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DOE OSTI2024

Assessment of Readout Techniques for Passive Monitors

This fiscal year (FY) 2023 report on passive temperature sensors covers two main objectives: to demonstrate that the optical dilatometer can successfully process disc shaped silicon carbide (SiC) temperature monitors (TMs), and to demonstrate proof of concept for using the capacitance readout method to read printed melt wires. The SiC objective was successfully met by annealing and analyzing, via optical dilatometry, all eight 3-mm SiC discs provided by the Nuclear Science User Facilities (NSUF) Idaho State University (ISU) Nanostructured Steels for Enhanced Radiation Tolerance (N SERT) experiment, which was irradiated at Idaho National Laboratory (INL)’s Advanced Test Reactor (ATR). Per the ISU N SERT experiment, capsule 1 (KGT 3828 1 and KGT 3828-2) had a design temperature of 300°C +/- 50°C and an exposure of 2 dpa +/- 10%; capsule 2 (KGT 4600 and KGT 4609) had a design temperature of 300°C +/- 50°C and an exposure of 6 dpa +/- 10%; capsule 3 (KGT 4639 C and KGT 4639-D) had a design temperature of 500°C +/- 50°C and an exposure of 6 dpa +/- 10%; and capsule 4 (KGT 3841 3 and KGT 3841 4) had a design temperature of 500°C +/- 50°C and an exposure of 2 dpa +/- 10%. The target exposure rates, in dpa, are the neutron damage for various types of nanostructured steels. All but three SiC TMs (KGT 4600, KGT 4639 D, and KGT 3841 4) revealed averaged peak irradiation temperatures that fell within the design temperature ranges. The three SiC TMs that did not fall within the design temperatures ranges were at least 100°C below that target temperature. Furthermore, SiC TM KGT 3841 C revealed two irradiation regimes: one closer to the 300°C design temperature, and the other closer to the 500°C design temperature. Also, all the SiC TMs’ averaged peak irradiation temperatures came in anywhere between 20°C and 240°C below the irradiation temperatures predicted by the thermal models. This showed the optical dilatometry method to be a reliable and less time intensive process for determining averaged peak irradiation temperatures from passive SiC TMs such as rods and discs. Under the Advanced Sensors and Instrumentation (ASI) program in FY-23, Boise State University (BSU) proposed to demonstrate proof of concept for using a capacitance readout technique applicable to printed melt wires; however, they were stymied by the complexity of the capacitance readout method. In support of the BSU work, INL developed an additively manufactured (AM) ceramic package for encapsulating the new melt wires. Inks were synthesized at BSU that used new protocols rather than following previously established protocols implemented at INL, and testing of various temperatures was conducted at BSU to evaluate the melting behaviors of the printed melt wires. The result was that the capacitance readout technique showed promise but also created more challenges than originally anticipated. For example, the tin ink synthesized at BSU showed unusual melting behaviors that did nothing to enhance the performance of the final printed melt wire prototype in terms of the capacitance readout method. To make the proof of concept work when applied to the printed melt wires, the ASI program would need to invest further resources and time. Consequently, the program is not planning to continue this proof of concept work in FY-24, based on the progress and findings achieved in FY-23.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Bioenergy Feedstock Library Annual Summary Report 2024

The Bioenergy Feedstock Library (BFL), part of the Biomass Feedstock National User Facility (BFNUF) located at Idaho National Laboratory (INL), is a physical sample repository and a web-accessible electronic database. The BFL stores physical and chemical characteristics of biomass and waste carbon sources for energy use, as well as samples generated from U.S. Department of Energy (DOE) Bioenergy Technologies Office (BETO) and U.S. Department of Agriculture-funded projects. The objective of this Bioenergy Feedstock Library Annual Summary Report for 2024, similar to the 2023 Annual Summary Report , is to focus on the updates to: (1) publicly available analytical data and equipment tracked through the BFNUF, (2) significant increases in the physical samples available for request, (3) sample and data archival progress from recent BETO-funded projects, and (4) publicly available data sets created upon request from BETO, INL projects, or outside entities compared to the previous annual summary reports. This report highlights key statistics and available data and information important for INL, BFL users, academics, and industry.

09 BIOMASS FUELS↗
DOE OSTI2023

Advancements in Development and Testing of Thermal Power Dispatch Simulators

Flexible plant operations and generation (FPOG) offer nuclear power plants (NPPs) the chance to leverage alternative, non-electric revenue streams while ensuring their continued role as reliable, clean, and constant sources of baseload electrical power. The excess thermal energy generated from NPPs during periods of low electricity demand can be channeled as raw materials to numerous industrial processes via a thermal power dispatch (TPD) system. Hydrogen production via high-temperature steam electrolysis (HTSE) is an optimal use case based on technical and economic feasibility. Researchers at Idaho National Laboratory (INL) have conducted previous works that developed and implemented TPD system models within the GSE Solutions Generic Pressurized Water Reactor (GPWR) simulator to support human-in-the-loop (HITL) scenario-based evaluations. The first part of this report documents modifications made to the GPWR TPD model and HMI from the previous iteration in line with a new Sargent and Lundy (S&L) TPD design with an automatic control system. The was done in collaboration with Westinghouse using their three-loop pressurizer water reactor (W3LPWR) simulator which contains an industrial grade automatic control system for the TPD. This was installed in the Human Systems Simulation Laboratory (HSSL) at INL. The second part of the report documents findings from an all-hands-on-deck integration and verification workshop that was conducted in the HSSL over several days. The research team comprised INL human factors and TPD experts, a nuclear engineer from GSE Solutions who implemented the revised TPD model for GPWR, the human-machine interface (HMI) prototyping and human factors team from the University of Idaho, and personnel with operations experience with pressurized water reactors. The workshop provided time and expertise to conduct the final activities to bring the operations, HMI, and simulator into a functional state. The goals of the integration and verification workshop were: 1. to install the revised GPWR TPD model into the HSSL 2. verify the TPD HMI prototype was functional 3. integrate the HTSE Simulink model to GPWR. 3. Issues were identified for resolution, but overall the workshop accomplished its goal to integrated and verify the majority of the intended functional. Future work will resolve the identified issues and use the integrated simulation to support an evaluation and demonstration in the next fiscal year.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Test Plan for DOE-GENIORS Dyanamic Irradiation Tests

Dynamic irradiation tests will be performed within the DOE-GENIORS collaboration, to be finished by the end of March 2021. These tests will be performed using three different setups, one at INL, one at CEA, and one at CIEMAT. The cis-mTDDGA/PTD chemical system will be used, applying both "extraction" (cis-mTDDGA - n-dodecane/HNO3) and "stripping" (cis-mTDDGA - n-dodecane/PTD - HNO3) conditions. This test plan provides the details and parameters to be used to perform irradiation at the INL using the INL Radiolysis Test Loop.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

NCERC-CL TREAT Free Field Characterization Foil Reaction Rate Results

This report summarizes the analysis and preliminary MCNP reaction rate modeling of free field characterization measurements performed in the Transient Reaction Test Facility (TREAT) at Idaho National Laboratory (INL) between 12/18/2023 and 2/22/2024. A series of seven irradiations were performed using the Big-BUSTER (Broad Use Specimen Transient Experiment Rig) core configuration. Foil sets including 19.5% LEU-Zr alloy wire, S, Au, Fe, Ni, Co, Ti, and Zr were deployed in each irradiation. Identical foil sets were supplied to Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and INL. All foils were supplied by INL. Additional details on the experiment can be found in [2]. The foil sets were shipped to the National Criticality Experiments Research Center Counting Laboratory (NCERC-CL) branch at LANL (NCERC-CL NISC) and were received on 3/25/2024. Additional details on receipt and counting are reported in [3]. All reactor dosimetry measurements were performed in adherence to the ASTM standards applicable to reactor dosimetry. Measured reaction rates from Au, Co, Fe, Ni, and Ti are reported. Select reaction rate ratios and simulated reaction rates and ratios are discussed. Reaction rates for all measured reaction products were modeled with MCNP 6.3.1 and a TREAT input deck supplied by Edward Lum. Neutron emission estimates calculated using the measured and simulated reaction rates show excellent agreement between the three fast-threshold n-p reactions with the percent differences being < 9% between the three reactions. The capture reactions had poorer, but still reasonable, agreement with < 20% percent differences between the neutron emission estimates of the three capture reactions. The percent difference between the fast-threshold and capture reactions is ∼ 100% for each reaction. This supports the TREAT model is not accurately modeling the neutron spectrum and additional measurements are required to characterize TREAT and match measurements to simulation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Finding ways to reduce nuclear waste: searching for the unknown one step at a time

In my home country, Venezuela, research has been stagnant. Due to the political turmoil and the crisis, many educated people have left the country in search for a better life. This has caused a deficit in any technological and scientific advances, making Venezuela one of the first South American countries to have its rate of publications decline by 29% in 2013. Currently, Venezuela lacks the infrastructure and the means to keep up with the research progress as compared to other countries in South America, such as Brazil. Since coming to the United States (US), and currently working for a national laboratory, the active research environment endorses a wide range of careers and engineering programs that allow researchers to thrive at any given field. Researchers have access to funds and tools to succeed in developing materials for the future. There are 17 national laboratories in the US, and all of these have a different research focus/objective. As examples, Los Alamos National Laboratory and Sandia National Laboratory focus is on national homeland security, weapon science, radiation effects, among others. Argonne National Laboratory focuses on nuclear energy, energy storage, high performance computing, etc. At Idaho National Laboratory (INL) the research focuses on innovating nuclear energy and clean energy resources, critical infrastructure materials, along with fuel cycle solutions to manage, dispose and find ways to recycle current and future radiological waste. Compared to other national laboratories, INL focuses slightly more on applied processes and how nuclear energy can be innovated to next reactor design and technologies. The research being conducted at INL made me apply for a Seaborg distinguished postdoctoral position. For the position itself, the researcher must submit a proposal related to actinide chemistry on a research field area. In this position, 50% of my time will be focused on my own proposal. The proposal that I am working on is focused on the innovation of nuclear energy and fuel cycle recycling, which is why I was mainly interested on working at this national laboratory. To give a bit more context of what my proposal is about, a little bit of background is necessary: After the nuclear fuel (UO2) is used in a reactor, the fuel matrix is then characterized by various fission products (FP). Among these FP (including rare earth elements, alkali/alkaline earths, and actinides), many can potentially be recovered through nuclear reprocessing technologies. In pyroprocessing, the used nuclear fuel undergoes electrochemical dissolution into a molten chloride salt mixture in an electrorefiner. Initially, uranium is reduced onto an inert cathode by applied potentials. However, numerous remaining FPs accumulate in the melt and pose challenges for recovery by an inert electrode, particularly the rare earth elements (e.g., Nd, Gd, Pr, Sm) due to their multivalent oxidation states and tendencies toward side reactions, leading to their dissolution in the electrolyte. These recovery challenges result in inefficiencies and necessitate the continual discarding of the molten chloride salt, thereby generating additional waste. Furthermore, the presence of rare earth elements and other fission products in the molten salt electrolyte alters its physical and chemical properties, affecting both uranium recovery efficiency and the longevity of the molten chloride salt. To improve the recovery efficiency of the FP, specifically rare earth elements, I am investigating the fundamental interactions between rare earth elements in the molten chloride salt and their metallic form. The kinetic pathways and the chemical reactions of these elements will give insights on how the recovery efficiency can be improved. The interactions and speciation of these elements are being studied by spectro-electrochemistry at high temperature environments in quartz and other ceramic materials (e.g., alumina crucibles). Some of the challenges I am facing specifically relates the reactivity of some of these elements with different glass and crucible materials. Although my research focuses on fundamental science, it will benefit the applied process by generating new scientific knowledge and closing the gap for an efficient recycling of the waste: one step at a time.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

In-Field Alpha Spectrometer Development FY2022 Mid-Year Report

In response to needs identified by the International Atomic Energy Agency (IAEA), Idaho National Laboratory (INL) has developed an In-Field Alpha Spectrometer (IFAS) to allow IAEA safeguards inspectors to collect samples of uranium hexafluoride (UF 6 ) at processing facilities, to perform field assessments to verify uranium enrichment. For sample collection, the IFAS method uses Single-Use Destructive Assay (SUDA) samplers, developed at Pacific Northwest National Laboratory (PNNL), which contain thin zeolite coatings that trap UF6 gas and convert it to the safer, more stable form uranyl fluoride (as a dihydrate, UO 2 F 2 ·2H 2 O). For alpha spectrometry, the IFAS instrument employs a large area silicon semiconductor transducer to detect and record alpha particle energy-deposition events. Over the course of this project INL, PNNL, Oak Ridge National Laboratory worked to optimize the SUDA sample design for alpha spectrometry, to optimize processes for manufacturing, loading, and shipping SUDA samples, and to allow the measurement of loaded SUDA samples. In fiscal year (FY) 2022 INL has bene investigating the reproducibility and measurement precision of IFAS measurements and the associated determination of uranium enrichment.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2024

Crossroads of Nonproliferation and Safeguarding Technologies for Implementation in Molten Salt Reactors

Idaho National Laboratory (INL) recently conducted a workshop endorsed by the National Nuclear Security Administration (NNSA) under the auspices of the Defense Nuclear Nonproliferation R&D (DNN R&D) Office's Safeguards Portfolio. The workshop's primary objective was to foster an engaging dialogue among researchers, with a specialized emphasis on the safeguards pertaining to molten salt reactors. At INL, the installation of the state-of-the-art Molten Salt Thermophysical Examination Capability (MSTEC) is underway. This shielded argon glovebox facility, designed for both irradiated and non-irradiated actinide materials, represents the cutting edge of research infrastructure. MSTEC is poised to serve as a pivotal research platform in the realm of molten salt technology, with significant implications for safeguards applications. Participants of the workshop had the opportunity to tour the facilities, including the site where MSTEC is being installed, as well as to observe the INL's molten salt and pyroprocessing research hot cells. The event featured insightful presentations delving into molten salt chemistry and the MSTEC project. Each participating laboratory contributed to the discourse with presentations on their respective research efforts addressing safeguards in relation to molten salt reactors. The workshop culminated with a generative brainstorming session, where participants shared their thoughts on strategic integration with partner agencies, aiming to synergize efforts in advancing the field of nuclear safeguards.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of Engineered Form MOFs CaSDB and HKUST-1

Idaho National Laboratory (INL) and Pacific Northwest National Laboratory (PNNL) have studied different sorbents for the purpose of separating krypton (Kr) and xenon (Xe). PNNL has researched various metal-organic frameworks (MOFs) that have characteristics theoretically suitable for the separation. Based on computational analysis, PNNL chose Calcium–4,4’–sulfonyl dibenzoate (CaSDB) and Hong Kong University of Science and Technology MOF1 (HKUST–1). The raw MOFs are supplied in a powder form and must be incorporated into a robust engineered form to be viable for gas separation. In 2016, PNNL produced an engineered form of CaSDB by compressing and grinding the raw MOF material. This form was sent to INL for adsorption testing. Gentle mechanical agitation of the compressed form produced a large amount of powder, resulting in the conclusion that the form was not viable for the application. In 2019, INL successfully produced two engineered forms, CaSDB–EF1 and HKUST–EF2, by incorporating the active MOF powders into a polyacrylonitrile (PAN) binder. PNNL also produced two engineered forms, CaSDB–10pmma and HKUST–10pmma, by using polymethyl methacrylate (PMMA) as a binder. Surface area analysis indicated that the forms could be viable for separations in which the CaSDB–EF1 retained 84% of the raw CaSDB surface area and CaSDB–10pmma retained 48%. HKUST–EF2 retained all the raw HKUST–1 surface area, whereas HKUST–10pmma retained only 28%. Mechanical stability testing demonstrated that the PMMA forms formed a fine powder whereas, the PAN forms did not. In general, adsorption testing showed that CaSDB–10pmma had a lower Xe breakthrough capacity than CaSDB–EF1, but a higher saturation capacity at both room temperature and at 253 K. The CaSDB MOF was intended to be used as a room temperature sorbent to capture Xe. However, both CaSDB engineered forms displayed relatively short initial break through times. Reduced temperature increased the Xe capacity of both engineered forms. HKUST–1 was expected to be effective only at sub-zero temperatures. At 195 K, HKUST–10pmma had an initial Xe breakthrough capacity of 119 mmol/ kg, compared to HKUST–EF2 at 339 mmol/ kg. The saturation capacities were 275 and 383 mmol/ kg, respectively. HKUST–10ppm also had a lower Kr capacity of 1.7 mmol/kg, whereas the HKUST–EF had a Kr capacity of 3 mmol/ kg. Test results indicate that the PAN engineered forms are a more suitable choice for the complete separation of Xe and Kr, due to the increase in initial breakthrough capacity. CaSDB allows rapid breakthrough of Kr and Xe regardless of the type of engineered form. At 253 K, it is more viable, but is still not the most effective sorbent. HKUST–EF2 demonstrated viability for Xe capture at 195 K, which may merit further study, however it may be cost prohibitive.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Towards Net Zero: Modeling Approach to the Right-Sized Facilities

As the concentration of greenhouse gases (GHGs) in the atmosphere increases, the concerns about carbon emissions are growing. Several net-zero initiatives are taking place around the globe to achieve a balance between the GHGs put into the atmosphere and those taken out. While most efforts present a sectorized approach, this paper describes the importance of integrating information across different sectors for effective modeling of carbon emissions and holistic reduction opportunity analysis. Using the Idaho National Laboratory (INL) campus as a test case, this work provides a web-based tool for INL stakeholders to use when engaging in strategic planning to achieve carbon emissions reduction. This net-zero engineering support tool (NEST) uses historical data as foundational information for applying the modeling framework. Prediction of CO 2 emissions throughout project completion integrates various approaches and schedules aimed at energy conservation, fleet decarbonization, and other GHG reduction activities. Using NEST, stakeholders can visualize carbon emissions, electricity consumption, and costs for decision making when planning the pathway for reaching carbon net zero. The INL’s initiative to transition into an EV fleet was used to demonstrate the developed framework and the advantages of using NEST. It was shown that electrifying different fossil-fueled campus vehicles before 2030 with aggressive replacement schedules require high annual capital expenditure (CAPEX), which may not be available. The tool allows decision makers to test different replacement schedules and prioritize those that yield CAPEX below a certain threshold while meeting target milestones. While the demonstration focused on vehicle electrification, the developed framework lays the foundation for further quantitative analysis of other GHG reduction activities.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2020

Modeling Time-Dependent Surrogates of Additive-Manufactured Nuclear Fuels Processes

Additive manufacturing (AM) technology is being increasingly adopted in a wide variety of application areas for its ability to rapidly produce, prototype, and customize designs. Recently, a hybrid AM technique was successfully developed at Idaho National Laboratory (INL) to manufacture nuclear fuels [1]. Despite the advantages, this AM technique needs optimization due to defects from a highly complex melting and sintering process. The complex metallurgical phenomena during AM processes are strongly related to parameters such as applied laser power, traveling speed, and scan style, which could lead to differences in density, residual stress, crystallographic texture, and mechanical properties. In addition, stochastic variations in laser energy interaction and associated multiscale/multiphysics phenomena cause variations in microstructure evolution and mechanical properties. Currently, researchers at INL are focusing on developing a comprehensive modeling framework, leveraging INL’s simulation tools MOOSE/MARMOT/BISON/RAVEN [2-4] to describe all steps of this AM process across multiple length scales. Although this advanced framework plays a critical role in enabling enhancements to traditional trial and error approaches for design and optimization of nuclear fuel materials, it remains computationally intense, limiting its use in sensitivity and optimization analysis. In this case, an accurate and inexpensive surrogate becomes an effective tool for providing a tractable approximation of the underlying underline physics. Surrogate models generally not based on the physics of a system are purely mathematical models used to capture the relationships between specific system inputs and outputs. Popular approaches, including neural networks [5], response surfaces [6], and subspace-based reduced order models [7], have been applied to a wide range of disciplines, such as nuclear reactor design, aerospace design and automotive design. In this summary, we employ advanced time-dependent surrogate models such as high-dimensional model representation (HDMR) [8] and physics-informed deep neural network (PINNs) [9] to accelerate the design and optimization of AM process.

42 ENGINEERING↗
DOE OSTI2020

ASNF Engineering Scale Drying Experiment Acceptance Criteria for Fabricated Items

This document specifies the acceptance criteria for items fabricated by Holtec International under the scope of INL Contract # 215725, consistent with the parameters described in INL/EXT-19-56017, Aluminum-clad Spent Nuclear Fuel Engineering Scale Drying Experiment Design. Verification is confirmed with initials (or the need for remediation may be identified) along with the date of inspection for each requirement. Verification may be performed by independent proxy (University of South Carolina participant, under INL Contract Number 207046) or by an INL employee, remotely by camera images.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

DOE Standard Canister Loading Methodology Poster

Over 250 different types of spent nuclear fuel (SNF) are managed by the U.S. Department of Energy (DOE) at the Idaho National Laboratory (INL). The current configuration of SNF at INL is not transportable off-site. In the Yucca Mountain Repository License Application, the SNF was to be loaded into DOE Standard Canisters which are capable of storage, transportation, and eventual disposal using the proper overpacks. The DOE Standard Canister will allow INL to place the SNF in a Road-Ready configuration in preparation for transportation off-site. The focus of this work is development of a general loading methodology for the DOE Standard Canister. This methodology is to be general enough that it will apply to all SNF to be loaded into the DOE Standard Canister. The main sections of the methodology to be outlined are prerequisites before loading (including any issues with materials such as inserts or baskets), loading of SNF, and canister closure and quality assurance (including welding and inspection). Each subsection will analyze and provide the bases for each step.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

FY20, 3rd Quarter, Multi-pass Hybrid Laser Arc Welding of Alloy 740H

Although progress has been slowed by telecommuting restrictions at INL and PSU, progress has been made by PSU on modeling and simulation of deep penetration laser welding. Additionally, both INL and PSU has received critical equipment, in the form of high-powered laser wobble heads, needed to verify the modeling and simulation results. The equipment was installed at INL and preliminary tests, based on a test matrix that was developed in this 3rd quarter, will be made early in the 4th quarter. Similarly, the PSU laser wobble head is expected to be installed early in the 4th quarter and preliminary experiments are expected to commence shortly thereafter. Completion dates on a few milestones have been impacted and new dates proposed.

36 MATERIALS SCIENCE↗
DOE OSTI2021

NRIC Final Water Management Options

The purpose of this paper is to inform advanced reactor developers of the conditions and requirements needed to manage the water used to reject heat from advanced reactor demonstrations at the INL. While there are many advanced reactor concepts, all of them require heat rejection, and some plan to use water for hear rejection. Cooling water for heat rejection is frequently the largest fresh-water load for a nuclear reactor, and therefore the focus of water management. The type of reactor and its purpose, the physical conditions at the INL, and the regulatory structure of the demonstration also influence the optimal method of heat rejection. Choosing the best method of rejecting heat will significantly influence the cost and probability of success for any advanced reactor demonstration. One of the important considerations in the deployment of demonstration plants at the INL site is water use and its implications. Advanced reactor developers need to consider the limitations of the water consumption based on the state and federal regulations. The water consumption may become a limitation when the demonstration reactor capacities reach to conventional nuclear power plants.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

Graphite Oxidation Studies Irradiated Graphite, GIF Summary, ASME Task Force, ASTM Activities

Introduction Comparison of Graphite Oxidation Rates for Irradiated and Unirradiated NBG-25 (Rebecca Smith and Austin Matthews, INL) Annealing effect over range of oxidation temperatures Irradiation effect (~6.5 dpa) over range of oxidation temperatures Dose dependency of irradiation effect Generation IV International Forum High Level Deliverable Summary of DOE ART Historic Oxidation Research (Cristian Contescu, ORNL) Publications issued over the last ~10 years Acute and chronic oxidation ASME Graphite Oxidation Task Force (C. Contescu, Paul Ryan, ORNL, Joe Bass, A. Matthews, R. Smith, INL) Collaborative effort headed by ORNL and engaging INL and others Initial review of relevant oxidation discussion in ASME code ASTM Activities (C. Contescu) Virtual Meetings New and Revised Standards

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

DE-FOA-0002687 - Request for Information on Industrial Decarbonization Priorities

Idaho National Laboratory (INL) collaborates with industrial, academic, FFRDC, and research organizations across all of the categories highlighted in the RFI. INL is cognizant that industry and research organizations provided detailed RFI responses in the categories aligned with their sector. INL’s RFI response outlines crosscutting principles and specific challenges to enable DOE to achieve industrial decarbonization targets across all categories.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2021

Phase Stability of Yttrium Titanium Oxide

This proposed study aims at elevating Idaho National Laboratory (INL)’s portfolio in nuclear material research on two major fronts: 1) expanding INL’s material characterization capabilities to provide holistic microstructural and microchemical characterization ranging from atomic scale to millimeter scale, and 2) filling the knowledge gap on the effect of microstructural evolution on material phase stability at extreme conditions resembling those inside nuclear reactors, using Yttrium-Titanium-Oxide (YTO) as a model material. Previous research indicated the kinetics of phase transition of YTO materials varies between bulk samples and powder samples, suggesting that microstructural characteristics such as defect, micro-strain, and crystallite size could play a significant role in phase transition. The effect of crystallite size on the thermodynamics of phase transition has been well documented for other ceramic materials at nanometer crystallite size, with potential impacts from surface and strain energy and defects. In this proposed study, we will perform microstructural (micro-strain, crystallite size, defects, phase) and microchemical (elemental mapping) analyses through in-situ high temperature measurements using X-ray diffraction (XRD; at millimeter scale for powder and bulk samples) and transmission electron microscopy (TEM, at nanometer scale and atomic scale for powder and thin foil samples), to provide a comprehensive view of the effect of microstructural and microchemical evolution on phase transition and phase stability of YTO materials. These findings will also establish a roadmap and provide new insights on the phase stability of other materials that are critical to nuclear energy and other renewable energy infrastructure. Besides advancing our understanding of this fundamental material science question, this project will also enable the development of state-of-the-art material characterization capabilities at INL/MFC for microstructural and microchemical analysis at high temperature conditions. The new collaboration between the High-Resolution Materials Characterization group at INL and the Electron Microscopy group at the University of Florida (UF) in this project will open the possibilities for more fruitful and synergetic cutting-edge nuclear material research in the future.

36 MATERIALS SCIENCE↗