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DOE OSTI2017

Temperature compensated transfer function for probability of detection (POD) in high temperature ultrasonic NDE using low temperature signals - 249

Ultrasonic piezoelectric based transducers offer the potential for in-coolant NDT measurements and imaging required at high temperature for the safe operation of advanced small modular reactors (SMRs). Current work estimates a significant 97% reduction in the probability of detection (POD) for an artificial flaw size of 2{lambda} due to degradation of the transducer performance at elevated temperature. The source of variability for POD in the current work is the temperature dependence of material matrix of piezoelectric material PZT-5A used in the sensor. The physics based model developed using finite element method quantifies reduction in the scattering amplitude for the side drilled hole (SDH) from 15 C to 195 C. Using model assisted probability of detection (MAPOD) approach, sensor temperature effect on the POD is demonstrated for SDHs in the low carbon steel specimen. Scattering amplitudes for the room temperature simulated data is validated with the experimental data. A temperature correction factor and transfer factors are reported to transform the simulated temperature effect from physics based model to the experimental data for estimation of high temperature POD using low temperature ultrasonic measurements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Applying U.S. metal fuel experience to new fuel designs for fast reactors

With the increasing interest in small modular reactors or microreactors, developers are working to design and submit licensing approval requests of U–10Zr-fueled fast reactors. The developers and their proponents cite prior metal fuel experience (worldwide, but U.S. experience in particular for many developers) as the motivation and justification for their reactor concepts. The experience with metal fuel deployment in sodium-cooled fast reactors as well as the underlying irradiation testing database, provide a suitable basis for analytically justifying the use of metal fuel in new reactors. The evolution of metal fuel design and capability illustrates the importance of key fuel design parameters to consider in new applications of the prior experience: fuel smeared density, plenum-to-fuel volume ratio, the ratio of cladding radius to thickness, fuel composition, and cladding and duct materials. In-service operating and deployment conditions to be considered include fuel linear heat generation rate, fuel temperature, cladding temperature, peak burnup and peak fast fluence. Fuel designs and in-service conditions that are bounded by the database and experience are most easily addressed, but deviations from those previous parameters and conditions can be addressed by considering impacts on previously established behavior and applying other mitigating conservatisms, as appropriate. Here, the authors recommend any new deployment proceed with fuel surveillance and monitoring to mitigate risk, application of conservative measures to address uncertainties, and a fuel qualification program that addresses a range of in-service operating conditions with production fuel. The work reported should be of interest to students and regulators unfamiliar with metal fuel in fast reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Simulating self-powered neutron detector responses to infer burnup-induced power distribution perturbations in next-generation light water reactors

Understanding how 3D power distribution will be monitored throughout reactor core volumetric space in next-generation nuclear power reactors is crucial to the design, deployment, and licensing of these reactors. Although numerous techniques exist for 3D power distribution monitoring based on the response of both in situ and ex situ sensors currently implemented or proposed for use in the US reactor fleet, crucial details about these techniques are often unclear. The publicly available documentation does not include information such as how well these techniques are characterized and optimized in their implementations and the levels of uncertainty in the inferred 3D power distribution. The work described herein investigated a recently developed 3D power distribution inferencing method as applied to two next-generation reactor simulations: (1) the NuScale small modular reactor design and (2) the Westinghouse AP1000 design, both of which contain in-core strings of vanadium self-powered neutron detectors (SPNDs). This investigation considered a range of SPND string sensor densities, as well as a range of 3D power distribution axial segment sizes. In this work, SPND response simulation is informed by neutron flux calculations in representative homogenized cores. For the different sensor densities and power distribution axial segment sizes in these simulations, the average solution error, solver iterations, and run time were tracked to parameterize the sensor-core configuration.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Advancement of Supercritical Carbon Dioxide Technology through Round Robin Testing and Fundamental Modeling

Growing interest in supercritical carbon dioxide (sCO 2 ) cycles for advanced reactors is driving the need for corrosion data on candidate plant materials. The sCO 2 Brayton cycle is being considered for power conversion systems for a number of nuclear reactor concepts including the sodium fast reactor, fluoride salt-cooled high temperature reactor, high temperature gas reactor, and several types of small modular reactors. Multiple organizations have developed test facilities to address the corrosion data knowledge gap in high temperature, high pressure sCO 2 environments but, to date, there has been no formal test program among these organizations to validate the consistency of the data they produce. A demonstration of comparable and reproducible results enables a coordinated effort to explore the sCO 2 parameter space relevant to advanced reactor technology. This proposal establishes a round robin test plan for sCO 2 corrosion testing and the organization of a sCO 2 Materials Group to guide future materials testing directions. Furthermore, this proposal outlines efforts to elucidate the mechanisms of sCO 2 corrosion by performing identical tests in supercritical steam, testing model alloys with various composition and environmental conditions, accompanied by modeling efforts to study rate controlling mechanisms of corrosion and to help understanding the effects of impurities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Molten Salt Reactor Signatures and Modeling Study

Molten salt reactor (MSR) technologies, either liquid fueled and cooled or only liquid cooled, pose specific, unique challenges for safeguards of the special nuclear material during the operation, fueling, and maintenance of the reactor. MSRs are one type of Generation IV technologies being invested in and considered for U.S. domestic fabrication primarily for electricity and process heat production. These designs have generated growing commercial interest for several reasons, including high (≈40%) thermal efficiency, ease of fueling, improved use of uranium fuel, potential utilization of thorium fuel, and proposed inherent safety features. U.S. companies have several planned designs that differ in fuel, cooling, and neutron energy spectrum. Driven by commercial interest and the intent of licensing MSRs, the U.S. Nuclear Regulatory Commission (NRC) has developed a vision and strategy to accommodate non-light water reactors (LWR), which include MSRs (U.S. Nuclear Regulatory Commission 2019). In addition, the Department of Energy (DOE) through the Office of Nuclear Energy (NE) Office of Advanced Reactor Technologies (ART), …sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. Reactor types considering the use of salts, liquid metals, or gases for coolant fall under both ARC and aSMR. Therefore, Research Design & Development is being pursued by DOE-NE through national laboratories, universities, and international and industrial collaborations. Additionally, the U.S. is a member of the Gen IV International Forum (GIF). The GIF is a cooperative, multinational organization to guide and carry out research and development needed for the GEN IV reactor systems (Forum 2018). GIF evaluated numerous reactor concepts and down-selected to the six most feasible advanced reactor technologies: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), MSR, supercritical watercooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). In support of the growing interest domestically and internationally, the Materials Protection, Accounting, and Control Technologies (MPACT) campaign, under the DOE-NE Fuel Cycle Technologies (FCT) program, engages in R&D activities by developing advanced instrumentation and analysis for safeguards and security of modern, advanced nuclear fuel cycle (non-LWR) facilities. Because of the historic experience in the operation of the Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) (Robertson, MSRE Design and Operations Report Part I 1965), Oak Ridge National Laboratory (ORNL) is heavily engaged in the various R&D activities through the DOE complex related to MSRs including national technical leadership of the DOE-NE MSR campaign. This report discusses and presents the outcomes of the FY19 MPACT MSR Safeguards task. The challenges presented by MSRs for nuclear material accountancy and control (NMAC) and associated safeguards will be investigated. The objective of this research is to explore and compile the safeguards requirements and identify measurement signatures through an initial high-level MSR design and develop complementary advanced simulation and modeling capabilities. A high-level ORNL-developed MSR design called the Molten Salt Demonstration Reactor (MSDR) (Bettis, Alexander and Watts 1972) was used as the target reactor design for this research. The MSDR model incorporates technology from the MSRE and the Molten Salt Breeder Reactor (Robertson, Conceptual Design of a Single-Fluid Molten-Salt Breeder Reactor 1971). But the MSDR is a 750 MWth graphite moderated liquid fueled (low-enriched uranium) MSR compared to the MSRE’s of 7.5 MWth. The focus of this report is to discuss the evaluation of novel signatures, correlations, and indicators to understand the applicability of current safeguards instrumentation to MSRs using the modeling results from the MSDR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Integration of Wireless Sensor Networks and Battery-free RFID for Advanced Reactors

To address an important need for Nuclear Power Plants (NPPs) to significantly reduce the amount of the required cables for sensor data communications, this Phase I SBIR effort successfully developed and demonstrated a novel low-cost proof-of-concept prototype of a secure wireless sensor network backbone communications system. This system combines commercially available low power, low cost XBEE wireless communications network and passive (battery-free) Radio Frequency Identification (RFID) systems to report individual sensor data and their location through the containment wall for rapid response to anomalies in nuclear facilities. This WIreless Sensing and Locating (WISLO) system network architecture allows non-intrusive wireless collection of sensor data with the sensor’s accurate location information from inside of the instrumentation or containment area with minimal need for power sources. The multi-node sensor data in the containment area is wirelessly transmitted to outside through the metal reinforced concrete walls without the need for any batteries. The WISLO system could be readily used in current reactor fleet and future advanced reactors, as well as in small modular reactors (SMR). The WISLO system is the first dual frequency battery-free trough the wall communications backbone system developed specifically for use in nuclear power plants. It not only can modernize the sensor monitoring practices in the existing reactor fleet, but also offers a secure, low-cost solution to advanced reactors and SMRs by removing cables and issues related to them such as cable integrity, reporting delays, and installation and maintenance costs. The intent is to improve process safety, reliability, efficiency, and cost effectiveness to the monitoring and maintenance process in current and future plants. As we address the next productization and manufacturing capabilities in the next phase, this system will serve the needs of many commercial and government applications in remote monitoring of sensor data that demand battery-free transmission of sensor data such as Internet-of-Things (IOT).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Economics of Hybrid-Energy Systems and the Impact on Advanced Reactor Design and Operation

The economics of electricity production in the U.S. is changing with the introduction of recent federal regulations limiting carbon dioxide emissions in the electric power industry. This has already resulted in the closing of coal plants, plants which in the past to a degree served to balance production with daily variation in load. The resulting loss of electricity production is being made up with wind and solar renewables and gas-fired units, and with an increasing need for energy storage. It would appear that with the loss of the baseload component of the coal plants, the duty-cycle of the gas-fired units may necessarily increase, encroaching on the traditional nuclear plant domain with its inexpensive fuel. Gas plants are costly to operate on a continuous basis with relatively volatile fuel prices. It appears then that nuclear plants that are capable of flexible operation have the potential on an economic basis to displace gas-fired power units. For flexible operation with renewables advanced reactors have advantages over the current generation of pressurized water reactors (PWR) and the new small modular reactors (SMR). For PWRs and SMRs, pellet-cladding mechanical interaction, xenon poisoning, and thermal stress concentrations in structures place limits on the all-important response time for operation with renewables. These phenomena are, however, not as limiting for the liquid-metal fast reactor (LMR). This may permit greater power ramp rates than in a light water reactor which bestows an economic advantage in energy systems with renewables. An economic model is developed for studying electric grid options including plant mix and market rules for the case of a high renewable fraction. It returns the minimum cost solution from among a mix of plant types including nuclear, gas, and storage (i.e. hybrid energy systems) for a given fraction of electricity production by renewables. This model can be used to assess the economic advantage of nuclear plants over gas plants and advanced reactors over light water reactors in an electric grid with a high fraction of renewables. The model has an explicit treatment of plant dynamic effects, an important feature for analysis of hybrid energy systems with a renewables production component that is unpredictable and uncontrollable.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

DOE OSTI2025

Development of Predictive Models for Advanced Reactor Autonomous Control

Advanced reactor designs including microreactors and small modular reactors will contribute to the clean production of cheap energy, and autonomous control for advanced reactors is an appealing option for reducing cost. However, there is a lack of industry experience applying autonomous control for advanced nuclear reactors. To accelerate the development and industry acceptance of autonomous control software for nuclear reactors, we aim to demonstrate autonomous control of the Purdue University research reactor (PUR-1) using INL-developed model predictive control (MPC) methods. To prepare for this demonstration, data-driven predictive models based on process data collected from PUR-1 have been developed and integrated with MPC and used to control a physics-based model of PUR-1. A data-driven dynamics model and a gated recurrent unit (GRU) network were both trained on process data from PUR-1. The dynamics model was shown to effectively control the reactor model with MPC when provided reactivity as a control variable but failed to control the model through the control rod positions. The GRU network produced more accurate predictions than the dynamics model when evaluated on operational data, and future work will include the evaluation of the GRU network in the controller.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Operational performance risk assessment in support of a supervisory control system - 251

A supervisory control system is being developed for multiunit advanced small modular reactors to minimize human interventions during normal and abnormal operations. In the supervisory control system, control action decisions are made based on a probabilistic risk assessment that employs event trees and fault trees. Although traditional probabilistic risk assessment tools are implemented, their scope is extended to normal operations, and the application is reversed to assess the success of non-safety related systems and to enable continued operation of the plant. This extended probabilistic risk assessment approach is called operational performance risk assessment (OPRA). OPRA helps to identify available paths, combine control actions for maintaining plant conditions within operational limits, and to quantify the likelihood of success of these operational trajectories to optimize the selection of alternative actions without activating reactor protection system. In this paper, a case study of OPRA in a supervisory control system is demonstrated for the Advanced Liquid Metal Reactor (ALMR) Power Reactor Inherently Safe Module (PRISM) design, specifically the power conversion system. The scenario investigated involved a condition in which the feedwater control valve that was observed to be drifting to the closed position. Alternative plant configurations that would allow the plant to continue to operate at full or reduced power were identified using OPRA. Dynamic analyses were performed with a thermal-hydraulic model of the ALMR PRISM system using Modelica to evaluate the magnitude of safety margins. Successful recovery paths for the selected scenario were identified and quantified using the supervisory control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Demonstration of Passive Phoretic Decontamination by Accident Simulation in a Scaled Reactor Containment

Reactor containment performance is a critical defense-in- depth barrier for the prevention of radioactive release in the event of a nuclear accident. One of the functions of the containment is to assist in the decontamination of aerosolized radioactive particles through both 'passive' or natural processes and 'active' or mechanical processes. Current United States Nuclear Regulatory Commission (NRC) guidance (Reg. guide 1.183) allows the use of aerosol natural deposition correlations to quantify postaccident nuclear containment decontamination factors. Large Light Water Reactor (LLWR) correlations have previously been developed to quantify the contribution of natural phenomena to post-accident aerosol decontamination. Specific correlations must similarly be developed to support the safety analyses of Small Modular Reactors (SMRs). For LLWR plants, decontamination is achieved 'actively' by mechanical systems such as containment sprays and 'passively' via natural phenomena such as gravitational settling, condensation, impaction, thermophoresis, diffusiophoresis and hygroscopic effects. Integrated pressurized water reactor (iPWR) SMR designs do not have containment sprays; therefore, they rely primarily on passive natural phenomena for decontamination in the integrated reactor vessel. Analytical studies and inferences from LLWR test data suggest that the smaller containment vessel volume to surface area ratio, coupled with the thermal hydraulic properties of iPWR plants, would significantly enhance decontamination factors for these smaller designs. The research aims to quantify the amount of decontamination from natural phenomena for the range of iPWR physical and thermal hydraulic characteristics. A preliminary assessment has shown that diffusiophoresis, thermophoresis and hygroscopic effects are the major areas with a research gap for iPWR plants; hence these areas are the foci of the research. These first-of-a-kind experiments will be described and preliminary analytical results will be presented that show the expected effects of SMR-specific correlations on passive decontamination of aerosolized radionuclides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Code Validation of a SBLOCA test in SMART Integral Test Loop

Integrated small modular reactors (iSMRs) are favored because of their unique characteristics, such as enhanced safety without any large break loss of coolant accidents (LBLOCAs), low initial construction costs, and multi-use of desalination and electricity generation. Many iSMRs have been developed thus far. In Korea, an iSMR called SMART was developed after a thorough validation of the components and systems, and finally received its design certificate in 2012 for the first time in the world. To ensure the performance of its safety systems, an integral test loop, SMART-ITL, is constructed. It contains the primary system, secondary system, passive safety systems, and an auxiliary system. The passive safety systems include a passive residual heat removal system (PRHRS), passive injection system with a core makeup tank (CMT) and a safety injection tank (SIT), and an automatic depressurization system (ADS). The height is preserved while the diameter is scaled down to 1:7. A series of SBLOCA tests have been performed with 4 trains of PRHRS, 1 train of CMT and SIT, and ADS. In parallel with the experiments, simulation of a representative test has been conducted to track the underlying thermal hydraulics. SMART-ITL is an integral test loop for SMART and a series of SBLOCA tests for confirming the performance of passive safety systems have been conducted. In this paper, a system code simulation on a SBLOCA test using SMART-ITL is presented. The reactor vessel water level, break flow rate, and CMT and SIT water levels are predicted well. (authors)

42 ENGINEERING↗

DOE OSTI2018

Solutions and Challenges Posed by the Increasing Use of Technology Within the Nuclear Industry - 18023

Whereas the increasing influence of technology within the nuclear industry can be seen by many as providing significant advancements and solutions across many areas of the sector - it can also be seen to conversely represent a considerable number of challenges. Unlike the well-publicised technological advancements that have accelerated both operations and decommissioning at nuclear facilities around the world, the challenges that technology have brought to the nuclear industry are either not fully understood or have yet to be entirely assessed. One of the greatest technological benefits to the nuclear industry exists from the increased use of robotic technologies - specifically unmanned aerial vehicle (UAV) systems and other autonomous platforms, as part of routine and emergency site monitoring. The use of this technology by other (unauthorised) parties, however, represents a significant challenge, and one that has yet to be fully un-assessed. Other technologies, their advantages and challenges within the nuclear industry, discussed in this work include the potential / growing use of Small Modular Reactor (SMR) systems and the 'Internet of Things' (IoT). (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

SMR Current Status: Development Needs and Global Perspectives

Defined as nuclear reactors with a power output up to 300 megawatts electrical (MWe) by the International Atomic Energy Agency (IAEA) and targeted for multipurpose applications, small modular reactors (SMRs) have been recognized as a very promising, clean, affordable, and sustainable energy source by many countries. At present, more than 80 SMRs are under design, development, demonstration, deployment, and beyond (4D+) phases worldwide. This study focuses on the current world status of SMRs and focuses on the necessary developments to accelerate the process of adopting SMRs as a major energy source globally. SMRs are not a new concept, but they do represent a new vision for older concepts if the challenges inherent within them are mitigated with strategic and realistic solution approaches. The major challenges for SMRs 4D+ like any new reactors are: (a) qualifying the advanced fuel-to-reactor design; (b) supporting rapid scaled/prototypic experimentations; (c) maintaining local and global codes, standards, and licensing; (d) supply chain issues;(e) effective cradle-to-grave nuclear fuel cycle and fuel material transportation, and (f) mitigating financial and environmental risks. These upfront challenges can be mitigated with a synergistic solution approach among the various stakeholders: industry, academia, research, government, and international entities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Presentation: SMR Current Status: Development Needs and Global Perspectives

Defined as nuclear reactors with a power output up to 300 megawatts electrical (MWe) by the International Atomic Energy Agency (IAEA) and targeted for multipurpose applications, small modular reactors (SMRs) have been recognized as a very promising, clean, affordable, and sustainable energy source by many countries. At present, more than 80 SMRs are under design, development, demonstration, deployment, and beyond (4D+) phases worldwide. This paper focuses on the current world status of SMRs and focuses on the necessary developments to accelerate the process of adopting SMRs as a major energy source globally. SMRs are not a new concept, but they do represent a new vision for older concepts if the challenges inherent within them are mitigated with strategic and realistic solution approaches. The major challenges for SMRs 4D+ like any new reactors are: (a) qualifying the advanced fuel-to-reactor design; (b) supporting rapid scaled/prototypic experimentations; (c) maintaining local and global codes, standards, and licensing; (d) supply chain issues; (e) effective cradle-to-grave nuclear fuel cycle and fuel material transportation, and (f) mitigating financial and environmental risks. These upfront challenges can be mitigated with a synergistic solution approach among the various stakeholders: industry, academia, research, government, and international entities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

SMR Current Status: Development Needs and Global Perspectives

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Design and Operation of an Electric Peaking/Industrial Steam Production Unit for SMRs

Various methods of Thermal Energy Storage (TES) can be coupled to nuclear (or renewable) power sources to help absorb grid variability caused by daily load demand changes and renewable intermittency. Our previous research has shown that coupling a sensible heat TES system to a Small Modular Reactor (SMR) allows the reactor to run at effectively nominal full power during periods of variable electric demand by bypassing steam to the TES system during periods of excess capacity. In this paper we demonstrate that this stored thermal energy can be recovered; allowing the TES system to act as a peaking unit during periods of high electric demand, or used to produce steam for ancillary applications such as desalination. For both applications the reactor can operate continuously at approximately 100% power.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

FAST Irradiations, Postirradiation Examinations, and Modeling of U-Mo for Light Water Reactor Applications

Many next generation light water reactor (LWR) concepts, such as mobile small modular reactors, are seeking to use smaller core dimensions than conventional reactor types. Smaller reactor cores require an increase in fissile material to maintain reactivity. For non-proliferation purposes, enrichment increases are limited to less than 20% (high assay low enriched uranium, [HALEU]) and so higher uranium density fuels than UO 2 must be considered. To this end, uranium-molybdenum alloys were tested using the Fission Accelerated Steady-state Test (FAST) approach. The experiment test matrix is focused on identifying the temperature transition between low swelling and high fission gas retention to break away swelling and low fission gas retention. This paper documents the results of irradiation tests and post-irradiation examinations (PIE) including neutron radiography, rodlet profilometry, fission gas collection analysis, and optical metallography. The results of these tests showed that unconstrained U-Mo fuels (solid, Na-bonded rodlets) have a swelling threshold between 400-450°C with minimal fission gas release below this point. Higher temperature solid fuel showed microstructural zoning with small pore networks while lower temperature solid fuels have a uniform microstructure with large pore networks. U-Annular Mo fuels where swelling had some self-constraint imposed upon it, were shown to have much reduced swelling compared to their solid counterparts as well as very low fission gas release for irradiation temperatures up to 500°C. These initial results show that the use of U-Mo in constrained fuel geometries could be used as a high uranium density HALEU fuel for LWRs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Securing Future Energy Supplies: From Renewables to Microreactors

This session will provide insight into how future energy deployments, critical to national-level programs focused on reducing carbon emissions, can be secured-by-design using lessons learned from current energy infrastructure. It will begin with an overview of current threats and risks associated with renewable energy assets and systems, primarily wind and solar, focusing on their control architecture and key system functions for both efficient and safe operations. This talk will then translate the key takeaways from current renewable infrastructure into applications for securing future energy systems, including microreactors and small modular reactors (SMRs), based on planned concepts of operations and control. Microreactors and SMRs are intended to be factory-assembled with commercially available components and deployed in more remote or distributed environments, necessitating centralized control centers, remote monitoring, and offsite maintenance and technical support. All of these factors lead these assets to a security posture and controls more similar to today's renewable energy assets than today's nuclear reactors, which represents a significant shift in mindset for the nuclear industry. This talk will provide justification for this shift as well as a path forward to motivate securing these groundbreaking technologies from the outset of their design and deployment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗