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DOE OSTI2023

Advanced Reactors Integrated Energy System: Thermal Energy Storage Island Design

The main topic of this research is integrated energy systems (IES) designed for pairing industrial thermal energy loads with advanced reactors (ARs). The Idaho National Laboratory (INL) Crosscutting Technology Development IES program and the National Reactor Innovation Center (NRIC) are seeking to develop, design, and construct an AR-IES demonstration platform that couples the thermal output from an AR operating at the INL/NRIC Demonstration of Microreactor Experiments (DOME) test bed in the Experimental Breeder II dome to a variable capacity load emulator (i.e., air-cooled radiator) and sensible thermal energy storage (TES) via a molten salt thermal energy transfer fluid. In the rapidly evolving landscape of energy supply and distribution, flexibility has emerged as a prized attribute, surpassing the traditional notions of stability and baseload generation capability. This shift in priorities is particularly evident in the context of nuclear power plants (NPPs), where adaptability over constant output is becoming more important. As our energy infrastructure and resources embraces the rise of distributed energy generation, the inherent variability in net demand continues to grow. Moreover, the use of nuclear energy as a source of heat for decarbonizing the industrial sector is becoming a very pressing topic. In such environment, advanced NPPs are poised to enter a more competitive energy market, delivering both, flexible electricity and heat. This shift motivates the exploration of TES systems, designed to empower NPPs with nimble responsiveness to market fluctuations, flexible heat delivery capabilities, and redefine their role in the energy field. TES systems offer the unique advantage of storing nuclear energy in its original form as heat, thereby affording unparalleled flexibility in its subsequent utilization.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Drop Analysis of Department of Energy Standard Canister with Fort Saint Vrain SNF

DOE manages over 300 types of SNF, most of which are located at the INL site. The Idaho Cleanup Project and INL are collaborating on the Road-Ready Capability Demonstration Project, which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF at the INL site for ?road-ready dry storage?. ?Road-ready dry storage? is a SNF management concept where SNF is packaged into dry and sealed canisters, which are then placed in on-site storage in anticipation of later transportation. The forward-looking goal of the Demonstration is establishing the foundation for a large-scale road-ready dry storage program at the INL site. The Demonstration will first package Fort Saint Vrain SNF currently stored at INL into several DOE Standard Canisters. These Standard Canisters will then be loaded into another commercial transportation or storage containment system (e.g., storage cask or transportation package). The Standard Canister is a class of standardized canisters designed for containing the large variety of DOE-managed SNF during interim storage, transportation and/or disposal at a geological repository. One critical aspect of road-ready dry storage is the ability to license the DOE Standard Canisters and its associated transportation package to 10 CFR 71. Depending on the SNF and transportation strategy, the Standard Canisters may have to maintain structural integrity under normal conditions of transport and hypothetical accident scenarios (i.e., drop events). The Standard Canisters have been tested and analyzed under various SNF loading configurations and accident drop events in support of the Idaho Spent Fuel Facility and other DOE programs. However, no analysis has been completed to support the recent Demonstration. This analysis will consider the Ø0.5 m × 5.1 m Standard Canister under drop scenario(s) considered in previous INL tests and analyses, including the 9 m drop at 80 degree off vertical. However, this analysis will consider the more recent Fort Saint Vrain loading configurations proposed for the Demonstration. This analysis will performed using strain-based acceptance criteria established by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 3. It will be compared to previous analyses and form the foundation of further formal calculations that will be used to support licensing efforts of the road-ready dry storage system at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Technical report Letter: RAFM, ODS steels and MMLC for Nuclear energy application

The lifetime, thermodynamic efficiency, safety and economic viability of new generation fission and fusion reactor concepts can largely be tied to the mechanical performance and stability of structural alloys under extreme environments. In this context, engineered nano materials could have broad-reaching impact on the future of advanced nuclear fuel-cycle and reactors. These systems are characterized by a large number density of interfaces which are efficient sinks for point defects and moderately biased; therefore limiting the deleterious effects of irradiation. Broadly, nuclear nano-technology deals with the use of the latest engineered-nanomaterials for improving the nuclear power performances and safety in all areas of nuclear energy production to bring new generations of nuclear power units. New advanced fuel assembly designs also have implications for securities and safeguards. To support the readiness for potential future license applications, an understanding of the technologies that would enable new reactor designs in the areas of component performance and domestic safeguards is necessary. This technical report letter work explores the technical issues and potential regulatory considerations associated with developing and adopting fuel claddings made of advanced nano- materials. Specifically three classes of nanomaterials are considered: (i) reduced activation ferritic/martensitic (RAFM) steels, (ii)oxide dispersed steels (ODS) and (iii) multi-metallic layered composites (MMLC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Recent Advances in Picochlorum renovo Strain Development

Microalgae are amongst the most efficient phototrophs for the reduction of CO2 to industrially relevant products and product intermediates. We have recently identified a novel species in the genus Picochlorum (Picochlorum renovo) that demonstrates high areal productivity (>30 g/m2/day), high temperature and salinity tolerances, and rapid growth rate (-2 hour doubling time): attributes essential for economically viable industrial scale deployment for low value commodities (e.g. fuel and chemical precursors). Following screening of >300 algal isolates to identify Picochlorum renovo, we developed baseline genetic tools necessary for genetic engineering of the nucleus and chloroplast, and further built upon these tools to establish a protein secretion system. We have also deployed genetic engineering tools to establish the utilization of phosphite as a selectable marker in the chloroplast and nuclear genomes, while concurrently acting as a potential crop protection strategy. We have further implemented these tools to enable the utilization of formate as a novel carbon source for phototroph cultivation. Explicitly, formate toxicity was assayed, followed by introduction of a formate dehydrogenase to allow formate utilization. Cultures supplemented with formate grew to a higher density when cultivated under ambient CO2, highlighting the potential for this strategy to increase growth. Additionally, we have established a synthetic transcription factor and core promoter system which lays the foundation for tunable, high expression engineering in the nuclear genome, and potentially can be applied for universal algal genetic engineering. Finally, current work has focused on direct photosynthetic production of chemicals, and generation of mutant libraries. Combined, our work has established a robust genetic toolbox for Picochlorum renovo, which we have deployed for strain development purposes to introduce biotechnologically relevant traits.

algae↗
DOE OSTI2021

Materials and Fuels Complex Five-Year Mission Strategy (FY21- FY25)

The Idaho National Laboratory (INL), through its designated mission of advancing innovative nuclear energy solutions, is actively engaged in the research, development, demonstration and deployment of advanced nuclear technology, as well as in fostering private-public partnership for technology development. Key to the success of INL’s mission is the Materials and Fuels Complex (MFC), the only complex in the U.S. that hosts a world-class assemblage of facilities, capabilities and instruments for handling, testing, and characterizing radioactive materials. Driven by its mission/vision of “Engineering and Experiments that Drive the World’s Nuclear Energy Future,” MFC is at the center of INL’s – and indeed the Department of Energy’s – advanced nuclear technology development initiatives, providing essential capabilities such as engineering-scale high-assay low-enriched uranium (HALEU) fuel production, reactor demonstration facilities, post-irradiation examination, and transient irradiation testing. Furthermore, MFC provides an ideal environment for test beds that are utilized for research, development and demonstration (RD&D) activities on used fuel treatment, nuclear non-proliferation, forensics, and nuclear power sources used for space exploration missions conducted by the National Aeronautics and Space Administration (NASA).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Materials and Fuels Complex (FY2021-2025 Five-Year Mission Strategy)

The Idaho National Laboratory (INL), through its designated mission of advancing innovative nuclear energy solutions, is actively engaged in the research, development, demonstration and deployment of advanced nuclear technology, as well as in fostering private-public partnership for technology development. Key to the success of INL’s mission is the Materials and Fuels Complex (MFC), the only complex in the U.S. that hosts a world-class assemblage of facilities, capabilities and instruments for handling, testing, and characterizing radioactive materials. Driven by its mission/vision of “Engineering and Experiments that Drive the World’s Nuclear Energy Future,” MFC is at the center of INL’s – and indeed the Department of Energy’s – advanced nuclear technology development initiatives, providing essential capabilities such as engineering-scale high-assay low-enriched uranium (HALEU) fuel production, reactor demonstration facilities, post-irradiation examination, and transient irradiation testing. Furthermore, MFC provides an ideal environment for test beds that are utilized for research, development and demonstration (RD&D) activities on used fuel treatment, nuclear non-proliferation, forensics, and nuclear power sources used for space exploration missions conducted by the National Aeronautics and Space Administration (NASA).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

Draft ASME Boiler and Pressure Vessel Code Cases and Technical Bases for Use of Alloy 617 for Constructions of Nuclear Component Under Section III, Division 5

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code currently only allows five materials for use in construction of nuclear components for high temperature service. These are: 2.25Cr-1Mo and V-modified 9Cr-1Mo steels, Types 304 and 316 stainless steels and the high nickel Alloy 800H. Since 2005, the US high temperature gas-cooled reactor program has been characterizing elevated temperature mechanical properties of Alloy 617 as the leading candidate construction material for the intermediate heat exchanger. After analysis of these experimental results, along with historical data and additional results available through the Generation IV International Forum, Very High Temperature Reactor, Materials Program Management Board Materials Handbook, a draft ASME Code Case to allow nuclear construction with Alloy 617 for temperatures up to 1750°F (954°C) has been developed. This report contains the Code Case for Low Temperature Service Construction of Section III, Division 5, Subsection HB, Subpart A, Class A and Subsection HC, Subpart A, Class B components, which has been approved in Section II, Materials, and Section III, Rules for Construction of Nuclear Facility Components. Supporting technical justification for the low temperature Code Case is also included. This Code Case allows use of Alloy 617 up to 800°F (425°C). This report also contains an updated draft of a Section III, Division 5, Subsection HB, Subpart B, Class A Code Case for Alloy 617 to qualify it for use in construction of nuclear components up to 1750°F (954°C) for service life up to 100,000 hours. The draft contained in Appendix 4, subject to editorial revision and approval by the ASME Special Task Group on Alloy 617 Code Qualification, will be submitted for approval by letter ballot by the appropriate ASME Committees. The technical justification supporting the Code Case is presented in Appendix 5 of this report. This background document is part of the information package that will be submitted with the Code Case for ballot.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Krypton Concentration using HZ-PAN

Idaho National Laboratory (INL) has developed and tested engineered sorbents to separate and capture volatile fission products such iodine, xenon (Xe), and krypton (Kr) from off-gas streams. As noble gases, Xe and Kr can be difficult to capture and separate. Historically, cryogenic distillation has been used to execute the separation. However, performing this separation from bulk air streams is expensive and can pose significant hazards. INL has successfully developed two different sorbents for Xe and Kr capture, silver mordenite polyacrylonitrile and hydrogen mordenite polyacrylonitrile (HZ-PAN). Adsorption studies to date successfully separated and captured Kr from carrier gas streams, but those studies focused primarily on initial separation and capture. Successful utilization in a used nuclear fuel (UNF) reprocessing facility, however, requires further concentration of the Kr to minimize long term storage volumes. This study focuses on the concentration of Kr utilizing HZ PAN as the concentrating media. Concentrating Kr during desorption is ideal for two reasons: one, to produce Kr that can be used for commercial and research applications, or two, to minimize the volume of the radioactive gaseous waste stream for disposal. This report investigates the Kr concentrating potential of HZ-PAN by running multiple Kr adsorption-desorption cycles, collecting the desorbed effluent from a saturated HZ-PAN column, and loading it onto a fresh column. This study demonstrates that four adsorption-desorption cycles can transform a 150 ppmv Kr stream into a 37.5% Kr stream, an overall concentration factor of 2497. Each concentration step results in successively smaller volumes desorbed, reducing volume by a factor of approximately 40,000. Depending on the operational goals, during desorption one could collect smaller volume fractions of high concentration Kr (~ 67%). This demonstration should be considered proof-of-concept. Further refinement is necessary to develop optimum operating schemes to integrate into UNF off-gas treatment.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2026

Pixelated plastic scintillator array manufacturing using fast-, photo-curable resin

Pixelated plastic scintillator arrays can serve as high efficiency and high resolution neutron imaging detectors. Manufacturing these arrays is intensive in both time and labor. This article presents a fabrication method based on additive manufacturing for two-dimensional plastic organic scintillator arrays using a custom-built automated assembly machine and a custom photocurable resin that has significant non-aromatic acrylate oligomer content. The process involves two main stages: fully autonomous production of one-dimensional layered arrays, followed by semi-autonomous cutting and stacking to form two-dimensional pixel arrays. One-dimensional arrays were manufactured at a rate of around 4 layers per hour with minimal defects and tight dimensional tolerances, while two-dimensional arrays up to 7 x 7 pixels and 70 mm in length were completed in approximately 3.5 hours. Final arrays exhibited dimensional deviations of less than 0.5 mm. Two-dimensional arrays read out by a multi-anode photomultiplier tube demonstrated per-pixel position resolution and pulse-shape discrimination, enabling gamma–neutron interaction separation in mixed radiation environments.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Development of Radiation and Fire Resistant Polyurethane Foam in Support of Deactivation and Decommissioning [Poster]

FoamBag™ is a two-part polyurethane foam that has been down selected as a possible engineering solution for the decommissioning of radiologically contaminated pipes. Our main objective consists of establishing an internal barrier (plug) to mitigate the release of residual contamination prior to cutting operations and required pipework for deactivation and decommissioning projects. To ensure worker safety, we will also be analyzing hazards associated with the off gasses present during curing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

AI-based design of a nuclear reactor core

The authors developed an artificial intelligence (AI)-based algorithm for the design and optimization of a nuclear reactor core based on a flexible geometry and demonstrated a 3× improvement in the selected performance metric: temperature peaking factor. The rapid development of advanced, and specifically, additive manufacturing (3-D printing) and its introduction into advanced nuclear core design through the Transformational Challenge Reactor program have presented the opportunity to explore the arbitrary geometry design of nuclear-heated structures. The primary challenge is that the arbitrary geometry design space is vast and requires the computational evaluation of many candidate designs, and the multiphysics simulation of nuclear systems is very time-intensive. Therefore, the authors developed a machine learning-based multiphysics emulator and evaluated thousands of candidate geometries on Summit, Oak Ridge National Laboratory’s leadership class supercomputer. The results presented in this work demonstrate temperature distribution smoothing in a nuclear reactor core through the manipulation of the geometry, which is traditionally achieved in light water reactors through variable assembly loading in the axial direction and fuel shuffling during refueling in the radial direction. The conclusions discuss the future implications for nuclear systems design with arbitrary geometry and the potential for AI-based autonomous design algorithms.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Installation of Lysimeters and Monitoring Wells in Support of the Revised Approach to E-Area Performance Monitoring

Savannah River Site (SRS) has developed a revised approach to performance monitoring of the E Area Low Level Waste Facility (ELLWF) that adds a saturated zone component to the existing vadose zone monitoring program (Kubilius, 2019). The revised approach specifies modifications to the vadose zone monitoring program, the addition of new lysimeters at Engineered Trench 3 (ET3), and the installation of water table monitoring wells to complement the existing well network surrounding the ELLWF. This project summary report is intended to provide the final construction details and layout for the new vadose zone lysimeters at ET3 and the water table monitoring wells installed as part of the new saturated zone component of the revised performance monitoring program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Completion document for MRT 8824 - Upgrade NIF’s gaseous radiochemistry diagnostics (RAGS) to support weapons science

This milestone highlights the successful upgrade and deployment of the Radiochemical Analysis of Gaseous Samples (RAGS) system, which delivers high-quality measurements (<20% uncertainty) of activated gaseous species from NIF implosions. These measurements are critical for supporting Stockpile Stewardship Program (SSP) relevant platforms, including LANL’s Double Shell and LLNL’s Pushered Single Shell campaigns. The RAGS diagnostic technique enables analysis of short-range mixing in implosions using high-Z shells, which are otherwise inaccessible to conventional x-ray diagnostic methods. By facilitating the investigation of high-Z material mixing into fusion burn, this system provides essential data for quantifying and interpreting results in high-energy density (HED) experiments. This report details the physics motivation, diagnostic fundamentals, planned and enacted upgrade work, and the quantification of uncertainty for the upgraded system.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Gear Test Assembly – Experimental Testing and Analysis of Gears and Bearings – FY2024

The Gear Test Assembly (GTA) is an experimental test apparatus built and installed in the Mechanisms Engineering Test Loop (METL) at Argonne National Laboratory (ANL). The focus of GTA has been to test larger radial spur gear sets machined from Inconel 718 along with a variety of bearing components that are to be used in the fuel handling machines built for use in pool-type sodium fast reactors. The testing in GTA also informs the design and material choices for the full-scale Gripper Test Assembly (GrTA) being developed by Argonne.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

High-Temperature Sodium System Structural Health Monitoring (SHM) Technologies at the Mechanisms Engineering Test Loop (METL)

This report documents the deployment of non-destructive testing (NDT) technologies at the Mechanisms Engineering Test Loop (METL). METL became operational in September 2018 with the mission to provide an ecosystem for Advanced Reactor Development (ARD). METLs flagship facility’s primary purpose is conducting small to intermediate scale tests for Sodium Fast Reactors (SFR). Its resemblance to a commercial SFR’s intermediate heat transport system, prototypic operating conditions, and industrial construction practices/materials provides the overarching benefit of establishing a proving ground for devices of all technology readiness levels as well as commercial off-the-shelf equipment yet to be demonstrated on a sodium system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Effects of Electron Irradiation on Candidate Materials for Target Window in Accelerator-Driven Mo-99 Production

Room-temperature (RT) axial tensile tests were conducted in 2018 and 2021 with flat, sub-sized, dog-boned samples of non-irradiated and irradiated Inconel 718 and irradiated beryllium (Be). The sample gauge length was 7.62 mm. Tensile properties of interest are the engineering yield stress (YS), ultimate tensile stress (UTS), uniform elongation (UE) and total elongation (TE). The electron dose of the 2021 samples was ≈2 times the dose of samples tested in 2018. Relative dose rates of 0, 1 (2018), and 2 (2021) are assigned to these samples. The 2021 electron-irradiated Be samples were brittle with a low failure stress of 338±64 MPa based on two samples. The failures occurred in the gauge section. The 2018 results for electron-irradiated Be indicated a slight decrease in ductility and strength for the one electron-irradiated sample that failed in the gauge section.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Lessons Learned from Open-Source Software Training Toward Expanding the Fusion Workforce

Within the United States (U.S.), the Fusion Innovation Research Engine (FIRE) Collaboratives seek to accelerate fusion technology development through wide-ranging community-driven research activities that bring together industry, laboratories, research institutes, and academia. Increasing the maturity of key components and systems for future fusion power plants (FPPs) toward commercialization, will, by necessity, increase the need for knowledgeable engineers, designers, and researchers in industry capable of integrating and further improving these technologies within industry FPP concepts. Further, various modeling and simulation packages support these programs and are under-development within them to drive design iteration and the development of FPP digital twins. To derive the greatest benefit from model output and capabilities, training these same specialists will be vital. Thus, the development of effective and accessible software onboarding and professional development opportunities focused on fusion will be key to keeping the pace of growth high in the coming decade and beyond. A similar need exists within the advanced fission reactor community, driven by an ever-growing need for carbon-free baseload power for industrial and data center applications. Within the U.S. and around the world, several open-source packages and frameworks exist to support this endeavor, and two we will highlight here are the Multiphysics Object Oriented Simulation Environment (MOOSE) framework and OpenMC. These packages, notably, are also being utilized in the FIRE Collaboratives program. In this presentation, we will highlight the lessons learned from over 30 years of combined software development and training experience, focused on developing strong technical software foundations within the nuclear workforce, from students to professional engineers & scientists. We will connect this experience to present and emerging needs in the fusion energy community, and, finally, will outline possible paths forward to develop a large, robust, global fusion workforce.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Scintillation and cherenkov photon counting detectors with analog silicon photomultipliers for TOF-PET

Objective. Standard signal processing approaches for scintillation detectors in positron emission tomography (PET) derive accurate estimates for 511 keV photon time of interaction and energy imparted to the detection media from aggregate characteristics of electronic pulse shapes. The ultimate realization of a scintillation detector for PET is one that provides a unique timestamp and position for each detected scintillation photon. Detectors with these capabilities enable advanced concepts for three-dimensional (3D) position and time of interaction estimation with methods that exploit the spatiotemporal arrival time kinetics of individual scintillation photons. Approach. In this work, we show that taking into consideration the temporal photon emission density of a scintillator, the channel density of an analog silicon photomultiplier (SiPM) array, and employing fast electronic readout with digital signal processing, a detector that counts and timestamps scintillation photons can be realized. To demonstrate this approach, a prototype detector was constructed, comprising multichannel electronic readout for a bismuth germanate (BGO) scintillator coupled to an SiPM array. Main Results. In proof-of-concept measurements with this detector, we were able to count and provide unique timestamps for 66% of all optical photons, where the remaining 34% (two-or-more-photon pulses) are also independently counted, but each photon bunch shares a common timestamp. We show this detector concept can implement 3D positioning of 511 keV photon interactions and thereby enable corrections for time of interaction estimators. The detector achieved 17.6% energy resolution at 511 keV and 237 ± 10 ps full-width-at-half-maximum coincidence time resolution (CTR) (fast spectral component) versus a reference detector. We outline the methodology, readout, and approach for achieving this detector capability in first-ever, proof-of-concept measurements for scintillation photon counting detector with analog silicon photomultipliers. Significance. The presented detector concept is a promising design for large area, high sensitivity TOF-PET detector modules that can implement advanced event positioning and time of interaction estimators, which could push state-of-the-art performance.

62 RADIOLOGY AND NUCLEAR MEDICINE↗