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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Non-Neutron Transmutation of Used Nuclear Fuel (Final Report)

The primary goal of this study is to develop a national facility concept for transmuting long-lived fission products (LLFP) to substantially reduce the disposal impact by minimizing the need for a geologic-timescale repository. As a charter for this study, the national transmutation facility was required to reduce the radiotoxicity and decay heat of LLFP isotopes by at least 90% relative to their values at discharge from a commercial LWR, while consuming less than 10% of the reactor's energy. The identified LLFP isotopes are Se-79, Zr-93, Tc-99, I-129, Sn-126, and Cs-135, whose radiotoxicity is about 99% of the total radiotoxicity of all fission products at 1,000 years. Approximately ~72 kg of LLFPs is discharged every year from a 1,000 MWe commercial or advanced nuclear reactor. First, LLFP transmutation options with non-neutron beams (photons and protons) were explored. The study concluded that LLFP transmutation is feasible with high-energy, high-intensity photons or protons, but impractical on an engineering scale due to low transmutation rates and the high energy requirements to produce the desired photon or proton beams. As alternatives, LLFP transmutation options with neutrons from fission, fusion, and spallation reactions were additionally explored. The transmutation options using advanced critical reactors are attractive only for selective LLFP isotopes because the production rates of several LLFP isotopes (Zr-93, Sn-126, and Cs-135) from fission reactions are larger than the transmutation rates. The transmutation options with only spallation neutrons are favorable to transmute all LLFP isotopes, but as a tradeoff, the net transmutation rates are reduced. The national transmutation facility concept was developed following an exploration of transmutation options using various incident particles. The proposed national LLFP transmutation comprises a dedicated molten-salt reactor (MSR), a proton accelerator, and a spallation neutron-based transmuter. The MSR power was set at 300 MWt and 120 MWe, with the thermal power approximately 10% of that of a commercial 1,000 MWe PWR. The electricity generated by the MSR powers the accelerator and transmuter. The accelerator produces 1 GeV, 30 mA protons, which are introduced into the spallation neutron-based transmuter. The spallation neutron-based transmuter consists of a central spallation target and LLFP target pins merged in a heavy water tank. The six LLFP isotopes are separated into two groups. Tc-99, I-129, and Se-79, having larger neutron cross sections, belong to group A, while Zr-93, Sn126, and Cs-135, having smaller neutron cross sections, belong to group B. Then, for effective transmutation, LLFPs in groups A and B are transmuted in the dedicated MSR and in a spallation neutron-based transmuter, respectively. The estimated capital cost of the national transmutation facility is approximately $\$$3.1B, and its annual O&M cost is expected to be ~$\$$182M. Radiotoxicity and decay heat of LLFPs were calculated and compared with those of the original LLFPs. It was assumed that the targets were made with elementwise LLFP rather than isotopic LLFP, owing to the potentially high cost of isotopic separation from used nuclear fuels. The decay heat of LLFPs can be reduced by more than 90% using a single national transmutation facility. However, radiotoxicity decreases by 79–84%, which does not meet the transmutation performance requirement, primarily because Cs-135 is produced rather than depleted. Thus, to meet the design requirement, Cs-135 should be separated from other Cs isotopes and irradiated in a spallation neutron-based transmuter. Then, radiotoxicity decreases by ~92%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Design-informed neutronics assessment of long-lived fission product transmutation in a tokamak fusion reactor blanket

This study presents a neutronics-based assessment of the feasibility and viability of transmuting six major long-lived fission products (LLFPs) from light-water reactors, namely 99 Tc, 129 I, 79 Se, 93 Zr, 126 Sn, and 135 Cs, within the blanket region of a tokamak fusion reactor, using the MIT ARC design as a concrete fusion configuration. Monte Carlo neutronics simulations were performed to evaluate LLFP transmutation and to compare the results with a reference boiling water reactor (BWR). The results indicate that transmutation of all six LLFPs is neutronics-feasible in fusion reactors, with transmutation half-lives significantly shorter than their natural decay half-lives. For elemental targets, transmutation of 135 Cs, 126 Sn, and 93 Zr was found potentially viable, as the net mass transmuted exceeded that achievable in the reference BWR under identical target volume and irradiation time. When isotopically separated targets were considered, transmutation of 126 Sn and 93 Zr appeared potentially viable. A parametric study demonstrated that plasma geometry modifications can enhance local neutron flux, increasing the transmuted 93 Zr mass by approximately 33% and reducing the transmutation half-life from approximately 240 years to 180 years. Repositioning the target and adjusting material layer thickness reduced the transmutation half-life of 93 Zr to 67 years and increased the net mass transmuted by a factor of 50. Furthermore, these results demonstrate that fusion reactors can enable LLFP transmutation beyond the practical limits of thermal fission reactors and highlight the critical role of reactor and blanket design optimization. Engineering and fuel-cycle considerations required for deployment are beyond the scope of this neutronics-focused study.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Symmetry transmutation and anomaly matching

We explore a situation where a global symmetry of the ultraviolet (UV) theory does not act faithfully on the local infrared (IR) degrees of freedom, but instead acts effectively as a higher-form symmetry. We refer to this phenomenon as symmetry transmutation, where the UV symmetry is “transmuted” into a higher-form symmetry in the IR. Notably, unlike emergent (accidental) symmetries, which are approximate, these symmetries are exact. We illustrate the ubiquity of this phenomenon in various continuum and lattice systems and provide examples where the ’t Hooft anomalies of the UV symmetry are matched by those of the new higher-form symmetry in the IR. We also show that in certain phases and for certain energies, the UV baryon-number symmetry of one-flavor QCD is transmuted into a discrete one-form global symmetry. Finally, we compare our symmetry transmutation to the well-known phenomenon of symmetry fractionalization.

Anomalies in Field and String Theories↗
DOE OSTI2026

Defect formation and transmutation behaviors in irradiated tungsten under thermal-neutron shielded and mixed spectrum conditions

Here, this study investigates the microstructural evolution of pure tungsten irradiated under thermal-neutron shielded and mixed spectrum conditions in the High Flux Isotope Reactor (HFIR). Four samples were irradiated at temperatures from 570 °C to 1130 °C up to 0.73 dpa. Neutron spectrum significantly influenced the accumulation of transmutation products, with Re+Os content estimated at ∼0.3–0.6% under thermal-neutron shielded conditions and ∼5.2% under the mixed spectrum condition. Irradiation temperature strongly influences tungsten’s microstructure, with dislocation loops and fine voids forming at lower temperatures and only larger voids and Re/Os segregation observed at higher temperature. Under thermal-neutron shielded conditions, dislocation loops and voids were observed at 570 °C and 790 °C. At the highest irradiation temperature (1130 °C), dislocation loops were no longer observed, while larger but less dense voids remained. Re and Os segregation to void surfaces was evident at 790 °C and 1130 °C, though no precipitation was observed. In contrast, under the mixed-spectrum condition, both spherical and needle-like Re/Os-rich precipitates were observed, frequently accompanied by large voids. Dislocation loops were not observed, but loop-like contrast within the precipitates suggests they may have nucleated on pre-existing loops. Irradiation-induced hardening was assessed for the shielded samples at 570 °C and 790 °C. Dispersed barrier hardening (DBH) analysis, based on TEM-resolved defects, revealed that voids were the dominant contributors to hardening, consistent with literature results. A schematic model is proposed to describe defect and precipitate evolution in tungsten under fusion-relevant transmutation-to-dpa conditions.

Dislocation loops↗
DOE OSTI2026

Data-driven method to estimate contamination from light ion beam transmutation at colliders

Collisions of relativistic light ions, such as oxygen, neon, and magnesium, have been proposed as a way to examine the system-size dependence of dynamics typically associated with the quark-gluon plasma produced in collisions of heavier ions such as xenon, gold, or lead. Recent efforts at both the Relativistic Heavy Ion Collider (RHIC) and Large Hadron Collider (LHC) have produced large datasets of proton-oxygen, oxygen-oxygen, and neon-neon collisions, catalyzing intense interest in experimental backgrounds associated with light-ion collisions. In particular, electromagnetic dissociation of light ions while they are circulating in a collider can result in beam contamination that is difficult to simulate precisely. Here we propose a data-driven method for evaluating the potential impact of beam contaminants on physics analyses. The method exploits the time dependence and smaller size of contaminant ion species to define control regions that can be used to quantify potential contamination effects. A simple model is used to illustrate the method and to study its robustness. Furthermore, this method can inform studies of recent LHC and RHIC data and could also be useful for future light-ion programs at the LHC and beyond.

Beam loss↗
DOE OSTI2025

A data-driven method to estimate contamination from light ion beam transmutation at colliders

Collisions of relativistic light ions such as oxygen, neon, and magnesium, have been proposed as a way to examine the system-size dependence of dynamics typically associated with the quark-gluon plasma produced in collisions of heavier ions such as xenon, gold, or lead. Recent efforts at both the Relativistic Heavy Ion Collider (RHIC) and Large Hadron Collider (LHC) have produced large datasets of proton-oxygen, oxygen-oxygen, and neon-neon collisions, catalyzing intense interest in experimental backgrounds associated with light ion collisions. In particular, electromagnetic dissociation of light ions while they are circulating in a collider can result in beam contamination that is difficult to simulate precisely. Here we propose a data-driven method for evaluating the potential impact of beam contaminants on physics analyses. The method exploits the time-dependence and smaller size of contaminant ion species to define control regions that can be used to quantify potential contamination effects. A simple mode

Accelerator Physics (physics.acc-ph)↗
DOE OSTI2025

ToF-SIMS spectral analysis of pristine and neutron irradiated single crystal tungsten

Time-of-flight secondary ion mass spectrometry (ToF-SIMS) has many promising features in studying materials including high spatial resolution and high mass accuracy of elements, molecules, and isotopes. Its ability to resolve isotopes is especially attractive in studying transmutation products of single crystal tungsten (SCW) post neutron irradiation. Tungsten (W) is a contender of plasma facing materials (PFMs) due to its high thermal and radiological stability. PFMs to be used in the construction of fusion vessels are subject to high temperature and neutron irradiation, resulting in changes to materials including transmutation, which ultimately impact material mechanical and thermal properties. We used IONTOF TOF.SIMS V instrument equipped with a 30 keV Bi 3 + primary ion beam to study pristine SCW and irradiated SCW speciemens. Scanning electron microscope coupled with focused ion beam (SEM-FIB) was used to reduce the dosage of neutron irradiated tungsten and prepare for specimens for SIMS analysis. Static ToF-SIMS spectra were obtained, and transmutation product peak identification was presented in this work. Identified molecules and molecular fragments were compared against isotope theoretical mass to charge ratios of tungsten, rhenium, osmium, and other relevant products. Our results show that ToF-SIMS provides a viable means to study transmutation products of W post neutron irradiation. Such applications are suitable to investigate transmutation effects on materials that are being considered and developed for fusion pilot plants.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Impact of Time Dependent Reactor and Sensor Physics on Core Power Synthesis

Online synthesis of the power distribution is critical in the operation and control of nuclear power reactors to ensure that the core is operating within safety margins, and to provide essential knowledge associated with the burnup of the fuel. In light water reactors (LWRs), power synthesis is achieved by using some a priori knowledge of the state of the reactor core and updating based on the signals coming from in-core sensors—namely, self-powered neutron detectors (SPNDs). This report aims to study the effects of fuel burnup and sensor degradation on the ability to accurately synthesize the power distribution in a LWR. Several modeling tools were used to simulate power synthesis based on the responses of SPNDs, with emitters made out of Rh or V. A representative pressurized water reactor low-enriched uranium (LEU) core was modeled using the Polaris/Purdue Advanced Reactor Core Simulator (PARCS) approach. The Monte Carlo N-Particle Transport 6 (MCNP6) code was used, as well, to calculate response functions between different segments of fuel to individual SPNDs; this is a crucial parameter for power synthesis. The Oak Ridge Isotope GENeration (ORIGEN) package in the Standardized Computer Analyses for Licensing Evaluation (SCALE) code was used to model the time-dependent isotopic transmutation in the SPND emitters. All these data were fed into a custom code that enacted the point-based iterative (PBI) method to simulate power synthesis. Developmental work was also performed on high-fidelity SPND models in the GEometry ANd Tracking 4 (Geant4) code, which enables higher-accuracy modeling of the current responses from SPNDs. In this work, five sets of time-dependent power synthesis test cases were conducted. In these test cases, systematic changes in the input conditions enabled an analysis of the effect of (1) slightly inaccurate a priori power distribution assumptions with respect to fuel burnup, (2) highly inaccurate a priori assumptions with respect to fuel burnup (such that burnup is not included in the a priori assumed distribution), and (3) differences between Rh and V SPNDs in terms of downstream consequences of the transmutation in the emitters. The authors discovered that one may permissibly have slightly inaccurate a priori assumptions of the fuel burnup (such that the level of burnup may be slightly under- or over-approximated by the accumulated burnup in approximately 9.3 full power days), but to not account for burnup at all in the a priori assumption leads to severe levels of error, approaching 25% at maximum. The authors also discovered that V SPNDs are extraordinarily robust in the low-enriched uranium fuel cycle considered in this modeling work, whereas Rh SPNDs undergo significant transmutation that can result in large errors in the synthesized power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Impact of Time-Dependent Reactor and Sensor Physics on Core Power Synthesis (Rev.1)

Online synthesis of power distribution is critical in the operation and control of nuclear power reactors to ensure that the core is operating within safety margins and to provide essential knowledge associated with the burnup of the fuel. In light-water reactors, power synthesis is achieved by using some a priori knowledge of the state of the reactor core and updating based on the signals coming from in-core sensors—namely, self-powered neutron detectors (SPNDs). This report examines the effects of fuel burnup and sensor degradation on the ability to accurately synthesize the power distribution in a pressurized water reactor (PWR), considering the typical low-enriched uranium (LEU, 3%-5% enrichment) fuel cycle as well as the higher enrichment LEU+ (5%-8% enrichment) fuel cycle. Several modeling tools were used to simulate power synthesis based on the responses of SPNDs, with emitters made out of Rh or V. A representative PWR LEU core was modeled using the Polaris/Purdue Advanced Reactor Core Simulator (PARCS) approach. The Monte Carlo N-Particle Transport 6 (MCNP6) code was used as well to calculate response functions between different segments of fuel to individual SPNDs; this is a crucial parameter for power synthesis. The Oak Ridge Isotope GENeration (ORIGEN) package in the Standardized Computer Analyses for Licensing Evaluation (SCALE) code was used to model the time-dependent isotopic transmutation in the SPND emitters. All these data were fed into a custom code that enacted the point-based iterative method to simulate power synthesis. Developmental work was also performed on high-fidelity SPND models in the GEometry ANd Tracking 4 (Geant4) code, which enables higher-accuracy modeling of the current responses from SPNDs. In this work, five sets of time-dependent power synthesis test cases were conducted. In these test cases, systematic changes in the input conditions enabled an analysis of the effect of (1) slightly inaccurate a priori power distribution assumptions with respect to fuel burnup, (2) highly inaccurate a priori power distribution assumptions with respect to fuel burnup (such that burnup is not included in the a priori assumed distribution), and (3) differences between Rh and V SPNDs in terms of downstream consequences of the transmutation in the emitters and the extended nature of the LEU+ fuel cycle in comparison with LEU. The authors discovered that one may permissibly have slightly inaccurate a priori assumptions of the fuel burnup (such that the level of burnup may be slightly underapproximated or overapproximated by the accumulated burnup in approximately 9.3 full power days), but to not account for burnup at all in the a priori assumptions leads to severe levels of error, approaching 25% at maximum (for LEU). The authors also discovered that V SPNDs are extraordinarily robust in both the LEU and LEU+ fuel cycles considered in this modeling work, whereas Rh SPNDs undergo significant transmutation that can result in large errors in the synthesized power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

MX precipitate behavior in an irradiated advanced Fe-9Cr steel: Helium effects on phase stability

As part of an ongoing series aimed at optimizing Fe-9Cr reduced activation ferritic/martensitic (RAFM) alloys for fusion energy systems, this study explores MX precipitate behavior under dual-ion irradiations, specifically examining correlations between helium transmutation and irradiation-induced damage. Utilizing single and dual-beam ion irradiation, the research explores the combined effects of helium (10–25 appm He/dpa), temperature (400–600 °C), and damage levels (15–100 dpa) on the microstructural evolution of CNA9 steel, a variant of Castable Nanostructured Alloys (CNAs). The study demonstrates that helium co-implantation hinders radiation-enhanced coarsening of MX-TiC precipitates at 500 and 600 °C, maintaining MX-TiC precipitate stability at moderate damage levels (15 dpa) but failing to prevent complete precipitate dissolution at higher damage levels (≥50 dpa) when irradiated at 500 °C. Here, a generalized precipitate stability model suggests that helium-induced suppression of diffusion alters the balance between recoil resolution and back diffusion for MX-TiC precipitates, enhancing the current understanding of precipitate behavior under damage and transmutation simulated dual-ion irradiation conditions.

Characterization↗
DOE OSTI2025

Computing chemical potentials with machine-learning-accelerated simulations to accurately predict thermodynamic properties of molten salts

The successful design and deployment of next-generation nuclear technologies heavily rely on thermodynamic data for relevant molten salt systems. However, the lack of accurate force fields and efficient methods has limited the quality of thermodynamic predictions from atomistic simulations. Here we propose an efficient free energy framework for computing chemical potentials, which is the central free energy quantity behind many thermodynamic properties. We accelerate our simulations without sacrificing accuracy by using machine learning interatomic potentials trained on density functional theory (DFT) data. Using lithium chloride as our model system, we compute chemical potentials with DFT-accuracy for solid and liquid phases by transmuting ions into noninteracting particles. Notably, in the liquid phase, we demonstrate consistency whether we transmute one ion pair or the entire system into ideal gas particles. By locating the temperature where the chemical potential of solid and liquid phases cross, we predict a melting point of 880 ± 18 K for lithium chloride, which is remarkably close to the experimental value of 883 K. With this successful demonstration, we lay the foundation for high-throughput thermodynamic predictions of many properties that can be derived from the chemical potentials of the minority and majority components in molten salts.

Gibson, Luke D. [Oak Ridge National Laboratory (OR↗
DOE OSTI2025

Mg and native defects in cubic silicon carbide from first principles

The diffusion of Mg defects in 3C-SiC is studied using the density functional theory. Mg has the highest burn-in rate as a transmutant in 3C-SiC when it is placed in high-energy neutron irradiation environment of a fusion reactor. The presence and evolution of transmutant defects impact thermal and mechanical properties of this important structural material. This study is focused on understanding the structure, stability, and evolution of Mg defects and the interaction of Mg with native defects in 3C-SiC. Our calculations of diffusion coefficients for different Mg defects suggest that Mg is likely to diffuse faster in pristine 3C-SiC than in the damaged one, in agreement with earlier experimental observations.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Post-irradiation Examination of Eurofer-97 Steel Irradiated to 20 dpa at 200–400°C in HFIR under the EUROfusion (ORNL-KIT) Collaboration Program

The Oak Ridge National Laboratory/Karlsruhe Institute of Technology (ORNL/KIT) collaboration focuses on research involving irradiation experiments and post-irradiation examinations (PIE) of isotopically modified Eurofer-97 steels for fusion reactor applications. This collaboration capitalizes on ORNL’s expertise in radiation effects on materials and its capabilities in irradiation and PIE. The research aims to qualify Fe-9Cr-based Eurofer-97 steel under simulated fusion conditions, which include high doses and high transmutation rates. A unique isotopic modification technique is utilized in the research, wherein high-transmutation isotopes such as Fe-54 and Ni-58 are alloyed into the Eurofer steel to align with the expected helium production rates in fusion reactor conditions. This report presents the results of post-irradiation mechanical testing activities for the EUROFER-97/2 specimens after irradiation to ~20 dpa at various irradiation temperatures. The irradiation doses and temperatures for the ES capsules ranged from 18.4–20.4 dpa and 202–256°C, respectively. The mechanical property datasets, obtained through baseline testing and PIE of the tensile and fracture specimens, include microhardness data from 93 irradiated and non-irradiated tensile specimens and 28 irradiated bend bar specimens, uniaxial tensile property data from 45 irradiated and non-irradiated tensile specimens, and fracture toughness data from the bend bar specimens. Further data analyses provide statistical information on the microhardness values and tensile properties, as well as the reference ductile-brittle transition temperature (T 0 ) data.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Diversion Path Analysis: A Proposed Methodology to Develop an MC&A Approach for Liquid-Fueled Molten Salt Reactors

Nuclear material control and accounting (MC&A) is a critical element of both the US Nuclear Regulatory Commission (NRC) and US Department of Energy (DOE)’s domestic safeguards and security requirements. NRC licensees are required, under Title 10 of the Code of Federal Regulations (10 CFR) Part 74 to establish and maintain an MC&A program that captures and records the quantities and locations of special nuclear material (SNM) at the facility. Along with physical protection, MC&A is a key element of domestic nuclear material safeguards that enables the NRC to ensure that SNM is controlled and accounted for. SNM, per 10 CFR Part 74, refers to plutonium, 233 U, and uranium enriched in the isotope 233 U or 235 U, but does not include source material. Periodic physical inventories, coupled with material balance evaluations, are effective and demonstrated tools to account for and detect theft or diversion of SNM in facilities containing SNM in bulk material form (i.e., not in discrete, countable items). Historically in the United States, these types of facilities have included fuel fabrication, conversion, and enrichment facilities. In comparison, reactors have relied on item counting of assemblies and control of SNM while in containment (e.g., a sealed reactor pressure vessel) because, to date, reactor fuel has been in item form. In liquid-fueled molten salt reactors (MSRs), unlike traditional light water reactors (LWRs) or bulk facilities, bulk SNM quantities can change significantly during operation as a result of depletion and transmutation. This introduces challenges to the use of traditional periodic physical inventories and material balance evaluations to detect theft or diversion of SNM in reactors that use SNM in bulk material form. Liquid-fueled (i.e., salt-fueled) MSR facilities are MSRs that use SNM within a salt eutectic as the fuel. The SNM is in a bulk material form any time it is outside of fresh or spent fuel storage containers. Some examples of when SNM will be in bulk form in the facility are during addition of fuel to the reactor system, while fuel is circulating in operation, and while fuel is in a drain tank. Periodic physical inventories and material balance evaluations can likely be effectively applied to many portions of an MSR facility, including all areas where depletion and transmutation are not significantly changing the quantities of SNM within the control area. Within an MSR facility, this would include fresh fuel receipt and loading, waste streams that may contain SNM, irradiated fuel storage outside of the reactor core, and any irradiated fuel processing that may happen after SNM has been removed from the reactor. All of these process steps could rely on measurements of SNM quantities compared with documented inventories. Any discrepancies from predicted (i.e., book) inventories and measured inventories could be quantified as inventory differences, consistent with traditional MC&A guidance from the NRC (e.g., in NUREG-1065 Revision 2, NUREG-2159 Revision 1, and RG 5.29 Revision 2). Within the reactor system, additions and removals to the book inventory include depletion of the SNM (e.g., fission of 235 U), which complicates the use of physical inventories. SNM control, however, can also likely be effectively applied to detect theft of SNM throughout a liquid-fueled MSR facility. To complement these approaches, prior technical reports have identified that a diversion path analysis may be a useful, risk-informed, and performance-based tool to determine suitable elements of an MC&A approach for the reactor system within a liquid-fueled MSR facility.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

An Integrated ML/AI Framework for Digitizing, Structuring and Searching DOE U-TRU-Fuels Data with Gap Analysis of Non-DOE Records

The U.S. Department of Energy (DOE) Advanced Fuels Campaign (AFC) is advancing transmutation fuel technologies to reduce long-lived radioactive waste by converting minor actinides into shorter-lived or stable elements through irradiation in sodium-cooled fast reactors. Key experiments such as AFC-1, AFC-2, FUels for the transmutation of Trans-URanium elements In phéniX (FUTURIX)-Fortes Teneurs en Actinides (FTA), and Experimental Breeder Reactor-II (EBR-II) X501 have provided fuel fabrication, irradiation, and performance data on various transuranic-bearing fuel forms. This report documents the creation of an artificial-intelligence assisted database, which has consolidated all DOE-owned data related to Transuranic (TRU)-bearing fuel experiments and stored across it across both the Idaho National Laboratory (INL) Nuclear Data Management and Analysis System and the INL high performance computing (HPC) infrastructure. A dedicated webpage, hosted on the INL HPC system, has been developed to support role-based access and data interaction. The database architecture allows researchers to navigate large, heterogeneous archives with far greater speed and accuracy than manual search and lays the foundation for future expansion into multimodal nuclear materials analysis environments. The database represents a major step towards a nationally integrated fuels database utilizing artificial intelligence tools.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A Brief Review of the Impact of Neutron Irradiation Damage in Tungsten and Its Alloys

Neutron irradiation poses a substantial challenge in the development and application of tungsten (W) and its alloys, predominantly in the framework of nuclear fusion and fission environments. Although W is well-acknowledged for its unique properties like its high melting temperature and higher resistance to sputtering, transmutation products, such as Re and Os, form and impact the alloy properties as a result of neutron irradiation. This transmutation effect accompanied by significant microstructure damage due to neutron irradiation can lead to the significant degradation of mechanical properties. This review surveys the literature focusing on the microstructural modifications post-irradiation and its impacts on the irradiation hardening. This review provides insights into the elaborative understanding on the neutron radiation damage on W and W alloys by exploring the microstructural evolution and hardness changes post-irradiation. The gaps and future opportunities for understanding neutron radiation damage in W are briefly summarized.

36 MATERIALS SCIENCE↗