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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Fusion Materials Semiannual Progress Report for the Period Ending December 31, 2023

This is the seventy-fifth in a series of semiannual technical progress reports on fusion materials science activity supported by the Fusion Energy Sciences Program of the U.S. Department of Energy. It covers the period ending December 31, 2023. This report focuses on research addressing the effects on materials properties and performance of exposure to the neutronic, thermal and chemical environments anticipated in the chambers of fusion experiments and energy systems. This research is a major element of the national effort to establish the materials knowledge base for an economically and environmentally attractive fusion energy source. Research activities on issues related to the interaction of materials with plasmas are reported separately. The results reported are the products of a national effort involving a number of national laboratories and universities. A large fraction of this work, particularly in relation to fission reactor irradiations, is carried out collaboratively with partners in Japan, Russia, and the European Union. The purpose of this series of reports is to provide a working technical record for the use of program participants, and to provide a means of communicating the efforts of fusion materials scientists to the broader fusion community, both nationally and worldwide.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Fusion Materials Semiannual Progress Report for the Period Ending June 30, 2024

This is the seventy-sixth in a series of semiannual technical progress reports on fusion materials science activity supported by the Fusion Energy Sciences Program of the U.S. Department of Energy. It covers the period ending June 30, 2024. This report focuses on research addressing the effects on materials properties and performance of exposure to the neutronic, thermal and chemical environments anticipated in the chambers of fusion experiments and energy systems. This research is a major element of the national effort to establish the materials knowledge base for an economically and environmentally attractive fusion energy source. Research activities on issues related to the interaction of materials with plasmas are reported separately. The results reported are the products of a national effort involving a number of national laboratories and universities. A large fraction of this work, particularly in relation to fission reactor irradiations, is carried out collaboratively with partners in Japan, Russia, and the European Union. The purpose of this series of reports is to provide a working technical record for the use of program participants, and to provide a means of communicating the efforts of fusion materials scientists to the broader fusion community, both nationally and worldwide. This report has been compiled by Stephanie Melton, Oak Ridge National Laboratory. Her efforts, and the efforts of the many persons who made technical contributions, are gratefully acknowledged.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Mk-IV Salt Crystallization Hot Finger Apparatus for Partitioning Used Electrorefiner Salt

Electrorefining is a controlled redox process used to regulate the behavior of ionic species. Through this process, metals can be deposited onto a cathode from an electrolyte solution in a controlled manner. The Mk-IV electrorefiner (Mk-IV ER) at Idaho National Laboratory is an engineering-scale, molten salt-based electrorefining cell that has been used for decades to recover metallic uranium from spent fuel. As a result, highly stable fission product chlorides have accumulated in the electrolyte. This accumulation results in changes to the salt’s properties, such as melting temperature, thermal conductivity, and density, as well as elevated product impurity and fissile materials criticality margin. These factors prompt the need for a salt regeneration process, such as melt-crystallization and species drawdown. This work focuses on providing a conceptual design to regenerate ER salt from used Mk-IV-ER salt in-situ, while minimizing salt waste volumes by concentrating the fission products in a final processed salt heal. We propose using a hot-finger crystallization apparatus design to fractionally crystallize salt in the Mk-IV-ER head space (or baffle space), allowing the collection of solid and liquid fractions. By using a cup-drain design, the used salt will be allowed to slowly solidify on the walls of a stainless-steel cup. The apparatus drain plug will then open to allow the liquid salt phase to drain to a lower cup, effectively separating the liquid phase from the solid phase. Under the hypothesis that the liquid phase salt concentrates the fission products, which is under examination in the accompanying work package, this separation allows the recovered solid salt to be reused while minimizing the high-level salt waste volume of used ER salt.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Spontaneous fission of the odd- Z isotope 255 Db

Experiments conducted at Lawrence Berkeley National Laboratory's 88-Inch Cyclotron Facility aimed to produce and study the decay of the previously unobserved isotope 255 Db. This isotope was produced in the 206 Pb( 51 V, 2⁢n) 255 Db reaction, separated from unreacted beam material and reaction by-products with the Berkeley Gas-filled Separator, and then implanted into a double-sided silicon-strip detector at the BGS focal plane. Decay properties of 255 Db were determined from the analysis of evaporation residue (EVR) fission and EVR-α–α correlations. The properties of this new isotope of dubnium differ dramatically from those of its neighboring Db isotopes. 255 Db was found to decay primarily by spontaneous fission (SF) with a small α-decay branch, where the average half-life of the observed decays was t 1/2 = 2.6$^{+0.4}_{– 0.3}$ ms. Theoretical calculations were performed using the Wentzel-Kramers-Brillouin approximation, with parameters calculated within a self-consistent microscopic approach, to see if these unique properties could be reproduced. A SF half-life estimate is obtained that closely matches the measured value, while simultaneously pointing out the sensitivities that need to be further constrained in future work.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

True molecular conformation and structure determination by three-dimensional electron diffraction of PAH by-products potentially useful for electronic applications

The true mol­ecular conformation and the crystal structure of benzo[e]di­naphtho­[2,3-a;1',2',3',4'-ghi]fluoranthene, 7,14-di­phenyl­naphtho­[1,2,3,4-cde]bis­anthene and 7,16-di­phenyl­naphtho­[1,2,3,4-cde]heli­anthrene were determined ab initio by 3D electron diffraction. All three mol­ecules are remarkable polycyclic aromatic hydro­carbons. The mol­ecular conformation of two of these com­pounds could not be determined via classical spectroscopic methods due to the large size of the mol­ecule and the occurrence of multiple and reciprocally connected aromatic rings. The mol­ecular structure of the third mol­ecule was previously considered provisional. These com­pounds were isolated as by-products in the synthesis of similar products and were at the same time nanocrystalline and available only in very limited amounts. 3D electron diffraction data, taken from submicrometric single crystals, allowed for direct ab initio structure solution and the unbiased determination of the inter­nal mol­ecular conformation. Detailed synthetic routes and spectroscopic analyses are also discussed. Based on many-body perturbation theory simulations, benzo[e]di­naphtho­[2,3-a;1',2',3',4'-ghi]fluoranthene may be a promising candidate for triplet–triplet annihilation and 7,14-di­phenyl­naphtho­[1,2,3,4-cde]bis­anthene may be a promising candidate for inter­molecular singlet fission in the solid state.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Source Term Reduction for Advanced and Small Modular Boiling Water Reactors

The United States Department of Energy (US DOE) is currently supporting the development of various advanced and small modular reactor (SMR) designs. Several of these designs have commenced license application with the US Nuclear Regulatory Commission (US NRC). These reactors have improved safety features that may significantly reduce radiological source terms in the event of design and beyond-design basis accidents. Specifically, some reactors feature a smaller containment volume relative to the available fission product depositional surface area, which supports increased fission product retention in the containment vessel. Pressurized water reactors (PWR) and boiling water reactors (BWR) with this feature include the integral pressurized water reactor (iPWR) and the BWRX-300 design by General Electric. A prior research program supported by the US DOE quantified the source term reduction associated with light water iPWRs and developed iPWR-specific theoretical models for fission product deposition rates. This program included a sequence of research projects that started with a feasibility study, development of theoretical models that predict higher deposition rates, and finally, development of empirical data for verification and validation of the theoretical models. The current project, which is a feasibility study, is the first step in a similar program to quantify the source term reduction associated with small and advanced light water BWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Diffusion and phase formation in the γ-uranium-technetium system

Phase formation in the U-Tc binary system at 800 °C was investigated using a diffusion couple experiment. Scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) identified four novel potential intermetallic phases - U 7 Tc 3 , U 13 Tc 12 , U 3 Tc 5 , and UTc 4 . Diffusion coefficients were calculated for the intermetallic phases using the Boltzmann-Matano method and were respectively found to be – 120, 38.2, 15.6, and 1.51 × 10 −18 m 2 /s. Tc also exhibits a solid-solution phase with high penetration into the U with a diffusion coefficient of ∼ 10 −14 m 2 /s. Furthermore, these findings expand the number of known U-Tc phases and provide the first diffusion coefficients for the U-Tc system, and contribute valuable data to the broader field of actinide metallurgy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

IER 517: Molybdenum Critical Experiment Design [Slides]

In nuclear fuels, U-Mo metallic fuels are used in new space reactor designs and new research reactor fuels. MITR, MURR, and NBSR reactors plan on converting from different HEU fuels to a U-Mo HALEU fuel. In spend nuclear fuel, 95 Mo is one of the 15 main absorbing fission products in irradiated LWR fuel. This makes it important for criticality safety studies in transportation and reprocessing. In structural materials. Molybdenum is found commonly in alloys that make up the structural materials of nuclear reactors such as type 316 SS. Molybdenum helps to improve high temperature performance and corrosion resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Optimal Chloride Salt Mixture for a Fusion Blanket

Deuterium-tritium fusion reactors cannot operate for a significant period without a closed tritium fuel cycle, according to a recent National Academies of Sciences (NAS) report on bringing fusion reactors to the US electrical grid. This fact places breeder blankets as one of the foundational systems for self-sustained fusion reactor operation. The typical functional requirements for a breeder blanket system include producing tritium, absorbing kinetic energy, transporting thermal energy, and being environmentally attractive. State-of-the-art research on liquid blankets has converged to primarily focus on (LiF) 2 and BeF 2 (FLiBe) molten salts and a metallic eutectic of lead and lithium (PbLi), but “virtually all of the technologies related to the tritium fuel cycle are at a low technological readiness level”. This work sought to explore optimum blanket configurations as it aligned with the Oak Ridge National Laboratory (ORNL) FY 2023 Laboratory Directed Research and Development Program’s research priority of developing and expanding the current understanding of fusion blanket science and technology. This purpose of this work was to address ORNL research priorities and NAS recommendations by investigating novel liquid blanket materials that could provide self-sustaining operation and draw on experience from research on molten salts used for advanced fission reactors, concentrated solar, and thermal energy storage. The hypothesis when proposing this research was that there could be chloride-based blanket designs that can exceed the tritium breeding ratios of (FLiBe) molten salt blankets while reducing the use of Be (FLiBe), avoiding the generation of HF (FLiBe), and minimizing magnetohydrodynamic (MHD)-perturbed flow fields (PbLi). The fastest and most cost-effective path to deploying liquid fusion breeder blankets could be from maximizing the synergistic technological overlap between fusion, fission, concentrated solar, and thermal energy storage industries.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Alpha Decay Chains as Thermal Power Sources: Analysis and Applications for RTGs

Radioactive sources can provide power in remote and environmentally harsh locations such as the arctic or space. The generators powered by such sources are rugged and can withstand extreme temperatures, lack of sunlight, and require no human intervention for multiple years. Radioisotopes are used in thermoelectric generators to provide power at remote sites and deep in space. Isotopes like Pu-238, Cm-244, and Am-241 are used in these generators by NASA for power in space probes and spacecrafts. These power sources deliver a steady supply of energy over extended periods of time. Alpha particles created during decay do not travel far in a material. Their kinetic energy is transferred to heat that we can then convert into energy. Unlike beta and gamma decay, the slower-moving alpha particles stop in the material, making their energy available for use. Energy from these natural decay processes provides a reliable source of power. Spontaneous fission is rare and unreliable, and unlike induced fission processes, alpha decay occurs naturally and does not require external management or ignition. The ideal properties of an isotope for use as a power source depend upon the intended use. For use in an Arctic research base over a period of several years, but less than a decade, an isotope that provides high power output over a shorter lifespan may be the most suitable option. Whereas, for deep space missions where a consistent power source for decades or perhaps more than 100 years is needed that would require a very different isotope. One with a much longer half-life that would provide consistent power throughout that time and survive in that state in for these extended periods of time. These examples represent two extreme sides in terms of time frames. By analyzing the power produced by different radioactive decay processes over time, we can evaluate the suitability of various isotope decay chains for specific uses. Some unstable isotopes undergo a series of radioactive decays, transforming into different isotopes at each step and resulting in a stable isotope. The lists of isotopes in these decay processes are known as decay chains. Some of these chains, illustrated in the figures below, are currently being investigated for use in radioisotope thermoelectric generators (RTGs) designed for a range of operational durations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

In-Situ Ion Irradiation to Add Irradiation Assisted Grain Growth to the Marmot Tool

In the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the MARMOT mesoscale fuel performance tool is used to inform the development of mechanistic materials models for the BISON fuel performance tool. The grain size of the fuel has a large impact on its performance, directly impacting heat conduction, fission gas release, creep, and fracture. Thus, atomistic and mesoscale MARMOT simulations have been used to investigate grain boundary migration and grain growth in UO 2 . However, the current grain growth models in MARMOT does not consider the effect of irradiation on causing grain growth. This research project experimentally studied and implemented irradiation effects on grain growth of UO 2 into MARMOT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2008

Development and Testing of Space Fission Technology at NASA-MSFC

The Early Flight Fission Test Facility (EFF-TF) at NASA-Marshall Space Flight Center (MSFC) provides a capability to perform hardware-directed activities to support multiple inspace nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations allowing for realistic thermal-hydraulic evaluations of systems. The EFF-TF is currently performing non-nuclear testing of hardware to support a technology development effort related to an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled reactor design, which builds on US and Russian space reactor technology as well as extensive US and international terrestrial liquid metal reactor experience. An important aspect of the current hardware development effort is the information and insight that can be gained from experiments performed in a relevant environment using realistic materials. This testing can often deliver valuable data and insights with a confidence that is not otherwise available or attainable. While the project is currently focused on potential fission surface power for the lunar surface, many of the present advances, testing capabilities, and lessons learned can be applied to the future development of a low-cost in-space fission power system. The potential development of such systems would be useful in fulfilling the power requirements for certain electric propulsion systems (magnetoplasmadynamic thruster, high-power Hall and ion thrusters). In addition, inspace fission power could be applied towards meeting spacecraft and propulsion needs on missions further from the Sun, where the usefulness of solar power is diminished. The affordable nature of the fission surface power system that NASA may decide to develop in the future might make derived systems generally attractive for powering spacecraft and propulsion systems in space. This presentation will discuss work on space nuclear systems that has been performed at MSFC's EFF-TF over the past 10 years. Emphasis will be place on both ongoing work related to FSP and historical work related to in-space systems potentially useful for powering electric propulsion systems.

Polzin, Kurt↗
DOE OSTI2023

The advanced characterization, post-irradiation examination, and materials informatics for the development of ultra high-burnup annular U-10Zr metallic fuel

U-Zr metallic fuel is a promising fuel candidate for Gen Ⅳ fast spectrum reactors. Previous experimental irradiation campaigns showed that the sodium thermal bonded U-10Zr fuel design can achieve a burnup of 10% fissions per initial heavy metal atom (FIMA). Advanced metallic fuel designs are pushing the burnup limit to 20% or even 30% FIMA. To achieve the higher burnup and eliminate the pyrophoric sodium, a prototypical annular fuel has been designed, fabricated, clad with HT-9 in the Materials and Fuels Complex, and irradiated in the Advanced Test Reactors of Idaho National Laboratory (INL) to a peak burnup of 3.3% FIMA. During irradiation, the mechanical contact between fuel and cladding acts as a thermal bond. The irradiation lasted for 132 days in the reactor. Recently, the archived fresh and irradiated fuel samples were characterized using advanced characterization capabilities in the Irradiated Materials Characterization Laboratory (IMCL) of INL. This article summarizes the results of advanced characterization and computer vision-based materials informatics to reveal the irradiation effects on U-Zr metallic fuel. Future work will focus on further implementation of advanced characterization and statistical data mining to improve the fidelity of fuel performance modeling and support U-Zr metallic fuel qualification for fast spectrum reactors.

Yao, Tiankai↗
NASA NTRS2011

Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology demonstration unit (TDU). In November, 2011 testing of a 37-pin core simulator (designed in conjunction with Los Alamos National Laboratory) for use with the TDU will occur. Previous testing at the EFFTF has included the thermal and mechanical coupling of a pumped NaK loop to Stirling engines (provided by GRC). Testing related to heat pipe cooled systems, gas cooled systems, heat exchangers, and other technologies has also been performed. Integrated TDU testing will begin at GRC in 2013. Thermal simulators developed at the EFF-TF are capable of operating over the temperature and power range typically of interest to compact reactors. Small and large diameter simulators have been developed, and simulators (coupled with the facility) are able to closely match the axial and radial power profile of all potential systems of interest. A photograph of the TDU core simulator during assembly is provided in Figure 2.

Houts, Michael G.↗
DOE OSTI2022

Three-dimensional microstructural characterization of FBR MOX fuel and the contribution of microstructural features to the thermal conductivity of the fuel

Combination of microstructural characterization, property measurements, and phase field modeling is used to investigate fast breeder reactor (FBR) mixed oxide (MOX) fuel irradiated to burnup of 13.7% fissions per initial metal atom (FIMA). Here, the fuel was characterized at different radial locations, which revealed that grey phase can be present in the central region if it nucleates on five metal precipitates (FMPs). In addition, in the mid-radial region FMPs do not diffuse out of the region once formed and the size of Pd–Te precipitates is dictated by the porosity present in the region. Thermal conductivity measurements were conducted as a function of radial location and the microstructure of the fuel was correlated with the observed trend. Reconstructions of the 3D solid and gaseous fission product structures in different regions of the fuel were used to simulate the effective thermal conductivity (ETC) of the respective regions and determine which microstructural feature has the strongest impact on thermal conductivity. Based on conducted assessment, FMPs and Pd-Te precipitates improve local conductivity of central and mid-radial regions even in the presence of grey phase, but defects are primary contributor to the degradation of thermal conductivity on the periphery of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Exploring Fission–Fusion Synergies to Accelerate Compatibility Understanding

To address the significant commercial interest in fusion energy, it will be necessary to accelerate the compatibility research associated with liquid breeders including Li, eutectic Pb–Li and LiF-BeF 2 (FLiBe) molten salt. Particularly for FLiBe, compatibility understanding is limited especially for fusion relevant materials such as reduced activation ferritic-martensitic steels, SiC and V alloys. The historical knowledge associated with molten salt reactors (MSRs) and recent work to commercialize MSRs can benefit fusion research. Recent experimental and modeling work has improved understanding and this knowledge can be applied to fusion relevant materials. For liquid metals (LMs), the comparisons to Li and Pb–Li are less direct but nevertheless can help guide the pathway toward commercialization. For Pb–Li, Al-rich coatings have been shown to inhibit dissolution and potentially increase operating temperatures. For commercialization, the experience with sensors and on-line cleanup can help guide future developments. Thus, it is worth considering the potential for fission-related research with LMs and molten salts to help accelerate fusion research.

36 MATERIALS SCIENCE↗