Engineering topics

Miller, Roger G.

Publications and source records attributed to Miller, Roger G..

13 records

DOE OSTI2024

Study of Mechanical Properties, Microstructure, and Residual Stresses of AISI 304/304L Stainless Steel Submerged Arc Weld for Spent Fuel Dry Storage Systems

The confinement boundaries of spent nuclear fuel (SNF) canisters are typically fusion welded. Welded microstructures, strain hardening, and residual stresses combined with a chemically aggressive, chloride-rich environment led to concerns that the welded canister may be susceptible to chloride-induced stress corrosion cracking (CISCC). A comprehensive understanding of the modification of stainless steel (SS) metallurgical and mechanical properties by fusion welding could accelerate the predictive analysis of CISCC susceptibility. This paper describes a submerged arc welding (SAW) procedure that was developed and qualified on 12.7 mm (0.5 in.) thick AISI 304/304L SS to produce joints in a way similar to actual SNF canister manufacturing. This procedure has the potential to reduce the production cost and weld CISCC susceptibility by using fewer welding passes and lower heat input than current industrial applications. Global and local mechanical behaviors and properties, as well as residual stress distributions on the welded joint, were studied. The results indicate that hardness values in the fusion zone (FZ) and heat-affected zone (HAZ) are slightly higher than that of the base metal. Strain localization was presented in the HAZ before the tensile stress reached its maximum value, and then it shifted to the FZ. The specimen finally broke in the FZ. High tensile residual stresses exhibited in the FZ and the nearby HAZ suggest the highest CISCC-susceptible spots. The maximum tensile residual stresses were along the welding direction, indicating that if cracks occur, they would be perpendicular to the welding direction. This study involved developing and qualifying a SAW procedure for SNF canister production. The new procedure yielded cost savings (SAW working efficiency increased by about 80%), improved mechanical properties, and presented moderate residual stresses. Analysis revealed that the welded joint’s low-stress and high-stress damage assessments may be affected by shifts in the strain localization spot under loading.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Complete the first phase of the comprehensive characterization of repair welding performed on irradiated Ni alloy 182

This report describes the research activities on post-weld evaluation and characterization of the quality and properties of welds made on the first phase of repair welding campaign on irradiated Ni alloy 182. The equipment and capabilities of the campaign were developed jointly by the U.S. Department of Energy, Office of Nuclear Energy, Light Water Reactor Sustainability Program, the Electric Power Research Institute, Long Term Operations Program (and the Welding and Repair Technology Center), and Oak Ridge National Laboratory. Irradiated nickel alloy 182 with doped boron ranging from 5 wppm to 15 wppm, were laser welded in the hot cell. The weld samples were cross-sectioned and characterized. The significant, on-going effort to weld irradiated alloys with high helium concentrations and comprehensively analyze the results will eventually yield validated repair techniques and guidelines for use by the nuclear industry in extending the operational lifetimes of nuclear power plants.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Collaborative Study on Weldability of Irradiated Stainless Steel 304 Materials Removed from The NRU Reactor

A collaborative research on developing advanced welding technologies for irradiated stainless steel 304 materials between CNL and ORNL has been established to support both U.S. and Canadian interests in evaluation of weld repair techniques on irradiated materials to support continuous operation of commercial nuclear power. The work utilizes unique material from the National Research Universal (NRU) reactor and the specialized welding hot cell facility at ORNL. The objective is to explore suitable welding technique and parameters and to determine the helium concentration limitation in terms of irradiated stainless-steel weldability. An additional objective is to develop the knowledge and understanding of cracking mechanisms induced by high concentration of helium in the process of weld repair of the irradiated structural alloys. This report summarizes the experimental welding evaluation on irradiated material from the NRU reactor containing helium concentrations ranging from 12 to 45 atomic parts per million (appm).

36 MATERIALS SCIENCE↗

DOE OSTI2023

A Technique for the Quantitative Characterization of Weld Microstructure and Application to Mo Welds

The choice of weld parameters determines the size, shape, and curvature of grains in the fusion zone (FZ) and heat-affected zone (HAZ) of welds while the mechanical properties of the welds are correlated to this microstructure. Here, a new technique to quantitatively evaluate these microstructural characteristics in both zones of welds has been applied to molybdenum gas tungsten arc welds fabricated using different weld parameters. Trends in microstructural characteristics in the FZ and HAZ were evaluated and correlated with changes to heat input, weld speed, and weld technique. The use of this approach showed that a 20 pct decrease in heat input caused a 20 pct decrease in the number of FZ grains with aspect ratios ≥ 4. The orientations of the FZ grain segments as a function of distance from the FZ centerline were significantly affected by the weld speed and its effect on weld pool shape. A 50 pct increase in weld speed caused a 20 pct decrease in grain segments orientated 60 to 90 deg from the normal to the direction of welding. This technique also captured differences in grain sizes and grain size anisotropy in the FZ between welds made with a constant current, pulsed current, and use of a 4-pole-magnetic oscillator.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Weldability of irradiated Stainless Steel 304 materials harvested from the National Research Universal (NRU) reactor

In this work, the weldability of the neutron-irradiated Stainless Steel 304 (SS 304) materials containing estimated helium levels of 12 to 44 appm has been studied through a Collaborative Research and Development Agreement between the Canadian Nuclear Laboratories and Oak Ridge National Laboratory. The work utilizes a SS 304 component harvested from the National Research Universal reactor. Laser beam welding was used to explore suitable welding conditions and associated parameters, to determine helium concentration limits for successful welding of irradiated stainless-steels. The experimental results show that maximum fusion zone void sizes, fusion zone void quantity and total heat affected zone helium-induced crack length generally increased with increasing He levels and effective weld heat input. A proper combination of welding parameters with low effective heat input, such as, laser power, weld travel speed and wire feed speed, were shown to improve the welding quality, and potentially reduce the length and probability of the formation of grain boundary helium-induced cracking and voids of irradiated SS 304 even with He concentrations as high as about 44 appm.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Light Water Reactor Sustainability Program: Complete the weld campaign on Ni-base irradiated materials using stress improved laser welding including the preliminary weld quality inspections

This report summarizes the most recent welding campaign on irradiated Ni-base alloy 182 and the preliminary weld quality inspections at the Radiochemical Engineering Development Center (REDC). Equipment and capabilities were developed jointly by the U.S. Department of Energy, Office of Nuclear Energy, Light Water Reactor Sustainability Program, the Electric Power Research Institute, Long Term Operations Program (and the Welding and Repair Technology Center), and Oak Ridge National Laboratory. Irradiated nickel alloy 182, with target helium contents of 5 atom parts-per million (appm), 10 appm, and 20 appm, were laser welded in the hot cell successfully. The significant, on-going effort to weld irradiated alloys with high helium concentrations and comprehensively analyze the results will eventually yield validated repair techniques and guidelines for use by the nuclear industry in extending the operational lifetimes of nuclear power plants. This report fulfills the FY 2023 milestone M3LW-22OR0406013, “Complete the weld campaign on Ni base irradiated materials using stress improved laser welding including the preliminary weld quality inspections”.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Cast Components for High Temperature Concentrated Solar Power Thermal Systems

Concentrating Solar-Thermal Power (CSP) components such as piping, valves fittings are required for use at temperatures up to 800°C. These are anticipated to be made using Nickel-based alloys such as Haynes ® 230 ® , Haynes ® 282 ® , or IN ® 740H ® and can contribute significantly to the cost of a CSP Gen 3 plant. Thus, there is a significant motivation to lower the cost of materials and components so that the capital costs can be minimized. In all these cases, the cost of the component has two contributions: 1. Materials cost, and 2. Manufacturing cost. Both materials cost and manufacturing costs must be kept low to attain the lowest possible cost. Materials cost can be lowered by using materials that have the ideal combination of properties at the lowest cost. Another avenue to lower cost is chose a manufacturing process that has the potential to lower the cost. Traditionally tubes are made from billets through a wrought process and can be expensive. An alternative process to consider is the centrifugal casting process where the tube is directly fabricated from molten metal. In this case, the molten metal is poured inside a cylindrical metallic mold with an insulating layer and is spun rapidly. A wide range of sizes (diameter, wall thickness, and length) can be cast using this process. The objective of this project was to develop the process for fabricating pipes and related components using a centrifugal casting process and to measure the properties of alloys fabricated using this process.

14 SOLAR ENERGY↗

DOE OSTI2021

Conduct weld campaign (FY-21-1) on irradiated materials provided by the Canadian Nuclear Laboratory (CNL), including baseline post-weld evaluation and testing

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Suppression of Helium Induced Cracking in Laser Repair Welding of Highly Irradiated Stainless Steels

Welding repair of irradiated nuclear reactor materials such as austenitic stainless steels used for the reactor internals has been challenging due to irradiation generated helium in the material matrix over an extended period of time. In this study, improved laser welding techniques, including the novel Auxiliary Beam Stress Improved (ABSI) technique that proactively manages the stresses during laser repair welding is investigated to substantially reduce or avoid the occurrence of intergranular helium-induced cracking (HeIC). Here, laser weld cladding, with and without ABSI technique, was applied to deposit weld metal on irradiated stainless steel (Type 304 L) blocks with various helium concentrations up to approximately 20 atomic parts per million (appm). Welded samples were cross-sectioned and analyzed using optical and electron microscopes. It was found that, with the improved laser welding techniques, no macroscopic (millimeter-sized) cracks were observed in any of the samples. Microscopic (less than 50 μm) cracks were present in the heat affected zone. Statistical analysis on the sample with the highest helium concentration (~20 appm) reveals that the ABSI technique could effectively reduce the size and amounts of microscopic cracks.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Produce ZrH Moderator Material

This reports complete Milestone M2.2.3 Produce zirconium hydride moderator material. In this milestone, we report the successful fabrication of FeCrAl alloys cladded zirconium hydride moderator. Different types of crucible designs were presented, including the basic design with bar cladding material, basic design with tube cladding material, and crucible design with position holding function. Mo and FeCrAl were used as the trial cladding materials. The welding techniques and procedures were reported and discussed. Totally three welding techniques were adopted in the development of moderator cladding, and they are electron beam welding (EBW), laser welding (LW), and gas tungsten arc welding (GTAW). After welding, all cladded crucibles were evaluated with two kinds of leak testing, the helium leak test for minor leakage and the bubble test for major leakage. Due to the concern of the poor neutronics performance, Mo was not ideal for future moderator cladding application. Therefore, detailed characterization of the Mo cladded zirconium hydride was not pursued. Instead, the characterization of FeCrAl cladded zirconium hydride was performed. We discussed two material conditions (i.e., as machined and pre-oxidized conditions) prior to the cladding process, evaluated thermal stability of cladded zirconium hydride moderator through directly measuring hydrogen release, and characterized the zirconium hydride following the thermal desorption measurement. The results showed pre-oxidized FeCrAl is capable of efficiently preventing hydrogen release from the moderator assembly.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Neutron diffraction illustrates residual stress behavior of welded alloys used as radioactive confinement boundary

Corrosion-resistant welded alloys are frequently used as a leak-tight boundary in critical applications that require confinement of hazardous and/or radioactive substances, including an increasing population of spent nuclear fuel (SNF) canisters. The behavior of residual stresses generated as a result of irregular elastic–plastic deformation during processes such as welding is one of today's key issues to a full understanding of the aging mechanisms that may compromise the confinement boundary. Whether such processes and any subsequent weld repairs, not subjected to post-weld heat treatment, would negatively affect the initial material by introducing through-thickness tensile stresses remains an open question. Here we report the first residual stress measurements using neutron diffraction on the welded joints of a SNF canister. We found significant tensile residual stresses in the as welded sample, indicating that initiation and through-thickness growth of cracks may be possible. Following repair, we observed a stress redistribution and introduction of beneficial compressive stresses. We anticipate our results will improve understanding of confinement susceptibility to aging and guide improvements in repair techniques.

36 MATERIALS SCIENCE↗

DOE OSTI2020

A Versatile Remediation Module for Remote Repair of Spent Nuclear Fuel and High-Level Waste Storage Containers

Oak Ridge National Laboratory (ORNL) successfully demonstrated the Versatile Remediation Module (VRM), a prototype module designed and built by ORNL for on-site remote repair of welded stainless steel storage containers for spent nuclear fuel and high-level radioactive waste. This paper describes the VRM prototype and its design features and components to support continued long-term storage or off-site transportation of spent nuclear fuel and high-level radioactive waste currently stored in storage containers. A remote (100 ft away from the simulated radiative environment) demonstration of the VRM was successfully performed on a full-scale mock-up welded stainless steel canister. The VRM is designed with features to accommodate remediation techniques beyond those currently selected and described in this paper. Therefore, many of the VRM’s features may benefit other remote nuclear or nonnuclear applications. The VRM is envisioned to serve as a development center to facilitate and enhance further development of new remediation technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Oak Ridge National Laboratory Technical Input for the Nuclear Regulatory Commission Review of the 2017 Edition of ASME Section III, Division 5, ‘High Temperature Reactors’

To assist the Nuclear Regulatory Commission in its decision making on endorsement of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code Section III, Division 5 (2017 Edition) for development of advanced non-light water reactors, the following Division 5 portions were reviewed: Article HBB-2000 Material; Article HCB-2000 Material; Article HGB-2000 Material; Mandatory Appendix HBB-I-14 Tables and Figures; and, Nonmandatory Appendix HBB-U Guidelines for Restricted Material Specifications to Improve Performance in Certain Service Applications. In addition to the 2017 Edition, the same parts of the 2019 Edition have also been reviewed as indicated in various sections of the report. This review was conducted by a collaboration of national laboratory and private sector participants with significant industrial experience, including some heavy lifting and deep diving from Clarus Consulting, LLC., all intended to achieve an objective, independent, and practical perspective. The report provides recommendations, descriptions of the evaluation methods, and the source references for the data used. To build confidence required for endorsement of the Code, this review was conducted as a verification and validation of the above Code contents. The objective of verification is to ensure that the Code is free of error – direct or implied; contains the information needed for its use, including proper coverage of the Code-specified materials for the intended application, and completeness and adequacy of references to other portions of the Code. The objective of validation is to authenticate that the Code tabulations and graphs represent design inputs consistent with what are determined using rules and methods specified by the Code. The authentication process used data that were assembled and/or generated independent of Code development, while the methods of analysis followed Code-specified methods where appropriate. The designated portions for this review cover the five alloys codified for high temperature reactor applications in Division 5, i.e. 316 SS, 304 SS, 800H, 2¼Cr-1Mo, and 9Cr-1Mo-V, regarding their general requirements, permitted specifications and design stress intensity values for pressure-retaining applications, deterioration in service, fatigue acceptance test, permissible weld materials, tensile and yield strength, expected minimum stress-to-rupture values (including for Alloy 718), weld stress rupture factors, permissible materials for bolting use, and restricted specifications in certain service applications. Additionally, stress intensity values for bolting materials including 316 SS, 304 SS and alloy 718 were reviewed. Analysis and discussion are also provided on contents outside of these designated Code portions where it was deemed relevant and necessary to develop a technically sound understanding of issues relating to the designated portions. Due to unavailability of sufficient test data on welds during the review period, the weld stress rupture factors in Tables HBB-I-10.14A to E, which cover a total of ten tables for the five alloys welded with twenty-eight different weld metals (some with similar properties), have been deferred to a future review effort. The review identified mainly two types of issues. The first type includes instances where the Code is found factually incomplete or incorrect, such as obsolete materials specifications listings, missing tabulation of stresses for bolting. Changes to the Code are recommended in these cases. The second type of issue includes instances where the Code tabulations and graphs are found to be less conservative than the review analysis results. In these cases, recommendations are made for further review and consideration where the difference in conservatism exceeds 10%, which is our threshold for questioning technical adequacy, meriting a risk assessment by the Nuclear Regulatory Commission and/or reactor designers. It is noted that this effort has been executed using all available data and established methods of analysis, including methods and criteria specified and used by the Code. As such, the findings that are presented in quantitative detail, in a format for convenient comparison with the Code, and with identification of where further review is recommended, should provide a sound technical basis for decisions about quantifying the implications of the reduced design margins and technical adequacy/inadequacy to form a basis for conditioning specific Code tabulation values on endorsement. Recommendations for specific changes to the Code, however, entail design conservatism considerations beyond the scope of this review effort, and are not made in this report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗