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Here, understanding the performance of the selected experiment based on the priorities of the study is fundamental when creating and choosing a design.
Tritium is produced from neutron interactions with both lithium and beryllium. Large quantities of tritium are generated in Molten Salt Reactors (MSRs) which use LiF/BeF2 (FLiBe) as the fuel salt. Tritium is unique among the radionuclide hazards as it readily permeates through metal structural materials at high temperatures. All metal surfaces are potential release paths for tritium. For adequate safety analysis and eventual licensing of new reactors, predictive models for tritium transport and release from MSRs must be developed. These models must account for the multiple transport phenomena involved with tritium: fuel salt phase mass transport, dissociation/recombination reactions on metal surfaces, interstitial diffusion through the metal structure, and salt or gas phase mass transport in the downstream fluid. These models also must also be validated with representative experiments. Our previous report outlined tritium transport phenomena involved in MSRs, made suggestions on gaps in the transport dataset, and proposed an experimental test stand to test combined transport effects – tritium transport through pipe walls in a convective salt flow. In this report, we summarize an updated analysis framework for tritium transport in MSRs, report our results on hydrogen and deuterium permeation through Hastelloy N, and describe the final design of the Molten Salt Tritium Transport Experiment (MSTTE, pronounced “misty”). The MSRE provides the only wholistic experimental data set for tritium transport in MSRs and understanding the transport phenomena involved in the MSRE is crucial for future model development. One set of parameters in our analysis framework was unknown for the MSRE—surface reaction rates for tritium on Hastelloy N. This warranted our hydrogen and deuterium permeation campaign to assess the permeability, diffusivity, and solubility of hydrogen isotopes in clean Hastelloy N. Surface reaction rate constants were probed by low pressure measurements, however, no surface effects were observed in the limits of our permeation apparatus. Permeation experiments on oxidized Hastelloy N were not performed for this report but are planned in future work. The experimental test stand, MSTTE, measures combined transport properties of the salt-metal system. MSTTE is a forced convection FLiBe loop with custom designed test section to measure tritium transport through candidate structural materials. We use MSRE relevant dimensionless numbers to design and scale the test section. Hastelloy N is a candidate loop and test section material due to the relevance for the MSRE and related designs, however, other metals are being considered (e.g. 316H SS) which may better align with current vendor concepts.
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Photovoltaics and concentrating solar thermal power are two ways for generating electricity from sunlight, albeit through different methods. Parabolic trough style powerplants represent 3.6 GW of electricity production, but many of these plants are aging and being replaced with photovoltaics. An alternative option that could be employed to leverage the sunk capital cost associated with the primary optics would be the design of a pure photovoltaic retrofit working within the existing plant architecture. Here, a secondary optical concentrator is designed to use the existing primary optics of a parabolic trough type solar thermal powerplant. The design is a v-shaped secondary concentrator resulting in a predicted concentration ratio on a 20 mm wide target of 94. The concentrating photovoltaic receiver for retrofit of an RP-3 based parabolic trough has been constructed using multi-junction concentrator photovoltaic cells and experimentally demonstrated here for the first time. Calculated performance of the cells based on cell specifications should result in 31% efficiency at 85 °C. On-sun efficiencies were measured at an average value of 21% with operational temperatures between 55 and 120 °C. Levelized cost of electricity calculations predict the system to have the potential to be below 7¢/kWh based on predicted efficiencies and 13¢/kWh based on the measured values at cell costs of $5/cm 2 .
The metallic fuel safety performance under unprotected design-basis transients is a key consideration for the deployment of advanced sodium fast reactors (SFRs). Reliable data are needed to validate advanced safety codes, reduce uncertainty in cladding failure thresholds, and strengthen confidence in licensing approaches. To address this need, this report develops blueprints for a conceptual sodium loss-of-flow (LOF) experiment in the Mk-IIIR loop at the Transient Reactor Test Facility (TREAT), providing the technical foundation for future integral testing.
The accurate interpretation of uranium metal 230 Th/ 234 U and 231 Pa/ 235 U radiochronometry model ages requires an understanding of how uranium parent nuclides and decay progeny ( 230 Th and 231 Pa) behave during uranium metal casting. In order to directly measure the spatial distribution of 230 Th and 231 Pa in uranium metal before and after vacuum induction melting (VIM), Los Alamos National Laboratory identified uranium metal feedstock, characterized the metal feedstock, conducted a controlled casting experiment of an approximately 120 kg uranium metal rod, and characterized the cast metal. Herein this study presents radiochronometry results and quantified 230 Th and 231 Pa VIM separation factors from bulk uranium.
With the advent of large language models (LLMs), in both the open source and proprietary domains, attention is turning to how to exploit such artificial intelligence (AI) systems in assisting complex scientific tasks, such as material synthesis, characterization, analysis and discovery. Here, we explore the utility of LLMs, particularly ChatGPT4, in combination with application program interfaces (APIs) in tasks of experimental design, programming workflows, and data analysis in scanning probe microscopy, using both in-house developed APIs and APIs given by a commercial vendor for instrument control. We find that the LLM can be especially useful in converting ideations of experimental workflows to executable code on microscope APIs. Beyond code generation, we find that the GPT4 is capable of analyzing microscopy images in a generic sense. At the same time, we find that GPT4 suffers from an inability to extend beyond basic analyses for more in-depth technical experimental design. We argue that an LLM specifically fine-tuned for individual scientific domains can potentially be a better language interface for converting scientific ideations from human experts to executable workflows. Such a synergy between human expertise and LLM efficiency in experimentation can open new doors for accelerating scientific research, enabling effective experimental protocols sharing in the scientific community.
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.
The Advanced Test Reactor (ATR), and complimentary zero-power ATR Critical (ATRC) reactor, located at Idaho National Labs (INL), are undergoing conversion from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU). Both have a variety of testing locations that can receive large variations in flux due to its unique serpentine design, consisting of five lobes (see Figure 1). Initial criticality and power distribution throughout the core are controlled by core-external outer shim control cylinders (OSCCs). Distinct test loops allow for testing at specific temperatures, pressures, and irradiation conditions. The ATR is one of the key nuclear engineering research and testing facilities within the DOE National Laboratory Complex, and the ATRC supports its operation [1]. Currently, the Office of Material Management and Minimization (M3) within the National Nuclear Security Administration of the DOE is working to convert the remaining research reactors, including the ATR, from 93% HEU fuel to 19.75% LEU fuel (LEU) to support non-proliferation [2]. Extensive materials testing at INL and internationally has demonstrated that a high-density uranium molybdenum (U 10Mo) alloy can meet the performance requirements of the remaining high powered research reactors. However, there are many technical challenges to address before the conversion to LEU can be successful, including the accurate characterization of the reactor core physics with LEU fuel. Reactor physics safety evaluations currently use Monte Carlo for the 21st Century (MC21), a continuous-energy Monte Carlo radiation transport code [3]. Existing MC21 models of the ATR and ATRC cores have a validation basis for use in neutronics analyses with HEU fuel. The models are used to support safety analyses that include comparisons to the safety requirements for the reactors. However, the use of the LOWE element in the ATR and ATRC is not currently covered by the current model validation basis. To deploy the new fuel type, extensive computational reactor physics support is necessary to support the use of LOWE in the ATR and ATRC. Therefore, LOWE requires a rigorous validation basis, aligned with that of HEU fuel, that takes advantage of the existing software tools and processes currently used for the ATR and ATRC. The experiment to validate of the MC21 models for determining power, the Power Impact Validation Experiment, will consist of two flux runs in the ATRC, one with fully HEU loading and one with a single LOWE element. Both flux runs will be instrumented with 20 sets of azimuthal fission wires and 3 sets of axial fission wires, as shown in Figure 4. Standard flux run methodology will be used [4]. Power Impact Validation Experiment data will be compared against MC21 calculated data, both for absolute fission rate accuracy and to determine the relative change in fission rates between the two runs. The results of the Power Impact Validation Experiment and subsequent evaluations will provide the validation basis for MC21 for use with LOWE elements. Key features of the Power Impact Validation Experiment include: (1) Two flux runs to allow for LOWE perturbed measurements to be compared to already validated measurements taken from a full core of HEU fuel, (2) Optimization of instrumentation to balance analytical needs with practical considerations (e.g., limited time window to count beta particles from fission products), and (3) Standard ATRC core loading, including both driver positions and flux traps, to minimize cost while remaining representative of typical ATR core loading.
Computer simulations that demonstrate the value of novel approaches are crucial to developing more flexible and robust power systems operations with high penetrations of renewable energy at multiple geographic and temporal scales. However, optimization-based simulations that depend on forecast data often face challenges in evaluating performance, reproducing results, and testing under realistic simulation scenarios. In this paper, we develop scientific computing best-practices for the validation and reproduction of power systems operational models. We then employ two case studies to demonstrate the proposed validation and reproduction framework.
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The severe volumetric changes in Si particles during the Li (de)alloying process cause expansion and contraction of the electrodes, which along with excessive electrolyte reduction and solid electrolyte interphase formation brings about rapid decay in cell capacity. In this work, we use operando electrochemical dilatometry to quantify the (de)lithiation-induced expansion/contraction of Si-based and graphite-rich electrodes during electrochemical cycling. We evaluate the relationship between electrode capacity and dilation and observe that by increasing the Si contribution to the electrode capacity, the swelling is aggravated upon lithiation. For silicon-rich anodes, the electrode dilation can be higher than 300%, and the expansion profile consists of a combination of slow swelling at low lithiation followed by an accelerated increase at higher lithium contents. We investigate how electrode properties, such as porosity, affect the dilation profile and quantify the irreversible expansion of the electrodes. Finally, we discuss some of the challenges associated with the dilatometry technique and suggest experimental approaches for obtaining consistent and reliable data.
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