Engineering topics
Publications and source records attributed to Maloy, Stuart Andrew.
Hydrogen infused nuclear reactor core monoliths and processes for their production are disclosed. Such monoliths may function as both a core monolith and a moderator, providing structure and at least some moderation. Hydrogen infused monoliths may be complementary to or be used in lieu of separate moderators for thermalizing neutrons.
Techniques for using additive manufacturing (AM) to fabricate creep resistant ferritic/martensitic steel with improved high temperature strength are described. AM processing may be performed on Grade 91 steel powder. Beam powers from about 221 W to about 270 W may be used. Traverse rates from about 675 mm/s to about 825 mm/s may be used. Heat inputs ranging from about 55.7 J/mm3 to about 83.2 J/mm 3 may be produced. Creep resistant ferritic/martensitic steel, produced according to the present disclosure, has improved strain yield strength and ductility as compared to wrought steel.
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The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T m ) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300°C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≃100) relative to the Low N specimen (ΔHV≃160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.
Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe 2+ up to 600 peak displacement-per-atom (dpa) at 450 °C. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. Furthermore, the comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.
Joining nanostructured ferritic alloys (NFAs) has proved challenging, as the nano-oxides that provide superior strength, creep resistance, and radiation tolerance at high temperatures tend to agglomerate, redistribute, and coarsen during conventional fusion welding. In this study, capacitive discharge resistance welding (CDRW)—a solid-state variant of resistance welding—was used to join end caps and thin-walled cladding tubes of the NFA 14YWT. The resulting solid-state joints were found to be hermetically sealed and were characterized across the weld region using electron microscopy (macroscopic, microscopic, and nanometer scales) and nanoindentation. Microstructural evolution near the weld line was limited to narrow (~50–200 μm) thermo-mechanically affected zones (TMAZs) and to a reduction in pre-existing component textures. Dispersoid populations (i.e., nano-oxides and larger oxide particles) appeared unchanged by all but the highest energy and power CDRW condition, with this extreme producing only minor nano-oxide coarsening (~2 nm → ~5 nm Ø). Despite a minimal microstructural change, the TMAZs were found to be ~10% softer than the surrounding base material. These findings are considered in terms of past solid-state welding (SSW) efforts—cladding applications and NFA-like materials in particular—and in terms of strengthening mechanisms in NFAs and the potential impacts of localized temperature–strain conditions during SSW.
The performance of materials under extreme environments poses important fundamental questions about the behavior of condensed matter under far-from-equilibrium conditions. These conditions create challenges in materials design, testing and evaluation. One important aspect of advancing nuclear power is the discovery and investigation of new classes of materials able to withstand the harsh environments in nuclear reactors. However, investigating and understanding the performance of these materials requires sophisticated tools and characterization techniques and skills. Furthermore, the goal of this special issue is to present recent research on the characterization of the response of candidate advanced nuclear materials to extreme environments.
Beyond the current commercial materials, refining the grain size is among the proposed strategies to manufacture resilient materials for industrial applications demanding high resistance to severe environments. Here, large strain machining (LSM) was used to manufacture nanostructured HT-9 steel with enhanced thermal stability, mechanical properties, and ductility. Nanocrystalline HT-9 steels with different aspect rations are achieved. In-situ transmission electron microscopy annealing experiments demonstrated that the nanocrystalline grains have excellent thermal stability up to 700 °C with no additional elemental segregation on the grain boundaries other than the initial carbides, attributing the thermal stability of the LSM materials to the low dislocation densities and strains in the final microstructure. Nano-indentation and micro-tensile testing performed on the LSM material pre- and post-annealing demonstrated the possibility of tuning the material’s strength and ductility. The results expound on the possibility of manufacturing controlled nanocrystalline materials via a scalable and cost-effective method, albeit with additional fundamental understanding of the resultant morphology dependence on the LSM conditions.
This work highlights some limitations of thermal stability analysis via in-situ transmission electron microscopy (TEM)-annealing experiments on ultrafine and nanocrystalline materials. We provide two examples, one on nanocrystalline pure copper and one on nanocrystalline HT-9 steel, where in-situ TEM-annealing experiments are compared to bulk material annealing experiments. The in-situ TEM and bulk annealing experiments demonstrated different results on pure copper but similar output in the HT-9 steel. The work entails discussion of the results based on literature theoretical concepts, and expound on the inevitability of comparing in-situ TEM annealing experimental results to bulk annealing when used for material thermal stability assessment.
Tantalum is a refractory metal, it has low coefficient thermal expansion and high creep resistance. Its BCC structure makes Ta less prone to radiation damage. It is 2nd most ductile among refractories. Its oxide (Ta 2 O 5 ) layer is thin and stable at high temperatures and low oxygen concentrations.
The main purpose of this presentation is to demonstrate that a high throughput method for fabrication and mechanical testing produces local and bulk-like results.
FeCrAl alloys are being developed for accident tolerant fuels for current light water reactors. Specifically as fuel claddings with enhanced safety. These alloys are being developed to have excellent corrosion resistance in light water reactor coolant environments and structural integrity for longer durations in the case of a loss of coolant scenario where time is critical for safety. Tensile, shear punch, and Vickers hardness testing was performed on neutron irradiated FeCrAl alloys with three variations in composition. Testing was done at room temperature for comparison of the three techniques. FeCrAl alloys in this report were irradiated at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR) to 7 dpa and at 320 °C.
This document provides updates on the work package listed above for activities for the period 10/1/2020 - 6/30/2021. The specific milestone involved reporting on samples fabricated by DED-L AM from Grade 91 powder. Details from the prior report covering an earlier period are included here for completeness.
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The Nuclear Energy Agency (NEA) Expert Group on Innovative Structural Materials (EGISM) was established in 2008 under the guidance of the Nuclear Science Committee (NSC). Its objectives are to conduct joint and comparative studies to support the development, selection and characterisation of innovative structural materials that can be implemented in advanced nuclear fuel cycles, under long service lifetime and extreme conditions, such as high temperature, high dose/dose rate and corrosive chemical environments. In this context of growing interest and initiatives, the EGISM initiated at the beginning of 2018 an activity among its members to: Identify, in a non-exhaustive way, the currently existing programs on innovative materials and fabrication processes in NEA member countries and China; Establish a first cartography of the activities that are underway on these topics and identify common subjects and thematic; Propose a Technology Readiness Level scale to estimate the maturity of both innovative materials and fabrication processes; Carry out a reflection on what the enablers are to quickly climb this TRL scale, as well as the obstacles, in order to identify solutions to overcome them. This paper first gives definitions shared between the EGISM members on what are considered as advanced structural materials solutions. Next, some international initiatives for the accelerated development of high performance materials are presented both in non-nuclear and nuclear fields. Then, the methodology adopted for technology readiness assessment is explained. A non-exhaustive synthesis of the projects identified among the EGISM members on innovative structural materials and advanced manufacturing solutions such as additive manufacturing is presented. The TRL level of these projects is evaluated. A focus is also made on some of these projects to illustrate and explain the TRLs chosen as well as highlighting enablers or obstacles identified to climbing up the TRL scale.
In this work, ferritic-martensitic (F/M) HT9 along with fully ferritic HT9 specimens were irradiated using self-ions to 600 peak displacement-per-atom (dpa) at 450 °C. To investigate and minimize the carbon contamination effect on void swelling, two identical F/M HT9 specimens were prepared: one without a coating and the other one with a tungsten and silver coating on the surface prior to irradiation. The specimen without a coating showed carbide formation in the matrix, resulting from carbon contamination during irradiation, with no voids observed. The coated sample did not form matrix carbides, and voids were observed within the irradiated region. The same coating was applied to the ferritic HT9, and void swelling results of both the F/M HT9 and the ferritic HT9 alloys were compared after irradiation. The F/M HT9 exhibited a maximum of 2.6% swelling, while the ferritic HT9 showed a maximum of 0.8% swelling at a 400–500 nm depth, equivalent to 334 local dpa. The results were correlated with the instability of carbides in the ferritic HT9 under irradiation. Carbides in the F/M HT9 and the ferritic HT9 are both M23C6, but the matrix carbides in the ferritic HT9 were less stable under the irradiation, compared to the ones on the grain boundaries in the F/M HT9. The study shows the impact of chemistry on the void swelling which overpowers the microstructural factors such as grain size and phase.
Microstructural analyses and micro-pillar compression were conducted on 14YWT nanostructured ferritic alloy (NFA) to compare different processing pathways: hydrostatic extrusion and Pilger processing with varying annealing temperatures into thin walled tubing, and after hot extrusion and cross-rolling into a plate. Hydrostatic extrusion at 815 °C resulted in the smallest grain sizes and highest yield strength of 1.20 GPa. Pilger processing with annealing at 800 °C had fine grained regions and bands of coarse grains, leading to a large variation in yield strength of 0.9–1.40 GPa. Higher annealing temperatures of 1200 °C after pilger processing significantly increased the grain size and lowered the yield strength to 1.01 GPa. These tubes showed a stronger (111) crystallographic texture in the normal direction and elongated grains in the extrusion direction. Characterization of the nano-oxides using TEM reveals more numerous, smaller oxides present in tubing processed at lower temperatures. Overall, this work shows NFA tubing after hydrostatic extrusion and pilger processing can lead to fine grained microstructures and texturing leading to higher yield strengths at lower annealing temperatures (e.g. 800 °C).
The axial tensile properties of C26M tube were successful measured at Los Alamos National Laboratory. Test were conducted at a constant strain rate of 10 -3 s -1 and three different temperatures (22, 300, and 600°C). As typically expected, an increase in the testing temperature lead to a decrease in both the yield stress (YS), from 710 to 273 MPa, and ultimate tensile strength (UTS), from 730 to 290 MPa. The ultimate tensile strength was observed almost immediately post-yielding for all test conditions, and thus the uniform elongation was limited to 1.2-1.6%. The total elongation of the 600°C specimens was notably higher (~47%) than both the 22 and 300°C specimens, which exhibited a similar response (~10%).