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DOE OSTI2025

A model to assess Zircaloy’s mechanical property changes following a transient beyond critical heat flux

Maintaining the integrity of nuclear fuel rods is essential for ensuring public health and safety in nuclear power generation. During reactor operation, this integrity is confirmed by demonstrating compliance with established regulatory acceptance criteria. For moderate-frequency events, such as limiting transients and anticipated operational occurrences (AOOs), the current fuel integrity criterion is based on preventing boiling transition. This criterion assumes that prevention of boiling transition will prevent excessive cladding heating and, thus, fuel failure during normal operations. While conservative, this approach places significant constraints on core design, fuel cycle economics, and a plant’s ability to perform major power uprates, leading to suboptimal fuel utilization and inefficient carbon-free energy production. A more efficient approach could be achieved by revising the failure criterion to a material-specific limit rather than strictly preventing the boiling transition, since boiling transition per se is not a cause of fuel cladding failure. Here, as a result, a new licensing framework based on material properties, termed time-at-temperature (t@T), is needed. This approach would allow for brief periods of post–critical heat flux operation during an AOO without compromising safety. Implementing the t@T licensing strategy requires a robust technical foundation in material properties, which must be established through comprehensive data collection on both unirradiated and irradiated fuel and cladding materials. This foundation would enable the development of a safety basis that ensures safe operation while providing greater flexibility and efficiency for reactor operation. This paper documents a thorough review of the available data to establish a baseline knowledge that can inform the development of cladding mechanical models, as well as identify experimental data gaps that need to be addressed in future research. Machine learning and data informatics were utilized to extract the importance of parameters on the t@T parameter. Industry tools were used to perform baseline analyses to define the relevant transient conditions for data analysis. The subsequent review successfully identified applicable experimental data, as well as sufficient data to evaluate changes in cladding mechanical properties following an AOO transient. Rather than developing new models, this work coupled existing irradiation annealing and recrystallization models to calculate changes in hardness, yield stress, and ultimate tensile stress following an AOO event. The findings from this review were summarized to highlight the experimental data needs required to fill remaining gaps and support the development of future t@T licensing methodologies.

Cladding performance↗

DOE OSTI2025

Autonomous Operations for Advanced Reactors Utilizing Supervisory Control

Automation is a critical tenet of reactor plant operations as reliance on nuclear energy increases. Nuclear power plants require a large workforce which does not scale with output; that is, the cost per megawatt increases as reactor output becomes smaller. The economic viability of advanced reactors, particularly small modular reactors (SMRs) and microreactors, requires a significantly reduced onsite workforce. The logical solution is establishing a systematic process of elimination of reliance on human operators, and to the extent possible, replacing these actions with automated functions. In this paper, we propose a method for such transformation to establish a robust technical basis to enable transition to autonomy. Our method is based on finite state automata (FSA)—also known as finite state machines (FSMs). Relying on this method allows us to exploit the rich set of mathematical proofs available in the field of regular languages. FSA are one of the mathematical tools to model discrete event systems (DES). These properties are applied to produce an automated startup controller for the Massachusetts Institute of Technology Research Reactor (MITR). The startup procedure is captured in terms of discrete changes from one state to another while an independent supervisory control system directs the sequence of states and alerts a human in the event of an abnormal operation. First, the design and behavior of the MITR rod control system were modeled in Simulink. Then, the startup procedure was applied to the rod control system and the DES performed a startup by procedurally withdrawing rods to the subcritical position. The simulation also stops rod motion in response to an uncontrollable event and restarts rod motion once the event has been cleared.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

DOE OSTI2020

Use of Multi-Attribute Utility Analyses to Assess Transportation Routes from ISFSIs - 20535

With two consolidated interim storage facilities for used nuclear fuel (UNF) having applied for storage licenses, there is the potential need to transport UNF in the U.S. from shutdown reactor sites to complete the site cleanup and from operating reactor sites who may be running out of storage options (e.g., spent fuel pools are nearly full and dry storage facilities are reaching licensed capacity). At some of the shutdown reactor sites, only the independent spent fuel storage installation (ISFSI) exists and often there remains a degraded transportation infrastructure for performing shipments from these sites. At other sites (e.g., operating reactor plants), the existing transportation infrastructure on the site is unchanged and sufficient for moving UNF, but the environment around the route may have changed (e.g., what was once a rural setting has transformed into a more urban setting) and/or local stakeholders want their input taken into consideration in UNF transportation activities. So, although preliminary evaluations have been performed to identify viable modes of transport from these sites using rail, barge, and/or heavy haul truck, if multiple options exist for moving the UNF from the site (e.g., multiple truck routes or a truck and a barge route both exist) then an evaluation that takes input from multiple stakeholders should be considered to identified the preference (e.g., ranking) of routes from a site. Orano has supported the development of several studies examining the viable routes for removing UNF from some of these sites and has utilized a multi-attribute utility analysis (MUA) to establish a ranking of routes and associated modes to move the UNF from the ISFSIs to a Class I rail carrier. The ultimate result from the MUA is a list(s) of the most to the least favored/preferred routes from the site. This paper provides an overview of the MUA methodology and provides an example of its application to a site with a shutdown reactor. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Brittle nature and the related effects of zirconium hydrides in Zircaloy-4

The use of zirconium alloys such as Zircaloy-4 in nuclear reactors is limited by brittle hydride formation, for example in harsh reactor operating conditions such as those in the pressurized water reactors. Even though much research has studied this phenomenon, a more systematic, in-depth study of the effects of hydrides on the properties of Zircaloy-4, using various analytical techniques, can add more details to the understanding of these effects. In the current study, hydrogen charging was performed on Zircaloy-4 at different levels compatible with commercial reactor operating conditions and conditions in accident scenarios. Effects of hydriding on the Zircaloy-4 (α-(hcp)-Zr) tensile properties, together with an in-depth study of any relationship between tensile and the microstructural properties of the material, are discussed. While a moderate pinning effect on the equiaxed α-Zr grain size was observed, no significant change in bulk texture of the Zry-4 was revealed as the hydrogen level increased up to nominally 1230 wppmH. At a nominal value of 490 wppmH, Zircaloy-4 fractured at a low strain value just above the yielding point, and brittle tensile fracture was observed in a sample H-charged to nominally 1230 wppmH. The low ductility and brittle fracture of the samples were attributed to the presence of brittle inter-grain boundary hydrides. Finally, it was observed that brittle fracturing was also supported by intra-grain boundary hydrides that were affected by the tensile deformation of the sample.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Interface Tracking Investigation of Geometric Effects on the Bubbly Flow in PWR Subchannels

We report that absorbing heat from the fuel rod surface, water as coolant can undergo subcooled boiling within a pressurized water reactor (PWR) fuel rod bundle. Because of the buoyancy effect, the vapor bubbles generated will then rise along and interact with the subchannel geometries. Reliable prediction of bubble behavior is of immense importance to ensure safe and stable reactor operation. However, given a complex engineering system like a nuclear reactor, it is very challenging (if not impossible) to conduct high-resolution measurements to study bubbly flows under reactor operation conditions. The lack of a fundamental two-phase-flow database is hindering the development of accurate two-phase-flow models required in more advanced reactor designs. In response to this challenge, first-principles–based numerical simulations are emerging as an attractive alternative to produce a complementary data source along with experiments. Leveraged by the unprecedented computing power offered by state-of-the-art supercomputers, direct numerical simulation (DNS), coupled with interface tracking methods, is becoming a practical tool to investigate some of the most challenging engineering flow problems. In the presented research, turbulent bubbly flow is simulated via DNS in single PWR subchannel geometries with auxiliary structures (e.g., supporting spacer grid and mixing vanes). The geometric effects these structures exert on the bubbly flow are studied with both a conventional time-averaging approach and a novel dynamic bubble tracking method. Lastly, the new insights obtained will help inform better two-phase models that can contribute to safer and more efficient nuclear reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Automatic Modeling of VVER-1000 Power Maneuvers

A special version of the IR code is described. The purpose of this code is to generate recommendations for the reactor operator when operating in a maneuvering regime, automate boron regulation, and study the maneuvering characteristics of VVER-1000 fuel loads. The results of modeling of operation in a daily load schedule are presented for the example of a stationary run of VVER-TOI. The impact of variation of the parameters of power maneuvers (run, operating time a reduced power, loading rate) on the boron regulation parameters is analyzed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

An Autonomous Critical Data Extrapolator for the AGN-201m

Nuclear nonproliferation serves as a key goal, being undertaken by the International Atomic Energy Agency (IAEA). To recognize proliferation there are two pathways that states, who intend to use nuclear material for malicious purposes can take, diversion can misuse. Diversion is when fissile nuclear material is declared to the IAEA for non-weapon purposes, but then covertly removed. If the source of nuclear material, that is not declared and not fissionable, is placed inside the reactor core to create fissile material used to create weapons then the state is using the second pathway of proliferation, misuse. With the emerging development in areas of simulation and machine learning the creation of virtual models of reactor systems, digital twins, serve as a potential method to identify proliferation through detecting anomalous behavior in the reactor. A digital twin for a physical nuclear reactor has never been developed, as digital twins serve as an emerging technology. To investigate the process for development and use of a digital twin for a nuclear reactor Idaho State University’s AGN-201m serves as the nuclear reactor used for development of this digital twin. A data acquisition system has been installed to the reactor system allowing for the transfer of collected data from a reactor operation to Idaho National Laboratory’s Deeplynx data warehouse. When utilizing data to train reactor physics and machine learning models, a significant challenge encountered is the initial state of the data. Nuclear proliferation will have the capacity to be detected when the reactor immediately starts up, nor will it occur after the reactor shuts down. Generally, it will be detected when the reactor is operating at some desired power over a sufficient period for that specific reactor design. For the AGN-201m this will be when the reactor is critical (generally 1 mW or above) for a timespan that is within or less than the range of a regular business day. Datasets sent to Deeplynx have had to be manually cut to when the reactor is critical based on plots of power levels. This method is inefficient and laborious, especially when using multiple datasets at once to train a model. To provide a more streamlined approach an automated critical data extrapolator is developed, with capabilities of recognizing when the reactor operation first reaches criticality, and when the reactor undergoes a SCRAM and is shutdown.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2024

Summary of LWRS Research in Addressing RPV Research Gaps in NRC EMDA Report

Reactor Pressure Vessels (RPVs) are critical components in nuclear reactors, housing the reactor core and coolant under extreme conditions of temperature, pressure, and radiation. These harsh environments contribute to the degradation of RPV materials over time, presenting challenges for extending reactor operations beyond their original design lifespans. The NRC Expanded Materials Degradation Assessment (EMDA) report volume 3 have been instrumental in guiding research to support extending the operational life of light water reactors (LWRs) up to 80 years. It provides a comprehensive framework to address technical challenges related to aging and degradation mechanisms in RPVs. The LWRS program has played a key role in advancing this research, supporting projects such as the UCSB ATR-2 Experiment, material testing from Zion and Palisades reactors, and the development of advanced mini-compact tension testing techniques. These efforts have been crucial in identifying and addressing gaps in our understanding of RPV aging, contributing to the successful subsequent license renewals of eight LWR units in the U.S. The EMDA report volume 3, built on the Phenomena Identification and Ranking Table (PIRT) analysis from earlier versions of the EPRI Materials Degradation Matrix (MDM) and Issue Management Tables (IMTs), provides a detailed assessment of RPV degradation mechanisms. However, as EPRI has updated the MDM and IMT, it is important to revisit research priorities and methodologies to reflect these changes. The revised MDM and IMT may introduce new factors affecting long-term RPV performance and safety, highlighting the need for continued research and updated guidance to ensure the reliable and safe operation of reactors beyond 80 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Analog, Digital, or Enhanced Human-System Interfaces? Results of an Operator-in-the-Loop Study on Main Control Room Modernization for a Nuclear Power Plant

The second Operator Study of System Overviews (OSSO-2) was conducted with a three-person reactor operator crew in the Human Systems Simulation Laboratory (HSSL) at Idaho National Laboratory (INL) in August, 2017. The study supported control room modernization at a nuclear power plant and featured a benchmark comparison of three variants of a turbine control system (TCS): the existing analog TCS, a proposed standalone digital TCS with two displays, and the digital TCS with the addition of a third display consisting of a system overview screen. TCS prototypes were developed at INL to allow evaluation of operator performance and preferences during realistic turbine scenarios in the full-scope simulator. The study revealed that completion of turbine startup was several minutes faster with the digital TCS variants than with the conventional analog TCS. Eye tracking scan paths were more widely distributed in the overview vs. standalone TCS condition, suggesting the overview screen was cueing reactor operators to verify values across the boards. There was no significant difference in workload or situation awareness across the three interfaces. Reviewing key plant parameters showed smoother transitions during load following for the digital vs. analog TCS. Despite some performance advantages for the digital TCS variants, operators preferred the existing analog TCS. Open-ended responses suggested this finding may be more an artifact of familiarity than a reflection of dissatisfaction with the new TCS. OSSO-2 provides compelling evidence that the new digital TCS was used successfully by the operators without extensive training of rewriting of the operating procedures, suggesting high usability for the digital TCS design. Further advantages were realized through the addition of the system overview screen to provide crews with at-a-glance indicators of key turbine parameters.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2018

Fuel cycle and neutronic performance of a spectral shift molten salt reactor design

Here, the fuel cycle performance and core design of the Transatomic Power liquid-fueled molten salt reactor concept is analyzed. This advanced reactor concept uses configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from intermediate at beginning of life to thermal at end of life. With a harder spectrum during the early years of reactor operation, this spectral shift design drives captures in fertile 238 U. The converted fissile plutonium makes up over 50% of the fissile material in the fuel salt over the last half (~15 years) of reactor operation. A softer spectrum late in reactor life helps drive the fuel to a burnup of 90 GWd/MTU. Continuously changing physics necessitates time-dependent analyses resolved over long timescales (i.e., months to years), as this concept does not meet an equilibrium condition. The spectral shift and molten salt reactor material feeds and removals enable this concept to perform better in fuel cycle metrics, increasing resource utilization by more than 50% compared with a typical light water reactor (i.e., from ~0.6% to ~1%). Finally, these metrics are compared to similar fuel cycles using alternate technologies. Additional core design and analysis challenges associated with the spectral shift and use of molten salt reactor technology are identified and discussed.

20 FOSSIL-FUELED POWER PLANTS↗

DOE OSTI2023

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↗

DOE OSTI2023

Accurate Dosimetry for Future Advanced Reactor Deployment and Operation

Accurate Dosimetry for Future Advanced Reactor Deployment and Operation from a customer's perspective. The main discussion points will be why we should be interested in this topic of nuclear renaissance, practical examples, and Dosimetry's role in advance reactor operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Process Model-Based Validation of the Intensification of Biomass Fast Pyrolysis in a Fluidized Bed via Autothermal Operation

A model for the fluidized-bed pyrolysis of biomass is extended to enable the simulation of intensified autothermal operation. In this system, partial oxidation of char and pyrolysate species replaces an external heat source to supply the enthalpy of pyrolysis, greatly increasing the throughput of a given fast pyrolysis reactor. Oxidation reactions are compiled from CRECK and other literature sources and extended to cover the species found in pyrolysis products, including a derivation of the catalytic effect of ash on char combustion from experimental studies at Iowa State University (ISU). Results indicate a roughly 3-fold increase in biomass throughput for a given reactor volume at the pilot scale with minimal loss of valuable products, in agreement with published data from the 3 × 10 -2 m 3 reactor operated by ISU. Additionally, a proposal for a 250 ton/day commercial-scale biomass pyrolysis reactor is analyzed, showing that the same size reactor operated by heating the fluidizing gas externally would have only one-tenth the capacity; if operated by heating the sand externally, the sand would need to circulate at an impractical 0.7 bed volume per minute rate to maintain a 250 ton/day biomass throughput.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2019

Innovative Implementation of Consolidated Dry Fuel Storage at the INL - 19655

An innovative nuclear-fuel storage bucket design provides increased fuel storage capacity at the Irradiated Fuel Storage Facility (IFSF) on the Idaho National Laboratory (INL) Site. The IFSF is a dry storage facility that is at approximately 93% of its fuel storage capacity. This innovative fuel storage bucket provides for higher density packaging, thereby allowing for the receipt of more fuel at the IFSF and subsequent completion of the wet-to-dry fuel transfer scope at the INL's Idaho Nuclear Technology and Engineering Center (INTEC). The subject fuel originates from the Advanced Test Reactor (ATR) at the INL Site. All spent nuclear fuel (SNF) at INTEC is required by the Idaho Settlement Agreement (ISA) to be out of wet storage by December 31, 2023. The ISA is a tri-party agreement between the U.S. Department of Energy, the U.S. Navy, and the State of Idaho. Currently, the INTEC CPP-666 pools contain all of the INL's SNF remaining in wet storage, and INTEC's Spent Fuel Project is on schedule to meet the ISA deadline. Interestingly, the INL Site's environmental cleanup contract awarded to Fluor Idaho in 2016 also initially required the Spent Fuel Project to continue SNF receipts from the ATR into wet storage at the CPP-666 pools. The continued receipts of ATR SNF into the CPP-666 basin were able to be redirected to the IFSF because of the new storage package design, enabling the Spent Fuel Project to meet the ISA requirement to have all SNF out of wet storage by 2023. Complicating the need to redirect ATR receipts from wet storage is the lack of interim storage capacity at the ATR site. This lack of ATR storage capacity threatened to terminate reactor operations if the SNF could not be received at INTEC. The contractual requirement for continued ATR SNF receipts to prevent termination of reactor operations is resolved by the implementation of the innovative consolidated dry storage configuration in the IFSF. Consolidated fuel storage at the IFSF provides increased fuel storage capacity to facilitate meeting the ISA and the Fluor Idaho contract requirements. It provides for the completion of the ATR wet-to-dry fuel transfer scope and supports continued operation of the ATR by allowing the receipt of additional ATR SNF directly into IFSF dry storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Signal Processing of Multiplexed Optical PWM Signals for Sensor Arrays in Nuclear Environments

Safe and effective generation of terrestrial nuclear power greatly benefits from the actionable data provided by the array of sensors located throughout a plant to provide a holistic online indication of reactor operation. This array includes sensors for monitoring coolant flow and pressure, temperature and heat transfer, radiation levels, structure health monitoring, and other critical parameters for reactor operation. While sensors capable of measuring these parameters have been developed, the electronics used in the pre-amplification and analog-to-digital conversion of the small signals they produce are extremely sensitive and susceptible to damage by high-temperatures and radiation environments nuclear reactors encounter while generating power. The small signals from sensors in nuclear power plants (NPPs) are transmitted over long cable runs which introduce dispersion artifacts into the signals of interest as well as electromagnetic interference (EMI) from lighting fixtures, pumps, mains electricity, and other equipment. To overcome these challenges, a front-end digitization (FREND) platform has been developed to use radiation-tolerant electronics to multiplex, amplify, and optically encode signals from an array of sensors for transmission over an optical fiber to mitigate dispersion and EMI artifacts from long runs of electrical cabling. To recover the optically transmitted data, a signal processing scheme based on 1-dimensional template matching to an indexing channel is described herein and demonstrated to have an effective bit-depth of 9.2 bits (1%). This scheme has been validated in proof-of-concept, non-nuclear testing and preliminary experimental results show good agreement between the measured optical output and and sensor input signals. The FREND system represents a low-loss data link between sensors in nuclear environments and data acquisition hardware which is aimed at improving the signal-to-noise ratio of the data acquired from these sensors to provide better information to operators and researchers.

Sweeney, Dan↗

DOE OSTI2024

Measurement of atomic oxygen densities using TALIF on a dielectric barrier discharge: insights into the volume above a micro cavity plasma array

Dielectric barrier discharges, particularly micro cavity plasma arrays, offer significant potential for plasma-catalytic research due to their ability to ignite plasma in direct contact with a catalytic surface, enabling the observation of plasma-surface interactions. A key factor in their application is the generation of reactive species, such as atomic oxygen, within the cavities. These species can interact with both the surface (e.g. for activation or cleaning) and the gas being treated (e.g. for oxidation). Given the central role of oxygen atoms in plasma catalysis and their use as a model for more complex species, this work investigates the transport of these atoms out of the cavities. Two-photon absorption laser-induced fluorescence spectroscopy with picosecond laser excitation is performed in the volume above the cavities. The results are compared with a basic diffusion model. The reactor operates with a He/O 2 mixture at a flow rate of 1 slm and atmospheric pressure. Densities of up to 10 16 cm -3 are measured near the surface. Time-dependent measurements show that, at a distance of 350 µm from the surface, a density equilibrium is reached within less than 3 ms of reactor operation. Decay times due to ozone formation after the reactor is turned off are on a similar scale. Spatially resolved measurements show that the oxygen density decreases exponentially from the surface but remains detectable up to approximately 1 mm above the surface, indicating significant application potential. Variations in the O 2 admixture show a density maximum at 0.4%, confirming previous helium state enhanced actinometry measurements within the cavities.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Cyber threat assessment of machine learning driven autonomous control systems of nuclear power plants

We report 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.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2024

Survey of Neutron Flux Sensors for Monitoring Advanced Reactor Concepts Operating with Low Neutron Fluence Rate and Extended Uninterrupted Lifespans

The purpose of this report is to provide a survey of neutron sensors applicable to the low power levels (low neutron fluence rate) for reactor monitoring and controls. A survey of detectors for reactor power monitoring and controls from various commercial vendors—Reuter Stokes, Framatome, Exosens, and Mirion—is presented in the following sections based on application of reactor power ranges. Detector specifications are provided where available per each of the vendor’s catalogues and specification sheets found on their websites. While this is not an all encompassing list, it provides an example of detector options. Fission chambers and self powered neutron detectors (SPNDs) for in-core application with changes to the neutron-sensitive materials—fissile depots for fission chambers and emitters for SPNDs—for regenerative capabilities will be discussed in the latter sections.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗