Design and directed energy deposition of functionally graded dissimilar SS316 to TZM materials
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Materials science↗
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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
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Control of the Plasma-Material Interface for Long Pulse Optimization in EAST and KSTAR
This project's major goal was to understand and control the plasma-material interface to improve long pulse discharge and performance in the EAST and KSTAR devices, emphasizing the unique strengths of each machine. In particular, MIT has developed implanted depth markers to diagnose the net erosion of high-Z materials. The goals included the characterization and utilization of these markers in various high-Z materials of interest, including TZM, tungsten, and refractory metal base high entropy alloys.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Development of Hydrogen Permeation Barrier and Matrix Interaction Protection Coatings for TZM Enclosure of Hydride Moderators (Microreactor Program M3 Report)
It is important to develop a high-performance enclosure solution for hydride-based microreactor moderation components to effectively suppress hydride thermal decomposition and prevent hydrogen loss in harsh in-reactor environments. Currently, the DOE-NE Microreactor Program (MR program) is investigating the titanium-zirconium-molybdenum (TZM) canned yttrium hydride (YH 2-x ) solution, which provides acceptable hydrogen retainment performance up to ~700 °C. However, improvements in the enclosure are necessary for increasing long-term operation temperature beyond 700 °C and enhancing short-term survivability at even higher temperatures during power transients. One potential solution is the addition of a dedicated hydrogen permeation barrier, which can potentially improve the enclosure's performance. Furthermore, the hydride moderator components will be inserted into the microreactor matrix, which may consist of materials that are not chemically inert against TZM. Therefore, the TZM might need to be protected from the external environment through the application of another surface protection coating. To address these critical issues, two types of barrier coatings optimized for TZM-based moderator enclosure has been developed. A multilayer coating approach has been used to identify TZM-compatible barrier materials. Behavior of these coating designs against H 2 permeation and long-term high-temperature graphite matrix interactions has been examined in detail. Additionally, the developed coatings' behavior when exposed to both high-temperature thermal cycling and heavy-ion irradiation has also been investigated. Overall, the development of these barrier coatings is essential for achieving maturity for the metal hydride moderator technology. The results in this report will provide crucial insights into the behavior of the developed coatings and their potential for improving the long-term performance of TZM-based moderator enclosures.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Thermal and structural analysis of a new WEST reciprocating probe head with Titanium-Zirconium-Molybdenum armour material
The Tore Supra tokamak was transformed into an X-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support of the ITER tungsten divertor strategy. WEST began operation in 2017. A new reciprocating probe head, equipped with a Titanium-Zirconium-Molybdenum (TZM) heat shield, was recently built and operated on WEST during the C9 Campaign in early 2024. TZM was chosen as armour material for its very good mechanical properties between 700 °C and 1400 °C. This material was already extensively used for probe heads and other plasma facing components on the Alcator C-Mod tokamak. Here, this paper presents the design of the new WEST probe head. It is equipped with tungsten filaments that can be heated to emit electrons and provide a direct measurement of the plasma potential in the scrape-off layer, and it is also equipped with a pair of Langmuir probes. The probe plunges vertically into the plasma, reaches its maximum dive depth, then returns to its protected resting position in about 0.25 s. Thermal loads are calculated using as input the thermal properties of TZM and radial profiles of the heat flux measured in the past by other reciprocating probes, and scaled upwards to the expected heat loads in the WEST scrape-off layer. An analytical estimation of Electromagnetic (EM) loads will also be presented, as well as a thermal and structural analysis of the probe for 15 cycles at maximum plunge depth. The probe is observed by an infrared camera. Comparisons between the heat flux deduced by the Langmuir probe and the infrared radiance of the probe housing during plasma exposure are finally presented.
Diagnostic↗
Fabrication of a Mo based high temperature TZM alloy by non-consumable arc melting technique
High temperature structural materials are in great demand for power, chemical and nuclear industries which can perform beyond 1000 °C as super alloys usually fail. In this regard, Mo based TZM alloy is capable of retaining strength up to 1500 °C with excellent corrosion compatibility against molten alkali metals. Hence, currently this alloy is considered an important candidate material for high temperature compact nuclear and fusion reactors. Due to reactive nature of Mo and having high melting point, manufacturing this alloy by conventional process is unsuitable. Powder metallurgy technique has limited success due to restriction in quantity and purity. This paper deals with fabrication of TZM alloy by nonconsumable tungsten arc melting technique. Initially a ternary master alloy of Mo-Ti-Zr was prepared which subsequently by dilution method, was converted into TZM alloy gradually by external addition of Mo and C in various proportions. A number of melting trials were conducted to optimize the process parameters like current, voltage and time to achieve desired alloy composition. The alloy was characterized with respect to composition, elemental distribution profile, microstructure, hardness profile and phase analysis. Well consolidated alloy button was obtained having desired composition, negligible material loss and having microstructure as comparable to standard TZM alloy. (author)
36 MATERIALS SCIENCE↗
Progress Towards a Titanium-Zirconium-Molybdenum Alloy Coreblock Prototype Using Powder Bed Fusion
Microreactor designs make use of a coreblock for structural stability allowing for these devices to provide powder generation while being portable and inherently safe. In order to maximize energy output of the microreactors, materials must be able to withstand high temperature without melting and without substantial decreases in mechanical properties. As simultaneous projects looking at steels, graphite, and other materials are on-going, this work package will investigate titanium – zirconium – molybdenum alloy (TZM) as it allows for potentially higher microreactor operating temperatures than traditional reactor materials such as stainless steel. Notably, this alloy was developed in the 1960s for use in rockets and has a very high melting point (2623°C) but is difficult to form, machine and join. With limited needs in aerospace, literature is appropriately sparce in looking at the AM of TZM (examples include [1,2]). However, the properties appear to be an excellent match for the needs of the microreactor coreblock leading to the current project. In addition, small washers form the bulk of TZM parts needed in aerospace while the microreactor community was looking at larger, more complex parts. Therefore, AM of TZM is being explored to enable near – net shape of components with the needed complexity to impact the nuclear energy field.
36 MATERIALS SCIENCE↗
Thermal stability of oxidized ultrathin Li films on TZM for plasma facing components
For improved understanding of the behavior of Li plasma facing components (PFCs) in the presence of oxygen impurities, we report on the thermal stability of ultrathin (up to 1.0 nm) LiO x films on a titanium-zirconium-molybdenum (TZM) alloy substrate over the temperature range of 310–1400 K. LiO x films were prepared by either post-oxidation of Li films or Li films deposited on an oxygen-precovered TZM surface. Thermal stability of these films was studied using a combination of temperature programmed desorption (TPD), X-ray photoelectron spectroscopy (XPS), and low-energy ion scattering (LEIS). For post-oxidized Li films, TPD showed that no Li desorption occurred until temperatures above 620 K, and then Li desorbed from the surface via at least three desorption peaks. The O 1s XPS spectra on the post-oxidized films at 310 K indicate the formation of lithium oxide (Li2O) and peroxide/hydroxide (Li 2 O 2 /LiOH). The peroxide/hydroxide converts to oxide after heating to 680 K with no desorption of Li or O 2 , and then this film decomposes to liberate Li into the gas phase while leaving oxygen at the TZM surface. Heating the LiO x films to 1070 K led to a solid-state reaction with surface Mo atoms of the substrate to form a condensed binary lithium molybdenum oxide (LixMoOy) phase, which upon further heating decomposed to liberate gas phase binary lithium molybdenum oxide species. Li deposition on an oxygen-precovered TZM surface formed a complex, highly stable, oxygen rich, Li-O-Mo interfacial oxide. Li desorption from multilayer Li films on oxygen-precovered TZM surfaces occurred in a metallic Li multilayer peak and three other oxide-derived peaks. These results support the improved thermal stability of oxidized lithium as compared to metallic Li and have implications for Mo impurity transport into the plasma at high temperatures.
36 MATERIALS SCIENCE↗
Impact of divertor material on neutral recycling and discharge fueling in DIII-D
We report that experiments with the lower divertor of DIII-D during the Metal Rings Campaign (MRC) show that the fraction F of atomic D in the total recycling flux is material-dependent and varies through the ELM cycle, which may affect divertor fueling. Between ELMs, F C ~ 10% and FW ~ 40%, consistent with expectations if all atomic recycling is due to reflections. During ELMs, F C increases to 50% and F W to 60%. In contrast, the total D recycling coefficient including atoms and molecules R stays close to unity near the strike point where the surface is saturated with D. During ELMs, R can deviate from unity, increasing during high energy ELM-ion deposition (net D release) and decreasing at the end of the ELM which leads to ability of the target to trap the ELM-deposited D. The increase of R > 1 in response to an increase in ion impact energy Ei has been studied with small divertor target samples using Divertor Materials Evaluation System (DiMES). An electrostatic bias was applied to DiMES to change E i by 90 eV. On all studied materials including C, Mo, uncoated and W-coated TZM (>99% Mo, Ti, and Zr alloy), W, and W fuzz, an increase of E i transiently increased the D yield (and R) by ~10%. On C there was also an increase in the molecular D 2 yield, probably due to ion-induced D 2 desorption. Despite the measured increase in F on W compared to C, attached H-mode shots with OSP on W during MRC did not demonstrate a higher pedestal density. About 8% increase in the edge density could be seen only in attached L-mode scenarios. The difference can be explained by higher D trapping in the divertor and lower divertor fueling efficiency in H- versus L-mode.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
The RaDIATE High-Energy Proton Materials Irradiation Experiment at the Brookhaven Linac Isotope Producer Facility
The RaDIATE collaboration (Radiation Damage In Accelerator Target Environments) was founded in 2012 to bring together the high-energy accelerator target and nuclear materials communities to address the challenging issue of radiation damage effects in beam-intercepting materials. Success of current and future high intensity accelerator target facilities requires a fundamental understanding of these effects including measurement of materials property data. Toward this goal, the RaDIATE collaboration organized and carried out a materials irradiation run at the Brookhaven Linac Isotope Producer facility (BLIP). The experiment utilized a 181 MeV proton beam to irradiate several capsules, each containing many candidate material samples for various accelerator components. Materials included various grades/alloys of beryllium, graphite, silicon, iridium, titanium, TZM, CuCrZr, and aluminum. Attainable peak damage from an 8-week irradiation run ranges from 0.03 DPA (Be) to 7 DPA (Ir). Helium production is expected to range from 5 appm/DPA (Ir) to 3,000 appm/DPA (Be). The motivation, experimental parameters, as well as the post-irradiation examination plans of this experiment are described.
43 PARTICLE ACCELERATORS↗
Effect of Nickel on Brazed Joints Between TZM and Stainless Steel
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Materials science↗
Effect of Nickel on Brazed Joints Between TZM and Stainless Steel
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Materials science↗
Deformation Behaviour of TZM at Elevated Temperatures
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Materials science↗
Development of Surveillance Test Articles with Reduced Dimensions and Material Volumes to Support MSR Materials Degradation Management
This report details the efforts toward developing new surveillance test article designs with reduced dimensions and material volumes to support materials surveillance technology development for advanced reactors. Two fabrication methods for the surveillance test articles are described. Welded test articles were fabricated with 316H and A617 materials, and interlocking test articles were fabricated using A709 and titanium-zirconium-molybdenum (TZM). The preliminary results demonstrate the successful design and testing of the flat surveillance test articles. The report also describes ongoing efforts to use an induction heating test setup to increase the heat up and cool down rates in testing the surveillance test articles. A brief description of the planned FY-24 work is provided.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Erosion and re-deposition of lithium and boron coatings under high-flux plasma bombardment
Lithium and boron coatings are applied to the walls of many tokamaks to enhance performance and protect the underlying substrates. Li and B-coated high-Z substrates are planned for use in NSTX-U and are a candidate plasma-facing component (PFC) for DEMO. However, previous measurements of Li evaporation and thermal sputtering on low-flux devices indicate that the Li temperature permitted on such devices may be unacceptably low. Thus it is crucial to characterize gross and net Li erosion rates under high-flux plasma bombardment. Additionally, no quantitative measurements have been performed of the erosion rate of a boron-coated PFC during plasma bombardment. A realistic model for the compositional evolution of a Li layer under D bombardment was developed that incorporates adsorption, implantation, and diffusion. A model was developed for temperature-dependent mixed-material Li-D erosion that includes evaporation, physical sputtering, chemical sputtering, preferential sputtering, and thermal sputtering. The re-deposition fraction of a Li coating intersecting a linear plasma column was predicted using atomic physics information and by solving the Li continuity equation. These models were tested in the Magnum-PSI linear plasma device at ion fluxes of 10^23-10^24 m^-2 s^-1 and Li surface temperatures less than 800 degrees C. Li erosion was measured during bombardment with a neon plasma that will not chemically react with Li and the results agreed well with the erosion model. Next the ratio of the total D fluence to the areal density of the Li coating was varied to quantify differences in Li erosion under D plasma bombardment as a function of the D concentration. The ratio of D/Li atoms was calculated using the results of MD simulations and good agreement is observed between measurements and the predictions of the mixed-material erosion model. Li coatings are observed to disappear from graphite much faster than from TZM Mo, indicating that fast Li diffusion into the bulk graphite substrate occurred, as predicted. Li re-deposition fractions very close to unity are observed in Magnum-PSI, as predicted by modeling. Finally, predictions of Li coating lifetimes in the NSTX-U divertor are calculated. The gross erosion rate of boron coatings was also measured for the first time in a high-flux plasma device.
liquid metals↗
Electromagnetic Torque in Tokamaks with Toroidal Asymmetries
Lithium and boron coatings are applied to the walls of many tokamaks to enhance performance and protect the underlying substrates. Li and B-coated high-Z substrates are planned for use in NSTX-U and are a candidate plasma-facing component (PFC) for DEMO. However, previous measurements of Li evaporation and thermal sputtering on low-flux devices indicate that the Li temperature permitted on such devices may be unacceptably low. Thus it is crucial to characterize gross and net Li erosion rates under high-flux plasma bombardment. Additionally, no quantitative measurements have been performed of the erosion rate of a boron-coated PFC during plasma bombardment. A realistic model for the compositional evolution of a Li layer under D bombardment was developed that incorporates adsorption, implantation, and diffusion. A model was developed for temperature-dependent mixed-material Li-D erosion that includes evaporation, physical sputtering, chemical sputtering, preferential sputtering, and thermal sputtering. The re-deposition fraction of a Li coating intersecting a linear plasma column was predicted using atomic physics information and by solving the Li continuity equation. These models were tested in the Magnum-PSI linear plasma device at ion fluxes of 10^23-10^24 m^-2 s^-1 and Li surface temperatures less than 800 degrees C. Li erosion was measured during bombardment with a neon plasma that will not chemically react with Li and the results agreed well with the erosion model. Next the ratio of the total D fluence to the areal density of the Li coating was varied to quantify differences in Li erosion under D plasma bombardment as a function of the D concentration. The ratio of D/Li atoms was calculated using the results of MD simulations and good agreement is observed between measurements and the predictions of the mixed-material erosion model. Li coatings are observed to disappear from graphite much faster than from TZM Mo, indicating that fast Li diffusion into the bulk graphite substrate occurred, as predicted. Li re-deposition fractions very close to unity are observed in Magnum-PSI, as predicted by modeling. Finally, predictions of Li coating lifetimes in the NSTX-U divertor are calculated. The gross erosion rate of boron coatings was also measured for the first time in a high-flux plasma device.
liquid metals↗
High Temperature Moderator Cladding Report
The hydrogen permeation system at LANL has been modified and improved to perform high temperature testing of metals and metal/ceramic samples. This will allow us to demonstrate the ability to reduce hydrogen permeation for microreactor materials. For this program, two metals have been chosen for analysis, stainless steel as a base metal and TZM based on its high temperature properties and anticipated higher H permeation resistance. Samples have been fabricated and coated with Al 2 O 3 via atomic layer deposition in order to evaluate the potential to reduce hydrogen permeation with a thin layer ceramic coating. Current efforts include literature review of systems and sample fabrication to demonstrate feasibility, future work will focus on collecting hydrogen permeation data.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Report on Thermal Cycling Testing and Bimaterial Weld Development for a Passively Actuated Materials Surveillance Test Article
This report describes the continuation and conclusion of a thermal cycling test program designed to evaluate a family of passively actuated material surveillance test articles. Additionally, the report details a preliminary study of bimetallic welding between 316 stainless steel and TZM (tungsten-zirconium-molybdenum) alloy. The test program was ultimately successful in demonstrating the basic viability of the passively actuated test article concept as well as specifically validating the procedures developed to size specimens to mimic the mechanical response of operating components. The report discusses and summarizes the results of the complete test program which tested samples for more than 1,000 controlled thermal cycles. The weld development program included both stir friction and electron beam joining. The stir friction welding was ultimately successful, while the electron beam welding would require additional development. Ultimately, these test programs pave the way for implementing a materials surveillance program in future molten salt reactors (MSRs). The report conclusions outline a plan for transforming the research and developed described here into a definite set of practices and procedures for a surveillance program at a future operating reactor.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Solidification cracking of refractory alloys: a computational and machine learning study to investigate composition-dependence for improved weldability and additive manufacturability
Large-batch numerical, CALculation of PHAse Diagrams (CALPHAD)-based solidification cracking calculations are performed and then analyzed with machine learning methods to generate models that relate chemistry of refractory alloys to cracking susceptibility. Kou’s solidification cracking index is used to study the refractory alloys including O, N, C binary mixtures with Mo, Ta, Nb, and W, the molybdenum-based TZM, Niobium-based C103, and Tantalum-based T111 and Ta-10 W, as well as hypothetical refractory ternary alloys. Findings strongly validate Kou’s Crack Susceptibility Index (CSI) against Varestraint test data for Nb- and Ta-based alloys, establishing CSI thresholds where refractory alloys with CSI < 15,000 K are likely weldable, CSI > 15,000 K are prone to cracking, and CSI > 25,000 K are likely unweldable (or unprintable). Furthermore, interstitial elements C, N, and O significantly increase crack susceptibility, with some existing material specifications coinciding with peak cracking susceptibility concentrations. Finally, machine learning-derived elemental potency factors enable rapid prediction of CSI from alloy chemistry for C103, TZM, Ta-10 W, and T-111 alloys. These results provide practical guidance for feedstock selection, powder reuse limits, and alloy specification amendments for welding and additive manufacturing applications.
36 MATERIALS SCIENCE↗