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Autonomous System Inference, Trojan, and Adversarial Reprogramming Attack and Defense (Final)

In the world of ever-advancing technology, Autonomous Systems (AS) find extensive application, bolstering functionalities of critical infrastructures such as nuclear power plants. These systems, however, are increasingly becoming a target for nefarious activities, namely through inference attacks, trojan attacks, and adversarial reprogramming. This paper delves into a comprehensive exploration of machine learning (ML)-driven autonomous control systems within advanced nuclear reactor designs, revealing the vulnerabilities and proposing strategies for defense against potential cyber-attacks. Advanced cyber-attacks against critical infrastructure and the energy sector are becoming more common. With the invention of autonomous control systems (ACS) within advanced nuclear reactor designs, system designers, reactor operators, and regulators must consider cybersecurity during the design and operational phases. This article provides a cyber threat assessment of machine learning (ML)- based digital twinning (DT) technologies in the context of advanced reactor ACS. A cyber-physical testbed was created to emulate nuclear reactor digital instrumentation and controls (I&C) and act as a basis for the ACS. The ACS was designed as two plant-level DTs predicting reactor malfunctions and determining control actions and two component-level DTs responsible for classifying component states and forecasting component inputs and outputs (I/O). Two duplicate ACS designs– one using a traditional ML framework and one using an automated ML (AutoML) framework– were created and tested against cyber-attacks on training data, real-time process data, and ML model architectures to determine their respective qualitative cyber-risk in terms of likelihood and impact. Both frameworks showed similar cyber-resilience against training, real-time, and ML architecture attacks, proving that neither is inherently more secure. Recommended safeguard and security measures are posed to system designers, reactor operators, and regulators to maintain the cybersecurity of ML-based DT technologies such as ACS, prompting a holistic view of shared responsibility for maintaining cyber-secure ML-based systems. As global reliance on generation III reactors begins to be critically assessed, the evolution towards advanced reactor systems utilizing digital instrumentation and controls (I&C) becomes not merely preferable, but essential. The integration of semi and fully autonomous control systems (ACS), powered by digital I&C and machine learning (ML)-based digital twinning (DT) technologies, emerges as a potent strategy to mitigate operations and maintenance costs, thereby enhancing the economic feasibility of novel reactor designs. However, with a staggering 500% and 380% increase in cyber-attacks reported against the energy sector by the United States Department of Energy (DoE) and the European Union respectively, a surge in cyber vulnerabilities specifically targeting the nuclear industry has been 2 markedly observed. Notable incidents, such as the W32.Ramnit spyware infiltration at the Gundremmingen nuclear power plant in Germany and the Dtrack spyware intrusion at the Kudankulam nuclear power plant in India, while not directly compromising core industrial control systems (ICS), underscore a compelling necessity to fortify cybersecurity protocols in safeguarding reactor systems against increasingly adept digital adversaries. In light of this, our investigation extends beyond conventional cybersecurity parameters, diving into the intricate web of potential vulnerabilities woven into ML-based DTs and ACS in advanced reactor systems. A crafted cyber-physical testbed and preliminary ACS were devised to act as a mirror, reflecting potential configurations of advanced reactor control designs. Moreover, this study is intertwined with a scrutinization of ML models, developed either through conventional, manually tuned methodologies or via automated means through AutoML, probing into their cyber-risk profiles within operational technology (OT) environments. Expanding on this, two distinct ACS blueprints were forged – one navigating through the corridors of traditional ML and the other traversing the path of AutoML – in an effort to holistically encapsulate the considerations pivotal to ML-based DT control system design. Employing the SANS Institute Industrial Control System (ICS) Kill Chain and the MITRE ATT&CK Tactics, Techniques, and Procedures (TTP) framework, a structured analysis was conducted, launching three targeted attacks against the training dataset, real-time dataset, and ML models, therein dissecting the potential cyber-attack implications against both ML frameworks within an ACS milieu. It is essential to note that three distinct categories of attacks were conducted against both ACS configurations, each encompassing three distinct ML-based DTs, cumulating in a total of 18 varied attacks. This exploration extends into the realms of Autonomous System Inference, Trojan, and Adversarial Reprogramming Attack and Defense, unraveling vulnerabilities, and opportunities for fortified defenses against such intrusions, particularly where ML-driven technologies, and by extension, ACS, are deployed. Final recommendations, articulated through a lens of security, safeguard, and implementation considerations, are presented for both traditional and AutoML models, anchoring upon the existing knowledge landscape and ML-based DT modeling for ACS, and are offered as a beacon to guide the nuclear industry through the intricate cybersecurity challenges that lie ahead.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Ceramic encapsulated metal phase change material for high temperature thermal energy storage

Thermal energy storage (TES) is a broad-based technology for reducing CO₂ emissions and advancing concentrating solar, fossil, and nuclear power through improvements in efficiency and economics. Phase change materials (PCMs) are of interest as TES media because of their ability to store large amounts of heat in relatively small volumes. The volume expansion during a phase change, typically between a solid and liquid, can cause breakage of protective coatings. This paper reports on the fabrication of a ceramic encapsulated metal (CEM) high temperature TES technology using a rotary calcining furnace and a fluidized bed chemical vapor deposition coating technique. Aluminum beads were chosen as the PCM because Al has a high melting point (660 °C), low cost, high heat of fusion, and an ability to form a thin, strong alumina layer capable of supporting the Al melt for subsequent processing. Quite remarkably, this study shows that 1 mm diameter Al can be fluidized up to at least 1500 °C in an appropriate atmosphere while maintaining a spheroid geometry. This allowed for producing a first of a kind CEM whereby Al particles were encapsulated in pyro-carbon (PyC) and high purity, dense chemical vapor deposited SiC. The CEM with a PyC only coating was exposed to thermal cycling to test the performance with a differential scanning calorimeter; the melting point and latent heat were measured to be 648.4 ± 2.8 °C and 293.3 ± 6.2 J/g respectively. It was demonstrated that the CEM design is possible to produce, laying the foundation for manufacturing of high temperature, tunable, TES media.

25 ENERGY STORAGE↗

Preliminary Feasibility of Printing, Microstructure Analysis and Mechanical Performance of a Down Selected Ni Alloy

This report is submitted as completion of a Milestone 3 deliverable under work package ORNL CT-23OR1304051 in support of the Advanced Materials and Manufacturing Technologies (AMMT) program. The AMMT program is aiming at the faster incorporation of new materials and manufacturing technologies into complex nuclear-related systems. An integrated approach, combining advanced characterization, high-throughput and accelerated testing, modeling and simulation including machine learning and artificial intelligence will be employed. While 316H (Fe-16-18Cr-10-14Ni-2-3Mo-0.04-0.1C) has been identified as a key alloy to be integrated into the AMMT accelerated alloy qualification approach due its relevance for many current and future nuclear energy reactors, many other alloys could be considered for the advanced fabrication of innovate high-performance nuclear components. ANL, INL, ORNL and PNNL are collaborating on identifying the most promising alloy candidates relevant for the AMMT program. A selection criteria matrix was established to evaluate the alloys considering their relative importance and technological readiness levels for nuclear energy applications, with a focus on laser powder bed fusion (LPBF). Due to the broad range of potential candidate alloys, ORNL and INL focused on Ni-based alloys, while ANL and PNNL mainly evaluated Fe-based alloys. PNNL previously published materials scorecards reports on several key alloys and this report is providing a broader overview of Ni-based candidate alloys, expending beyond alloys well-known to the nuclear community. Of particular interest are alloys that are currently commercially available in powder form due to the growing demand from other industries that have invested heavily in additive manufacturing. Among these alloys, Haynes 282 (Fe-20Cr-20Co-8Mo) was selected due to its superior strength at high temperature compared to the code-qualified alloy 617 (Ni-20-24Cr-10-15Co-8-10Mo). To assess the integration of this new alloy into the AMMT digital manufacturing framework, we performed the rapid optimization of alloy 282 printing parameters on a Renishaw 250 machine, the fabrication of sufficient materials for extensive characterization and mechanical testing both at ORNL and INL, and added the printing data into the digital platform via the Peregrine software. A detailed analysis of the LPBF 718 alloy was also conducted, with creep specimens being tested at 600-650°C and characterized by advanced electron microscopy. The alloy superior mechanical strength and great printability associated with the extensive database that has already been generated highlight the promising potential of LPBF 718 as a candidate alloy for the AMMT program. INL, ANL and PNNL have generated similar reports and all the information will be compiled into a final M2 milestone report to provide the AMMT leadership team with clear recommendations on the down selection of reactor materials, as well as establish a roadmap for the qualification of these selected alloys.

36 MATERIALS SCIENCE↗

NASA Utilization of Space Nuclear Systems for Robotic and Human Exploration Missions: Response to EO 13972: Promoting Small Nuclear Reactors for National Defense and Space Exploration

Space Nuclear Systems (SNS) technology development offers a wide range of capabilities to support NASA’s current and future missions. Executive Order (EO) 13972, “Promoting Small Modular Reactors for National Defense and Space Exploration” [1], issued 5 January 2021, directs NASA to define requirements for NASA utilization of nuclear energy systems for human and robotic exploration missions through 2040 and analyze the costs and benefits of such requirements.” Although it is premature to define requirements and cost for future exploration missions that have not yet been formulated, this report describes planned objectives and missions by 2040 that are enabled or enhanced by nuclear systems while taking into account a number of unique considerations for nuclear energy in the space environment. Nuclear energy systems are enabling for space missions and critical capabilities where conventional forms of energy production are impractical or impossible due to mass constraints, mission duration, or distance from the Sun. Space nuclear technologies available or in development for use by 2040 utilize radioisotope decay or nuclear fission and fall into three categories: heat, power, and propulsion. Current applications utilize radioisotope power systems that provide consistent and reliable performance in the sub-kilowatt power range. More advanced SNS can enable new mission objectives where high energy density solutions are critical, or where access to solar solutions is prohibitive. Higher power radioisotope and fission systems are under development within NASA for a wide variety of human exploration and science mission applications. Planned missions designed to use radioisotope systems include Dragonfly, a rotorcraft that will explore the surface of Titan, and Persephone, a mission concept for a Pluto orbiter. Nuclear fission systems have the key advantage of providing significantly higher power, lower mass solutions from tens to even thousands of kilowatts. Fission power is enabling to a sustained human presence on the Moon and developing a robust lunar economy. Fission propulsion is enabling for missions within and beyond cis-lunar space. This report examines NASA-envisioned mission applications and associated performance needs for SNS over the next twenty years leading to 2040 along with the unique technical considerations posed by space nuclear technology development. This includes engineering and operational logistics for ground handling, thermal management, survival of the space environment, operational safety, power requirements, and service longevity. Safety to the public, the NASA work force, and agency assets remains a top priority for NASA and particular attention is given to this aspect in the design, hardware assembly, ground operation, launch, and mission operation of an SNS. NASA relies on the Department of Energy as nuclear authority and its legacy of rigorous safety procedures as standards for ground development, test, transportation, and launch site operation. The principal concern is preventing unintended radiological release to the public or environment. Radioisotope system experience has established processes, including ground operation, transportation, and launch, that are considered directly applicable to emerging fission systems; however, fission systems have unique design needs that impact the safety and performance requirements. High efficiency power conversion from both fission and radioisotope systems requires high operating temperatures necessitating both passive and active thermal management to maintain safe and nominal operating conditions. Effective cooling and waste heat rejection have special considerations for space applications, whether in zero-g or reduced gravity. Fluid and heat transfer within the reactor system is not anticipated to be impacted by reduced or zero-g environments. Cryogenic working fluids and propellant supplies utilized in some space nuclear applications will need low mass, high capacity cryocoolers to meet the long-term storage and near zero-boiloff needs. Integrated, high power density SNS capable of being packaged in a single vehicle is a key consideration for NASA. Due to concerns for complexity and reliability, in space reactor assembly and reactor refueling are not current design considerations. Expanding into a new era for space exploration depends on mass-efficient, high-energy solutions to power deep-space vehicles, operate in harsh environments, and increase mission flexibility. NASA nuclear technology investments are targeting power for surface operations and propulsion for fast-transit, deep-space missions, all with the ability to reliably operate without the need for repair or refueling. NASA’s goals, enabled by nuclear technologies, provide for exciting advances in scientific objectives and human exploration, ushering in a new space age that enables a human presence on bodies beyond our Earth.

nuclear↗

Advanced Fabrication Techniques of Metal Hydrides for Science and Technology Applications (Abbreviated Final Report)

Lithium Hydride (LiH) atomic properties make it an excellent candidate for thermal energy storage, hydrogen storage, and nuclear reactor technology. High Energy Density (HED) experiments on LiH at the National Ignition Facility (NIF) can also provide critical Equation of State (EOS) and Hugoniot data. Density requirements for LiH vary per application and therefore physical and structural requirements for LiH are non-uniform. Historical production of LiH relies on casting processesthat are limited by density and grain-size control, which leads to unfavorable machinable characteristics. LiH manufacturing is further hindered due to its chemical reactivity and propensity to readily react with moisture. This work aimed at exploring advanced fabrication techniques for LiH such as 1) uniaxial powder pressing methods and 2) laser powder bed fusion (L-PBF) additive manufacturing. Uniaxial powder compaction offers advantageous tailorable mechanical and physical properties via density control, and L-PBF can produce net-shaped complex parts with unique microstructure. Uniaxial pressing proved successful in readily producing LiH with density control.

36 MATERIALS SCIENCE↗

Advanced Fabrication Techniques of Metal Hydrides for Science and Technology Applications (Full Technical Report)

Lithium Hydride (LiH) atomic properties make it an excellent candidate for thermal energy storage, hydrogen storage, and nuclear reactor technology. High Energy Density (HED) experiments on LiH at the National Ignition Facility (NIF) can also provide critical Equation of State (EOS) and Hugoniot data. Density requirements for LiH vary per application and therefore physical and structural requirements for LiH are non-uniform. Historical production of LiH relies on casting processes that are limited by density and grain-size control, which leads to unfavorable machinable characteristics. LiH manufacturing is further hindered due to its chemical reactivity and propensity to readily react with moisture. This work aimed at exploring advanced fabrication techniques for LiH such as 1) uniaxial powder pressing methods and 2) laser powder bed fusion (L-PBF) additive manufacturing. Uniaxial powder compaction offers advantageous tailorable mechanical and physical properties via density control, and L-PBF can produce net-shaped complex parts with unique microstructure. Uniaxial pressing proved successful in readily producing LiH with density control.

36 MATERIALS SCIENCE↗

Radiation Effects in Next Generation Used Nuclear Fuel Reprocessing Strategies

With the global community committed to significantly expanding nuclear energy capacity, the development of efficient used nuclear fuel (UNF) management strategies has become more critical than ever. These strategies are vital to fostering the widespread adoption of closed fuel cycles, which are essential for sustainable nuclear energy production and security. Achieving this ambitious goal necessitates a comprehensive understanding of radiation effects on next-generation technologies, as radiolysis can often limit the longevity and performance of these systems. This seminar will provide an overview of next-generation UNF reprocessing strategies, highlighting the latest advancements and innovative approaches in the field. Particular attention will be given to two key areas of recent research: 1. Radiation robustness and performance of advanced sulfur chloride-based chlorination technologies. We will explore the efficacy of sulfur chloride-based chlorination processes in the presence of surrogate cladding materials, specifically aluminum. These processes have shown promise in the dissolution, decontamination, and recovery of cladding materials for reuse. Detailed findings on how the composition and performance of these sulfur chloride solvents respond to radiation exposure will be discussed. 2. Impacts of metal ion complexation and direct dissolution conditions on monoamide-based reprocessing strategies. We will delve into the time-resolved and dose accumulation effects of irradiation on the direct dissolution of voloxidized uranium and rhenium using N,N-di-(2-ethylhexyl) butyramide (DEHBA) or N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) in pre-equilibrated n-dodecane solvent. The implications of these interactions on dissolution efficiency, radiolytic stability, and overall process performance will be examined. These studies aim to underscore the importance of understanding radiation effects in the development of next-generation UNF reprocessing technologies and the global transition towards more sustainable and efficient nuclear energy systems.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗

Advances in GDSA Framework Development and Process Model Integration

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and highlevel nuclear waste (HLW). A high priority for SFWST disposal R&D is to develop a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media. This report describes fiscal year (FY) 2020 advances of the Geologic Disposal Safety Assessment (GDSA) Framework and PFLOTRAN development groups of the SFWST Campaign. The common mission of these groups is to develop a geologic disposal system modeling capability for nuclear waste that can be used to probabilistically assess the performance of disposal options and generic sites. The capability is a framework called GDSA Framework that employs high-performance computing (HPC) capable codes PFLOTRAN and Dakota.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Structures, Systems, and Components Classification Criteria for the Versatile Test Reactor

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy (DOE), Office of Nuclear Energy. The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by the U.S. National Laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc. As part of the VTR authorization process by the DOE, it is necessary to identify the safety classification of structures, systems, and components (SSCs) to determine the proper design, surveillance, operation, and incident-reporting requirements. Since VTR is utilizing a risk-informed, performance-based authorization process based on the Licensing Modernization Project (LMP) described in NEI 18-04 [1], the criteria for SSC classification, which are reviewed here, includes both risk information from the probabilistic risk assessment (PRA) and prescriptive requirements.

Grabaskas, David↗

The Elevated-Temperature Cyclic Properties of Alloy 316H Fabricated by Powder Metallurgy Hot Isostatic Pressing (Rev.1)

Powder metallurgy hot isostatic pressing (PM HIP), an advanced manufacturing technique, is viewed by the Department of Energy as a “transformational technology” for heavy section manufacturing of nuclear components. Although PM HIP is a mature technology, it has not been qualified in Section III, Division 5 for elevated-temperature nuclear construction. This work seeks to demonstrate a PM HIP qualification method for elevated-temperature nuclear construction that is based on allowable material characteristics. Specifically, the elevated-temperature cyclic properties of PM HIP 316H are investigated. The creep-fatigue properties of PM HIP 316H are significantly degraded compared to those of wrought. Current PM HIP practices are not able to fabricate elevated-temperature nuclear components with creep-fatigue properties that are equivalent to or better than traditional wrought components. A research and development program is proposed for qualifying PM HIP 316H for elevated-temperature nuclear construction.

36 MATERIALS SCIENCE↗

Evaluation of In-Situ AM Process Monitoring Techniques and Potential for Detecting Process Anomalies and Undesirable Microstructures

The US Department of Energy’s Advanced Materials and Manufacturing Technologies (AMMT) program is pursuing rapid qualification of new materials for fabrication of nuclear relevant components using advanced manufacturing techniques. Particular interest is placed on code-qualifying stainless steel (SS) 316H processed by laser powder bed fusion (LPBF) additive manufacturing. A paradigm that incorporates data from in-situ sensing during the printing, ex-situ characterization, and advanced artificial intelligence–based models was established under the Transformation Challenge Reactor (TCR) program to develop a pedigree for each fabricated component that could be tracked from the feedstock to the component’s release for application. Under the TCR program, the Peregrine software was developed as a tool for incorporating the vast amounts of in-situ and ex-situ characterization data collected; all data stored on a rapidly growing digital platform. The digital platform allows for users to link site-specific process anomalies to the macro- and microstructure. The platform will eventually be able to predict component performance, which will be crucial to qualifying materials and components in risk-averse industries such as those supporting and building nuclear reactors. Current in-situ process monitoring techniques that are already integrated with software like Peregrine are advantageous for identifying process anomalies including powder spatter, component edge swelling, recoating-build interactions, and so on. However, additional data are required to fully predict the resulting microstructures needed for identifying relationships to component performance. The rapid cooling rates observed in LPBF are some of the highest of any bulk manufacturing process, resulting in heterogenous microstructures and typically causing anisotropy in mechanical properties. Moreover, evolved residual thermal stresses are high, which can cause severe defects such as delamination or cracking. Therefore, other in-situ monitoring methods are warranted for exploration to measure and map the thermal history, and potentially the stress state, of each build. This report summarizes different in-situ monitoring strategies proposed for LPBF with a focus on the more developed sensor systems. Novel capabilities for measuring melt pool temperatures are also addressed to better inform modeling efforts.

36 MATERIALS SCIENCE↗

Regulatory Research Planning for Micro-Reactor Development

A limited-scope Regulatory Research Development Plan (RRDP) is being developed to link important advanced non light water nuclear microreactor technology research and development (R&D) activities sponsored by the U.S. Department of Energy’s Office of Nuclear Energy (DOE NE), Science, and Technology to regulatory requirements and critical licensing needs likely to impact the deployment of a domestic commercial microreactor fleet. This document is a precursor to that forthcoming limited scope microreactor RRDP. This report is being published to aid in the identification of key fiscal year (FY) 2022 topical research opportunities (as conveyed to the RRDP development team) by respondent microreactor developers and suppliers in consideration of micro-reactor regulatory needs and challenges noted by industry, DOE, and Nuclear Regulatory Commission (NRC) stakeholders. Although this document does not contain the regulatory effects analysis and R&D recommendations that is expected in the final RRDP, this document is being released now to assist DOE researchers and investigators in FY22 research planning activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Enabling a Physical Twin for Control Methods Evaluation

Advanced nuclear reactors play an important role in the energy future of the United States and the rest of the world. They are designed and operated based on a different model than that of the current operating fleet, thus enabling deployment in remote locations and allowing for safe semi-autonomous or autonomous operations. Such characteristics require the development of a new reactor control paradigm. A significant factor in the development of control technologies and methods is integration of the various technologies and methods with each other and with hardware (both reactor system hardware and control hardware). A recent workshop on control of advanced reactors identified the lack of a flexible, expandable software/hardware infrastructure to enable such integration as a key gap. A previous phase of the current effort involved developing and demonstrating the Control and Optimization Modular Modeling Application for Nuclear Deployment (COMMAND) platform, which is capable of integrating autonomous-control-enabling technologies and methods, without the constraints imposed by existing software solutions. This platform will enable advanced reactor developers to deploy and test any developed technologies and methods by employing a common framework, and to couple them with their own models and hardware. The present phase of this effort entails using the Microreactor Automated Control System (MACS) platform, which was developed by the U.S. Department of Energy (DOE) Microreactor Program, to serve as a control method testbed. MACS can be used by advanced reactor developers to integrate their control related research activities with any reactor system. For the present effort, MACS was customized to mirror Idaho National Laboratory (INL)’s Microreactor Applications Research Validation and Evaluation (MARVEL) microreactor, and COMMAND was leveraged to enable MACS to emulate the physics of MARVEL, thus positioning MACS as a physical twin of MARVEL.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

2020 Budget Request for the DOE Computational Science Graduate Fellowship (CSGF) Grant

The Department of Energy Computational Science Graduate Fellowship (DOE CSGF) is essential for addressing the increasingly complex national workforce demands stemming from the growth of computational science and engineering challenges. Computational science and engineering (CSE) takes a multidisciplinary approach that utilizes scientific computing to tackle practical problems and provide technical tools across the spectrum of scientific discovery. The DOE CSGF specifically highlights high-performance computing (HPC) as a critical enabling technology in CSE, driving advancements in science and engineering that are vital to both the DOE and the broader economy. Over the past half-century, HPC has been an essential tool for DOE’s success. During this period, important missions, such as nuclear stockpile stewardship, have turned to HPC as an essential technology. Entire science disciplines have been transformed through the augmentation of scientific observation via HPC. At government laboratories, academic institutions, and in industry, DOE CSGF alumni are helping push traditional HPC boundaries while contributing to discoveries in high-energy physics, quantum information systems, fusion-reactor design, machine learning, additive manufacturing, nano materials for next-generation batteries and transistors, and advanced nuclear reactor modeling. In addition, HPC is used to address national health needs that will eventually point to cures both by helping cancer researchers manage and analyze huge troves of data, by simulating biological mechanisms, and by accelerating drug development. A 2023 report from the ASCAC Subcommittee on American Competitiveness and Innovation to the ASCR office, “Can the United States Maintain Its Leadership in High-Performance Computing?” says of the Program, “The CSGF program provides a barometer for disciplines that will be of interest to future DOE computing. Computational biology, machine learning, and quantum computing are among the subjects that began to swell in the ranks of CSGF applicants before the labs were hiring as high a percentage of employees in these categories.” The explosion of scientific and technological data has heightened the demand for advanced high-performance computing (HPC) to transform these data into meaningful scientific insights. As access to vast amounts of data increases, the fields of Machine Learning and Artificial Intelligence are experiencing a resurgence, enhancing the established practices of computational modeling and simulation. In its September 2020 subcommittee report on "AI/ML, Data Intensive Science, and High-Performance Computing," the DOE Advanced Scientific Computing Advisory Committee (ASCAC) specifically called for a fellowship program to train computational and data scientists to address exascale and data-intensive computing challenges. This integration of empirical and theoretical modeling will increasingly guide federal policymakers in making decisions that impact American society and future generations. It demands a workforce of highly skilled and intellectually agile computational scientists capable of navigating the rapid advancements in scientific computing within the DOE National Laboratory research environment. The DOE CSGF program has consistently addressed this critical need.

97 MATHEMATICS AND COMPUTING↗

Sockeye Heat Pipe Analysis Code Verification and Validation

Some of the most promising microreactor designs currently under development utilize heat pipe technology to transfer heat from the reactor core to the secondary side heat exchanger, due to the technology’s compactness, efficiency, passivity, and reliability. Sockeye is an engineering-scale heat pipe tool developed under the Nuclear Energy Advanced Modeling and Simulation Program to be used for the design and safety analysis of microreactors. Sockeye’s core capability lies in a 1D, two-phase, compressible flow model, used to simulate the working fluid inside the heat pipe. Sockeye is built on the Multiphysics Object-Oriented Simulation Environment framework, which allows for seamless multiphysics coupling with other Nuclear Energy Advanced Modeling and Simulation tools and can thus be used in a full-scale simulation of a microreactor assembly, which can include hundreds of heat pipes. This paper presents some initial verification and validation assessments performed for Sockeye. We demonstrate good agreement between Sockeye’s numerical results and steady-state analytic solutions for velocity and pressure drop, with differences attributable to the underlying assumptions made by the analytic solutions. We also show that Sockeye reproduces analytic predictions of key operational limits, such as the capillary limit and the sonic limit. Finally, we compare Sockeye results to experimental data for the SAFE-30 heat pipe module test.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Simulations of the High Temperature Test Facility using SAM

Under the support from the U.S. Department of Energy Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, an effort is being pursued to support the modeling and simulation needs of high-temperature gas-cooled reactor (HTGR) technology development. There is a particular need for advanced modeling and simulation methods and tools to predict thermal-fluid behavior in the nuclear reactor primary system during safety-related transients. This report focuses on one such activity related to HTGR: developing a model of the High Temperature Test Facility (HTTF) at Oregon State University using the system-level code SAM, and using the model to understand thermal response behavior in the facility. Note this activity is coordinated with the DOE-NE’s Advanced Reactor Technology Gas-Cooled Reactor Program

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Design and Prototyping of Advanced Control Systems for Advanced Reactors Operating in the Future Electric Grid (Final Report)

Despite its significant advantages as a baseload, low-carbon energy source, the U.S. nuclear power industry has faced increasing difficulties in maintaining economic competitiveness in a rapidly evolving energy market. The economic conditions faced by the current fleet of nuclear power plants (NPPs) in the U.S. deregulated electricity market require a concerted effort to mitigate specific cost factors. Many units are struggling to stay competitive, and some premature shutdowns have occurred. Besides, in response to the large penetration of renewable energy sources, the role of nuclear power plants as pure baseload units needs to be reconsidered. Based on these experiences, operational flexibility is considered a fundamental requirement for the next generation of nuclear reactors to be competitive in the future energy market. The deployment of advanced reactors capable of operating within a new power grid paradigm, known as the Integrated Energy System (IES) was investigated. This approach combines new reactor designs with Thermal Energy Storage (TES) technologies, allowing the nuclear reactor to maintain a steady power output without the need for constant adjustments in response to load demand fluctuations. With this configuration, the reactor operates as a baseload unit, experiencing only very gradual power transients, while the energy storage facility within the power conversion cycle acts as a peaking unit.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Brayton Power Conversion System Study to Advance Technology Readiness for Nuclear Electric Propulsion

Recently, there has been significant interest within the aerospace community to develop space based nuclear power conversion technologies especially for exploring the outer planets of our solar system where the solar energy density is very low. To investigate these technologies NASA awarded several contracts under Project Prometheus, the Nuclear Systems Program. The studies described in this paper were performed under one of those contracts, which was to investigate the use of a nuclear power conversion system based on the closed Brayton cycle (CBC).The investigation performed included BPCS (Brayton Power Conversion System) trade studies to minimize system weight and radiator area and advance the state of the art of BPCS technology. The primary requirements for studies were a power level of 100 kWe (to the PPU), a low overall power system mass and a lifetime of 15 years (10 years full power). For the radiation environment, the system was to be capable of operation in the generic space environment and withstand the extreme environments surrounding Jupiter. The studies defined a BPCS design traceable to NEP (Nuclear Electric Propulsion) requirements and suitable for future missions with a sound technology plan for technology readiness level (TRL) advancement identified. The studies assumed a turbine inlet temperature approx. 100 C above the current the state of the art capabilities with materials issues and related development tasks identified. Analyses and evaluations of six different HRS (heat rejection system) designs and three primary power management and distribution (PMAD) configurations will be discussed in the paper.

Patrick E. Frye↗