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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Development of Critical Experiments to Benchmark Moderator Temperature Reactivity Worth

The simulation of nuclear reactors requires being able to accurately model several different interrelated physical processes including neutronics, heat production, fluid/thermal physics, structural mechanics, fuel behavior, chemistry, and balance-of-plant. In order to design and use new types of nuclear reactors, modern codes have been developed which couple together the behavior of these different physical models to accurately consider the feedback effects between different physical effects. The SHARP toolset, part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project, is an example of an effort to combine models of neutronics, thermal-hydraulics, and structural mechanics in a coupled simulation to model any type of nuclear reactor as accurately as possible. The validation of multi-physics coupling techniques requires new experimental benchmark experiments specifically designed for that purpose. As a preliminary step towards design of these experiments, a series of measurements have been performed to demonstrate the temperature-dependent neutronic behavior of a well-characterized reactor assembly for comparison with the available codes. The facility adopted for these experiments is Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI). The RCF is unique in several aspects and provides several advantages to this type of experiment. Built in 1954 by the American Locomotive Company (ALCO), the RCF has been used to design small compact reactors. RPI acquired the facility in 1963, and the highly-enriched uranium fuel was later swapped for low-enriched fuel pins that were excess from the SPERT facility experiments. Nowadays, The RCF is regarded as a teaching and research facility, permitting the training of licensed Senior Reactor Operators, and offering a critical experiments laboratory course to students. The reactor core standard configuration consists of 332 or 333 UO{sub 2} fuel pins, containing a total of 35.2 grams of U- 235 each (4.81 % enriched). The pins are arranged in a regular lattice with a pitch of 1.6256 cm. The number and configuration of these pins are flexible, and the reactor's overall thermal power is limited to no more than 15 W. As the low thermal output of the reactor does not induce any measurable temperature changes, the effect of temperature in the reactor is observed by applying two 18- kW electric heaters to the water moderator, with mechanical agitators to maintain uniformity in moderator temperature. Before beginning experiments, each fuel pin inside the core was surveyed for mechanical defects (bending or warping) that may contribute to uncertainty in core behavior. Every fuel pin has been individually removed from the core and checked for any irregularities on its surface (caused by shocks or usage). About 12 % of the 333 pins were found to have a small degree of warping and were replaced with excess pins available in the facility. In doing this, we made sure all the pins are identical, thereby increasing the accuracy of future models. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Post-quench ductility of zirconium alloy cladding materials - 2016-0088

The ductility of zirconium-based alloys after steam oxidation at high temperatures, followed by quenching in water, is of significant interest for the nuclear industry. High-temperature steam oxidation and quenching of nuclear fuel cladding occurs during postulated loss-of-coolant accidents. The industry is currently establishing in-reactor operation limits for peak cladding temperature and time at temperature based on the measured ductile-to-brittle transition of different cladding materials, both in the as-fabricated condition and using hydrogen charging as a surrogate for in-reactor operation. Aiming to study possible differences in post-quench ductility caused by differences in chemical composition, extensive testing on Zircaloy-4, ZIRLO{sup R}, and Optimized ZIRLO{sup TM} cladding materials as a function of hydrogen content, extent of oxidation, and peak cladding temperature was performed. The results showed similar ductile-to-brittle transition for the alloys studied despite differences in chemical composition, with a decrease in the extent of oxidation to reach the ductile to brittle transition with increasing hydrogen content. Microstructural characterization was performed to investigate the evolution of the oxygen-enriched layer that forms underneath the oxide, both during the time at high temperatures and during cooling. Significantly different morphologies of the oxygen-enriched layer were observed with different cooling rates, particularly for alloys containing niobium. Possible mechanisms for the development of the oxygen-enriched layer are discussed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Post quench ductility of Zirconium alloy cladding materials - 6.2 STP-2016-0088

The ductility of zirconium-based alloys after steam oxidation at high temperature, followed by quenching in water, is of significant interest for the nuclear industry. High temperature steam oxidation and quenching of nuclear fuel cladding occurs during postulated loss-of-coolant accidents (LOCA). The industry is currently establishing in-reactor operation limits for peak cladding temperature and time at temperature based on the measured ductile-to-brittle transition of different cladding materials, both in the as-fabricated condition and using hydrogen charging as a surrogate for in-reactor operation. Aiming to study possible differences in post-quench ductility (PQD) caused by differences in chemical composition, extensive testing has been performed on Zircaloy-4, ZIRLO{sup R} and Optimized ZIRLO{sup TM} cladding materials as a function of hydrogen content, extent of oxidation and peak cladding temperature. The results show similar ductile-to-brittle transition for the alloys studied, in spite of the differences in chemical composition, with a decrease in the extent of oxidation to reach the ductile-to-brittle transition with increasing hydrogen content. Microstructural characterization was performed to investigate the evolution of the oxygen enriched layer that appears underneath the oxide, both during the time at high temperature and during cooling. Significantly different morphologies of the oxygen enriched layer are observed with different cooling rates, particularly for alloys containing niobium. Possible mechanisms for the development of the oxygen enriched layer are discussed (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Assess How Changes in Fuel Cycle Operation Impact Safeguards

Since the beginning of commercial nuclear power generation in the 1960s, the ability of researchers to understand and control the isotopic content of spent fuel has improved. It is therefore not surprising that both fuel assembly design and fuel assembly irradiation optimization have improved over the past 50+ years. It is anticipated that the burnup and isotopics of the spent fuel should exhibit less variation over the decades as reactor operators irradiate each assembly to the optimum amount. In contrast, older spent fuel is anticipated to vary more in burnup and resulting isotopics for a given initial enrichment. Modern fuel therefore should be more uniform in composition, and thus, measured safeguards results should be easier to interpret than results from older spent fuel. With spent fuel ponds filling up, interim and long-­term storage of spent fuel will need to be addressed. Additionally after long periods of storage, spent fuel is no longer self-­protecting and, as such, the IAEA will categorize it as more attractive; in approximately 20 years many of the assemblies from early commercial cores will no longer be considered self-­protecting. This study will assess how more recent changes in the reactor operation could impact the interpretation of safeguards measurements. The status quo for spent fuel assay in the safeguards context is that the overwhelming majority of spent fuel assemblies are not measured in a quantitative way except for those assemblies about to be loaded into a difficult or impossible to access location (dry storage or, in the future, a repository). In other words, when the assembly is still accessible to a state actor, or an insider, when it is cooling in a pool, the inspectorate does not have a measurement database that could assist them in re-­verifying the integrity of that assembly. The spent fuel safeguards regime would be strengthened if spent fuel assemblies were measured from discharge to loading into a difficult or impossible to access location. The primary driver for suggesting this shift in approach is the change in robotic technology and information technology in general. It should be possible, with minimal impact to the facility, to measure each assembly every time that it is moved in the pool, with the first measurements being made at discharge. The following conclusions were reached: The total neutron count rate can be accurately predicted at any future moment in time based upon the measured count rate at discharge, provided the initial enrichment and burnup of the assembly is known at discharge. It is expected that the total neutron count rate measured at discharge will be indicative of the initial enrichment and burnup of that assembly. If the automated robot were to focus on measuring the assemblies in the rack without moving them, the time available would increase immensely.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2016

Radiation tolerant joining for silicon carbide-based accident tolerant fuel cladding

Silicon carbide (SiC) and silicon carbide fiber reinforced, silicon carbide matrix (SiC-SiC) composites are candidate materials for nuclear fuel cladding in accident tolerant light water reactors and advanced high temperature reactors due to their chemical stability and retention of strength at elevated temperature and neutron flux. A reliable means to join silicon carbide components is imperative to the successful implementation of SiC-based structural nuclear materials as fuel cladding as well as more complex core internal structures. A comprehensive approach has been taken to address several of the significant challenges associated with SiC joining and fabrication of complex SiC structures. This work focuses on low activation joining materials that are expected to behave well under a variety of different reactor operating conditions. Pre- and post-irradiation mechanical testing and microstructural evaluation will be presented for specimens subjected to fast flux irradiation to 4.5 dpa at 730 C. degrees, as well as 8.7 dpa at 270 C. degrees. Joints did not exhibit any strength degradation under either irradiation condition. Additionally, the performance of joined cylindrical SiC-SiC composite specimens was evaluated following irradiation in pressurized water reactor (PWR) coolant chemistry to 0.5 dpa. Lastly, out-of-pile characterization of permeability has been conducted on joined cylindrical SiC-SiC specimens to assess performance of this joint material under additional conditions relevant to reactor operation. Details of the experimental set-up and relationships between microstructure and performance will be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2015

The decommissioning imperative - motivating progress in us used fuel management - 15185

When the Yucca Mountain project was terminated in 2010 there was approximately 3500 metric tons of used fuel stored at Independent Spent Fuel Storage Installations (ISFSIs) at which there was no operating reactor (10 shutdown plant sites and the GE Morris facility). By 2015, four additional sites with five reactors will have shut down, causing the amount of used fuel 'stranded' at sites where there is no operating reactor to nearly double to approximately 6900 metric tons which will be stored at a total of 15 sites. At each of these sites, once the reactors are decommissioned, the only thing preventing the land from being returned to communities for other uses is the remaining presence of the used fuel. The cost of maintaining 15 stand-alone ISFSIs is born by electric ratepayers and taxpayers. Going forward, the costs of maintaining stand-alone ISFSIs is certain to grow even if the number of stand-alone ISFSIs does not grow beyond the already announced shutdown of the Oyster Creek reactor in 2019 (which would increase the amount of stranded used fuel by another 10% and create a 16. stand-alone ISFSI). One reason for this cost escalation will be the need for ISFSI owners to renew storage licenses, and implement aging management programs to support these renewals, due to the continued delay in the federal program for removing the used fuel. Storage licenses at 10 of 16 stand-alone ISFSIs will expire by 2021, which is the earliest projected date at which the Department of Energy (DOE) has indicated it could have a pilot interim storage facility available - and nearly 3 decades before DOE projects a repository will be available. The burden and cost of continuing to renew storage licenses - and manage aging storage facilities - at 16 stand-alone sites instead of moving stranded used fuel to a single consolidated storage location or repository will likely be significant. Industry resources are increasingly being channeled into dry storage aging management activities as public and regulator interest in potential long-term degradation mechanism continues to grow. Whether the ultimate destination for used fuel is Yucca Mountain or some other location, the current stalemate preventing progress in the federal program to move it must be broken - and the increasing prominence of the decommissioned reactor quandary may very well be a pivotal factor in breaking that stalemate. (authors)

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

Epsilon Metal: A Waste Form for Noble Metals from Used Nuclear Fuel

Epsilon metal (e-metal) is the metallic phase that forms as inclusions at the grain boundaries in the UO2 fuel during reactor operation. This metal is composed of Pd, Mo, Rh, Ru, and Tc. These metallic inclusions are insoluble in strong acid and remnants of these metallic inclusions have been found in the UO2 matrix that remains from the natural reactors in Gabon that were active 1.8 billion years ago, therefore e-metal should be an excellent waste form for the immobilization of the long-lived isotopes 107Pd (6.5 ? 106 a) and 99Tc (2.13 ? 105 a), with 99Tc being the isotope of interest for repository performance. Therefore, the chemical durability of this potential waste form is assessed in this study. Typically, corrosion rates for metallic materials are measured electrochemically because they are quick, inexpensive, and can reveal the mechanism by which a metal corrodes, at least initially. However, in a repository the waste form would be subjected to slowly flowing water without an applied electrical potential over long time periods. Therefore, the corrosion rates of e-metal specimens were measured with both electrochemical tests and the single-pass flow-through test (SPFT). Potentiodynamic and potentiostatic polarization results suggest that a thin passive film exists on the alloy surface, which seems to be responsible for its high corrosion resistance. Additionally, X-Ray photoelectron spectroscopic results suggest that Pd oxides are significantly enriched in the passive film Results from the SPFT show that the dissolution rates were weakly dependent on pH. Only Mo and Re were found in solution and were used for the calculation of the dissolution rates. In general, the electrochemically determined corrosion rates agree reasonably well with the initial dissolution rate measured with the SPFT test, but they are about one or two orders of magnitude higher than the steady state rates. The causes for this discrepancy are discussed.

epsilon-metal, nuclear waste form, corrosion, sing↗