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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Materials Data on SrMg2 by Materials Project

Mg2Sr is Tungsten Carbide-like structured and crystallizes in the hexagonal P-6m2 space group. The structure is three-dimensional. Sr is bonded in a 12-coordinate geometry to six Mg atoms. There are three shorter (3.59 Å) and three longer (3.62 Å) Sr–Mg bond lengths. There are three inequivalent Mg sites. In the first Mg site, Mg is bonded to six equivalent Mg atoms to form distorted edge-sharing MgMg6 pentagonal pyramids. All Mg–Mg bond lengths are 3.07 Å. In the second Mg site, Mg is bonded to three equivalent Sr and three equivalent Mg atoms to form a mixture of distorted face, edge, and corner-sharing MgSr3Mg3 pentagonal pyramids. In the third Mg site, Mg is bonded in a 6-coordinate geometry to six equivalent Sr atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Mo2CN by Materials Project

Mo2N(C) is Tungsten Carbide-derived structured and crystallizes in the orthorhombic Pmm2 space group. The structure is three-dimensional. there are two inequivalent Mo+3.50+ sites. In the first Mo+3.50+ site, Mo+3.50+ is bonded to four equivalent C4- and two equivalent N3- atoms to form a mixture of distorted edge, face, and corner-sharing MoC4N2 pentagonal pyramids. All Mo–C bond lengths are 2.20 Å. Both Mo–N bond lengths are 2.20 Å. In the second Mo+3.50+ site, Mo+3.50+ is bonded to two equivalent C4- and four equivalent N3- atoms to form distorted MoC2N4 pentagonal pyramids that share corners with twelve MoC4N2 pentagonal pyramids, edges with six MoC4N2 pentagonal pyramids, and faces with two equivalent MoC2N4 pentagonal pyramids. Both Mo–C bond lengths are 2.19 Å. All Mo–N bond lengths are 2.20 Å. C4- is bonded to six Mo+3.50+ atoms to form distorted CMo6 pentagonal pyramids that share corners with four equivalent CMo6 pentagonal pyramids, corners with eight equivalent NMo6 pentagonal pyramids, edges with two equivalent CMo6 pentagonal pyramids, edges with four equivalent NMo6 pentagonal pyramids, and faces with two equivalent CMo6 pentagonal pyramids. N3- is bonded to six Mo+3.50+ atoms to form distorted NMo6 pentagonal pyramids that share corners with four equivalent NMo6 pentagonal pyramids, corners with eight equivalent CMo6 pentagonal pyramids, edges with two equivalent NMo6 pentagonal pyramids, edges with four equivalent CMo6 pentagonal pyramids, and faces with two equivalent NMo6 pentagonal pyramids.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on MgSi2 by Materials Project

MgSi2 is Tungsten Carbide-like structured and crystallizes in the triclinic P-1 space group. The structure is three-dimensional. there are three inequivalent Mg sites. In the first Mg site, Mg is bonded to six Si atoms to form MgSi6 octahedra that share corners with six SiMg3Si3 octahedra, edges with six MgSi6 octahedra, and edges with six SiMg3Si3 octahedra. The corner-sharing octahedral tilt angles are 17°. There are a spread of Mg–Si bond distances ranging from 2.87–2.89 Å. In the second Mg site, Mg is bonded to six Si atoms to form MgSi6 octahedra that share corners with six SiMg3Si3 octahedra, edges with six MgSi6 octahedra, and edges with six SiMg3Si3 octahedra. The corner-sharing octahedral tilt angles are 17°. There are four shorter (2.88 Å) and two longer (2.89 Å) Mg–Si bond lengths. In the third Mg site, Mg is bonded to six Si atoms to form MgSi6 octahedra that share corners with six SiMg3Si3 octahedra, edges with six MgSi6 octahedra, and edges with six SiMg3Si3 octahedra. The corner-sharing octahedra tilt angles range from 16–17°. There are a spread of Mg–Si bond distances ranging from 2.87–2.89 Å. There are four inequivalent Si sites. In the first Si site, Si is bonded to three Mg and three Si atoms to form distorted SiMg3Si3 octahedra that share corners with three MgSi6 octahedra, corners with three SiMg3Si3 octahedra, edges with three MgSi6 octahedra, and edges with nine SiMg3Si3 octahedra. The corner-sharing octahedra tilt angles range from 0–17°. All Si–Si bond lengths are 2.37 Å. In the second Si site, Si is bonded to three Mg and three Si atoms to form distorted SiMg3Si3 octahedra that share corners with three equivalent MgSi6 octahedra, corners with three SiMg3Si3 octahedra, edges with three MgSi6 octahedra, and edges with nine SiMg3Si3 octahedra. The corner-sharing octahedra tilt angles range from 0–17°. Both Si–Si bond lengths are 2.37 Å. In the third Si site, Si is bonded to three Mg and three Si atoms to form distorted SiMg3Si3 octahedra that share corners with three MgSi6 octahedra, corners with three SiMg3Si3 octahedra, edges with three MgSi6 octahedra, and edges with nine SiMg3Si3 octahedra. The corner-sharing octahedra tilt angles range from 0–17°. There are one shorter (2.36 Å) and two longer (2.37 Å) Si–Si bond lengths. In the fourth Si site, Si is bonded to three Mg and three Si atoms to form distorted SiMg3Si3 octahedra that share corners with three equivalent MgSi6 octahedra, corners with three SiMg3Si3 octahedra, edges with three MgSi6 octahedra, and edges with nine SiMg3Si3 octahedra. The corner-sharing octahedra tilt angles range from 0–17°.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na2O2 by Materials Project

Na2O2 is Tungsten Carbide-like structured and crystallizes in the orthorhombic Pbcm space group. The structure is three-dimensional. Na is bonded to six equivalent O atoms to form a mixture of distorted edge, corner, and face-sharing NaO6 octahedra. The corner-sharing octahedra tilt angles range from 31–52°. There are a spread of Na–O bond distances ranging from 2.40–2.43 Å. O is bonded to six equivalent Na atoms to form a mixture of distorted edge and corner-sharing ONa6 pentagonal pyramids.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on CaZn by Materials Project

CaZn is Tungsten Carbide-like structured and crystallizes in the trigonal R-3m space group. The structure is three-dimensional. Ca is bonded to six equivalent Zn atoms to form a mixture of distorted edge and corner-sharing CaZn6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Ca–Zn bond lengths are 3.12 Å. Zn is bonded to six equivalent Ca atoms to form a mixture of distorted edge and corner-sharing ZnCa6 octahedra. The corner-sharing octahedral tilt angles are 0°.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Li3H24Ru(ClO2)6 by Materials Project

(Li(H2O)4)3RuCl6 is Tungsten Carbide-like structured and crystallizes in the monoclinic C2/c space group. The structure is zero-dimensional and consists of four ruthenium(6+) hexachloride molecules and four Li(H2O)4 clusters. In each Li(H2O)4 cluster, there are two inequivalent Li1+ sites. In the first Li1+ site, Li1+ is bonded to six O2- atoms to form face-sharing LiO6 octahedra. There are a spread of Li–O bond distances ranging from 2.16–2.19 Å. In the second Li1+ site, Li1+ is bonded to six O2- atoms to form face-sharing LiO6 octahedra. There are two shorter (2.14 Å) and four longer (2.16 Å) Li–O bond lengths. There are twelve inequivalent H1+ sites. In the first H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the second H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the third H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the fourth H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.99 Å. In the fifth H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the sixth H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the seventh H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the eighth H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the ninth H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the tenth H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the eleventh H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. In the twelfth H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. There are six inequivalent O2- sites. In the first O2- site, O2- is bonded in a distorted water-like geometry to one Li1+ and two H1+ atoms. In the second O2- site, O2- is bonded in a distorted water-like geometry to one Li1+ and two H1+ atoms. In the third O2- site, O2- is bonded in a distorted water-like geometry to one Li1+ and two H1+ atoms. In the fourth O2- site, O2- is bonded in a distorted water-like geometry to two Li1+ and two H1+ atoms. In the fifth O2- site, O2- is bonded in a distorted water-like geometry to two Li1+ and two H1+ atoms. In the sixth O2- site, O2- is bonded in a distorted water-like geometry to two Li1+ and two H1+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on CoMo4N5 by Materials Project

Mo4CoN5 is Tungsten Carbide-derived structured and crystallizes in the orthorhombic Amm2 space group. The structure is three-dimensional. there are four inequivalent Mo+3.25+ sites. In the first Mo+3.25+ site, Mo+3.25+ is bonded to six N3- atoms to form a mixture of distorted corner, edge, and face-sharing MoN6 pentagonal pyramids. There are two shorter (2.13 Å) and four longer (2.18 Å) Mo–N bond lengths. In the second Mo+3.25+ site, Mo+3.25+ is bonded to six N3- atoms to form a mixture of distorted corner, edge, and face-sharing MoN6 pentagonal pyramids. There are four shorter (2.18 Å) and two longer (2.19 Å) Mo–N bond lengths. In the third Mo+3.25+ site, Mo+3.25+ is bonded to six N3- atoms to form a mixture of distorted corner, edge, and face-sharing MoN6 pentagonal pyramids. There are four shorter (2.16 Å) and two longer (2.19 Å) Mo–N bond lengths. In the fourth Mo+3.25+ site, Mo+3.25+ is bonded to six N3- atoms to form a mixture of distorted corner, edge, and face-sharing MoN6 pentagonal pyramids. There are four shorter (2.16 Å) and two longer (2.17 Å) Mo–N bond lengths. Co2+ is bonded in a 6-coordinate geometry to six N3- atoms. There are four shorter (2.07 Å) and two longer (2.36 Å) Co–N bond lengths. There are five inequivalent N3- sites. In the first N3- site, N3- is bonded in a 6-coordinate geometry to two equivalent Mo+3.25+ and four equivalent Co2+ atoms. In the second N3- site, N3- is bonded to six Mo+3.25+ atoms to form a mixture of distorted corner, edge, and face-sharing NMo6 pentagonal pyramids. In the third N3- site, N3- is bonded to six Mo+3.25+ atoms to form a mixture of distorted corner, edge, and face-sharing NMo6 pentagonal pyramids. In the fourth N3- site, N3- is bonded to six Mo+3.25+ atoms to form a mixture of distorted corner, edge, and face-sharing NMo6 pentagonal pyramids. In the fifth N3- site, N3- is bonded in a 6-coordinate geometry to four equivalent Mo+3.25+ and two equivalent Co2+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

MH-1A Sturgis Decommissioning and Dismantlement - 20281

The U.S. Army Corps of Engineers (USACE) with its prime contractor, Aptim Federal Services, LLC (APTIM), recycled 5,260 metric tons (11.6 million pounds) of material from Sturgis Barge (Sturgis) during the Decommissioning and Dismantlement of the MH-1A nuclear power reactor. The overall objective of the project was to reduce residual radioactivity associated with MH-1A to levels that permitted release of Sturgis for dismantlement and termination of the Army Reactor Office permit. By effectively applying waste hierarchy's three Rs - reduce, reuse and recycle - the Sturgis project not only minimized the amount of waste that required disposal at landfills, but also reduced the potential for long-term environmental liability emanating from these landfills. The project team completed the physical decommissioning efforts in June 2018 in Galveston, TX. In September 2018, radiological surveys were completed to demonstrate the vessel could be released for shipbreaking. Sturgis was towed from Galveston, TX to Brownsville, TX in late September 2018. Shipbreaking, dismantlement and recycling efforts began in early October 2018 and were completed on 15 March 2019. As part of the decommissioning effort in Galveston, the team shipped 69 shipments (860 metric tons) of low-level radioactive waste and radioactive components to the Waste Control Specialist (WCS) facility in Andrews TX for disposal. Certain radioactive components had to be transferred to the Department of Energy prior to placing the materials into the Federal Waste Facility located within WCS. However, most of the radioactive waste was characterized, profiled, approved and managed under the WCS permitted radioactive waste exemption process authorized and implemented by the Texas Commission on Environmental Quality (TCEQ) and the Radioactive Materials Division. This allows LLRW and LLMW to be shipped as regulated waste and then upon receipt at WCS through satisfying the relevant waste acceptance criteria the waste is exempted and placed into the WCS RCRA permitted cell. An additional 35 shipments (544 metric ton) of contaminated hazardous waste water were transported to U.S. Ecology in Robstown, TX for treatment/disposal. An additional 36 shipments (500 metric tons) of non-hazardous wastewater was sent to Republic Waste Services' facility in Fresno, TX. The disposal of these materials required close coordination with State of Texas regulators. During decommissioning, the project team recycled approximately 270 metric tons (600,000 pounds) of lead and steel. As part of the dismantlement in Brownsville, TX the team recycled approximately 5,000 metric tons (11 million pounds) of ferrous and non-ferrous material, limiting our disposal requirements to about 180 metric tons (400,000 pounds) of material (<4% from entire shipbreaking activity). Although the primary hazard being mitigated by this project was radiological, recycling was always a priority for the project. The team strived to achieve sustainability goals as we implemented this one of a kind project. Scrap metal recycling has a large positive impact on the environment and can also favorably impact project disposal costs. Steel is among the most recycled material in the world. Nearly 40% of the world's steel production is made from scrap. Recycling steel also requires 75% less energy than producing it from raw materials. By using recycled steel rather than virgin materials, 2.33 kg of carbon emissions are eliminated per kg of steel [1]. The project recycled more than 4,500 metric tons (10 million pounds) of steel, which eliminated about 10,400 metric tons (23 million pounds) of CO{sub 2}. By implementing a recycling initiative for the Sturgis project, the team was able to realize cost avoidance for disposal of scrap, cost savings from the metals recycled, plus the project provided benefits to the environment through our recycling efforts. Once the dismantlement was complete, the team prepared a detailed decommissioning closure report, which allowed for the termination of the Army Reactor Decommissioning Permit. While it not only reduced any potential long-term environmental liability, this project to decommission and dismantle a floating nuclear power plant is truly unprecedented - it is a prime example of the USACE mission which is: 'Engineering solutions for the Nation's toughest challenges'. This unique, one of a kind, historical power plant was never designed to be taken apart, and the available information about its construction was lacking in many details. The hazards that required mitigation dictated a painstaking and deliberate process in order to avoid any release to the environment and the community, and to protect the health and safety of the workers involved while keeping the waste hierarchy's three R's - reduce, reuse and recycle at the forefront. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Characterization of boundary precipitation in a heavy ion irradiated tungsten heavy alloy under the simulated fusion environment

In the concerted effort to identify materials capable of surviving the adverse environment of a fusion reactor interior, tungsten heavy alloys have been put forth as leading candidates. Experimental trials and behavioral studies have yielded positive results for their adoption by taking advantage of the alloy’s unique balance of inherently high fracture toughness and low sputter yields; yet due to their relative novelty in the fusion community, there remains a lack of understanding on the response of these materials to the extended high temperature irradiation environment of the reactor interior. Further, to alleviate this issue and provide the necessary data on the behavior of tungsten heavy alloys to the simulated fusion environment, a 90W-7Ni-3Fe alloy has been subjected to elevated temperature sequential Ni + and He + ion irradiations to mimic the expected displacement damage and gas production expected after five years of service as a plasma facing material component. Atomic-scale structural analyses and nanoscale chemical mapping have identified the formation of two distinct precipitation structures, a surface localized η-carbide and an a-W 2 C type tungsten carbide, both of which appear to originate at the bi-phase interface between W and the ductile phase. This irradiation assisted and induced precipitate formation respectively is anticipated to adversely affect these materials by selective embrittlement at bi-phase interfaces leading to a reduction in the material’s overall fracture toughness during prolonged high temperature irradiation. It is asserted that any potential exposure to C during fusion reactor operational service should be minimized as to prevent the formation of these phases.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Evaluation of Radiation Transport Through a Nuclear-Grade Sandwich Composite

The development and deployment of microreactors are often synonymous with their transportability; however, designing a shield which is appropriate for both transportation and operations is challenging. A recent approach is the use of composite style shield which combines the reactor pressure vessel and biological shielding into a single unit to enhance the shields effectiveness without degrading the structural properties. To assess the viability of a nuclear grade sandwich composite (NGSC), and understanding of how radiation transports through the shield needs to be understood to select the appropriate materials. This work examines a two-layered NGSC of varying thicknesses with two primary shielding materials (tungsten-tetraboride cermet and boron carbide). To assess the NGSC, quantities of interest including neutron attenuation, photon attenuation, energy deposition, displacements per atom, helium generation, and dose are examined. Altogether, it is found that a combination of both tungsten-tetraboride and boron carbide are necessary to act as an appropriate shield, where the order in which these materials are place in the composite are important to their effectiveness.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Control rods for light water reactors

A control rod for a nuclear fuel assembly is described herein that includes a neutron absorbing material having a melting point greater than 1500° C. that does not form a eutectic with a melting point less than 1500° C., and may further include a cladding material having a melting point greater than 1500° C. The cladding material is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material is selected from the group consisting of Gd2O3, Ir, B4C, Re, and Hf. The metal cladding or the absorbing material may be coated with an anti-oxidation coating of Cr with or without a Nb intermediate layer.

Lahoda, Edward J.↗
DOE OSTI2026

Retention and surface morphology evaluation of fine-grain dispersion-strengthened tungsten for plasma-facing component applications

This study exposed novel fine-grain dispersion-strengthened tungsten (W) to high fluence, low energy deuterium (D) and helium (He) plasmas to evaluate how material microstructure and composition affect hydrogen retention and surface morphology. Tested materials included fine-grain dispersion-strengthened tungsten (DSW) with 3 wt% zirconium carbide (ZrC) dispersoids, fine-grain dense W without any dispersoids (FGW), and coarse-grained polycrystalline ‘ITER-grade’ W. Samples were exposed to D 2 + and He + plasmas at fusion-relevant fluences (∼10 25 m -2 ) and ion energies (75 eV) over a range of temperatures (200 °C, 300 °C, 450 °C for D, 850 °C for He). Helium ion microscopy was performed on the exposed samples to evaluate surface morphology changes and material integrity. After D plasma exposure, the ZrC dispersoids showed near-surface degradation at exposure temperatures above 300 °C, but no detrimental morphology changes were observed for the adjacent W grains. After He plasma-exposure, nano-structured fuzz formation was observed in the tungsten matrix of all samples. The ZrC dispersoids maintained their integrity despite the surrounding fuzz growth, with clear delineation between the W fuzz and dispersoid regions. Thermal desorption spectroscopy showed that ZrC DSW consistently retained more D than the FGW by about a factor of 2 across all temperatures. At 200 °C and 300 °C, the ITER-W displayed lower D retention than both the DSW and FGW, however at 450 °C ITER-W showed the highest retention, about 50% more than DSW. He retention was comparable across all samples, with the highest retention observed in the fine-grain W, only 26% higher than in ITER-W. These insights on retention behavior will inform further optimization of these novel fine-grained tungsten materials with and without dispersoid additives.

Dispersion-strengthened tungsten↗
DOE OSTI2019

Development of Multicomponent EAM Potential for Ni-Based Superalloy

We initiated the development of multi-component EAM potentials for Aluminides and Carbides, key phases in Ni-based Superalloys. The goal is to utilize the MD simulation to understand the deformation dynamics that contribute to the formation of voids and creep initiation. For this purpose, we constructed the raw data from ab-initio (molecular dynamics) MD simulations fed into the potential development code and used Nickel as the base metal with the addition of a number of various elements including Aluminum, Chromium, Tungsten. We then developed the EAM potentials for the aluminide and carbide phases using the force-fitting code MEAMfit. Our generated potential reproduces the fundamental properties of the Ni3Al and Mo23C6 phases. We verified further the EAM potential through the thermal stability test at different temperatures and by reproducing the elastic constants consistent with the experimental values.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Additive manufacturing of complex objects using refractory matrix materials

A method for the manufacture of a three-dimensional object using a refractory matrix material is provided. The method includes the additive manufacture of a green body from a powder-based refractory matrix material followed by densification via chemical vapor infiltration (CVI). The refractory matrix material can be a refractory ceramic (e.g., silicon carbide, zirconium carbide, or graphite) or a refractory metal (e.g., molybdenum or tungsten). In one embodiment, the matrix material is deposited according to a binder-jet printing process to produce a green body having a complex geometry. The CVI process increases its density, provides a hermetic seal, and yields an object with mechanical integrity. The residual binder content dissociates and is removed from the green body prior to the start of the CVI process as temperatures increase in the CVI reactor. The CVI process selective deposits a fully dense coating on all internal and external surfaces of the finished object.

Terrani, Kurt A.↗
DOE OSTI2024

Additive manufacturing of complex objects using refractory matrix materials

A method for the manufacture of a three-dimensional object using a refractory matrix material is provided. The method includes the additive manufacture of a green body from a powder-based refractory matrix material followed by densification via chemical vapor infiltration (CVI). The refractory matrix material can be a refractory ceramic (e.g., silicon carbide, zirconium carbide, or graphite) or a refractory metal (e.g., molybdenum or tungsten). In one embodiment, the matrix material is deposited according to a binder-jet printing process to produce a green body having a complex geometry. The CVI process increases its density, provides a hermetic seal, and yields an object with mechanical integrity. The residual binder content dissociates and is removed from the green body prior to the start of the CVI process as temperatures increase in the CVI reactor. The CVI process selective deposits a fully dense coating on all internal and external surfaces of the finished object.

Terrani, Kurt A.↗
DOE OSTI2020

High Flux Helium Irradiation of Dispersion-Strengthened Tungsten Alloys and Effects of Heavy Metal Impurity Layer Deposition

Tungsten has been chosen as the plasma-facing material (PFM) for the divertor region in ITER and also a candidate PFM for future plasma-burning nuclear fusion reactors. During fusion device operation, PFMs will be exposed to low-energy He irradiation at high temperatures, resulting in sub-surface bubbles and surface morphology changes such as pores and fuzz. Carbide dispersion-strengthened W materials may enhance the ductility of W, but their behavior under high flux He irradiation remains unclear. In this work, the response of dispersion-strengthened tungsten materials to high flux, low energy He irradiation at high temperature is examined. Tungsten alloyed with 1, 5, or 10 wt. % tantalum carbide or titanium carbide exposed to these conditions result in surface pores, coral-like feature growth and sub-surface helium bubbles. Reactor-relevant helium irradiation (5 x 10 26 m -2 fluence) combined with high powered laser pulses to simulate off-normal reactor events does not significantly alter the surface morphology, as the surface nanostructures appear stable and cracks are only observed on a localized region of one sample. However, specimens show the development of an impurity layer on the surface, likely impurity deposition from the sample holder during irradiation, resulting in a mixed material layer on the surface. Helium bubbles exist in this impurity layer, and obscure conclusions about helium interactions with the carbide dispersoids. Nonetheless, it is clear that the dispersoid microstructure limits He bubble formation and subsequent surface nanostructuring, attributed to the dispersoid composition.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Comparison of void swelling of ferritic-martensitic and ferritic HT9 alloys after high-dose self-ion irradiation

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.

36 MATERIALS SCIENCE↗