary in 2009). Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity for 40- and 60-year license periods. If current operating nuclear power plants do not operate beyond 60 years (and new nuclear plants are not built quickly enough to replace them), the total fraction of generated electrical energy from nuclear power will rapidly decline. That decline will be accelerated if plants are shut down before 60 years of operation. Decisions on extended operation ultimately rely on economic factors; however, economics can often be improved through technical advancements. The U.S. Department of Energy Office of Nuclear Energy’s 2010 Research and Development Roadmap (2010 Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: 1. Develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. 2. Develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals. 3. Develop sustainable nuclear fuel cycles. 4. Understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗