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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 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Data recovery via covert cognizance for unattended operational resilience

One of the important premises of unattended operation, a highly promoted characteristic of fission batteries and advanced microreactors, is the ability to automate the analysis of sensors data used in support of operational monitoring and control. Here, to meet this vision, this work proposes a new monitoring and data recovery paradigm to ensure resilience against data corruption which may be the result of malicious intrusion into the reactor operational network. This is paramount to ensure 100% availability under contingency scenarios such as cyberattacks. In support of this vision, earlier work has presented the concept of covert cognizance and demonstrated its mathematical ability to identify and embed cognizance parameters under the noise-dominated null space of the sensors data. This work extends this concept and applies it in real-time to demonstrate three key characteristics: zero-impact, zero-observability, and data recovery, where the first characteristic is to ensure no impact on operation, the second is immunity to discovery by pattern recognition techniques, and the third is to allow recovery of corrupt or falsified data. Recognizing that fission batteries are designed to operate under steady state most of the time, we elect to employ a small modular reactor model under transient operational conditions to demonstrate the operational resilience enabled by the covert cognizance paradigm. Specifically, the PI controller is augmented with the covert cognizance modules to develop self-awareness and enable automatic data recovery. The developed modules are expected to be equally applicable to a wide range of advanced reactor technologies relying on full or partial unattended control.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
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 OSTI2023

Accident event progression, gaps, and key performance indicators for steam generator tube rupture events in water-cooled SMRs: A review

According to historical records of reactor-related incidents, a steam generator tube rupture (SGTR) is one of the most common occurrences at operating pressurized water reactors (PWRs). Such design-basis accidents (DBAs) could lead to a direct path for radionuclides to be released to the atmosphere via the safety and relief valves, making assessments of radionuclide discharge from operating nuclear power plants (NPPs) into the environment crucial for ensuring safety. In such analyses, the primary focus has been on the extent of radioactive release, not the potential damage to the core. Moreover, significant and timely intervention by operators is needed during the initial stages of SGTRs in order to avert overfilling of the SGs, as well as to restrict the dissemination of radioactive materials. Determining the event sequence and phases that occur during an SGTR incident in an advanced passive (e.g., AP1000) water-cooled nuclear reactor is crucial for implementing effective passive safety systems (PSSs) for a specific water-cooled small modular reactor (SMR) design. Here this study provides a comprehensive review of the accident event progression, associated physical phenomena, knowledge gaps, and key performance indicators that must be evaluated and assessed in thermal-hydraulics models, based on relevant test data, respectively.

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 OSTI2026

Investigating the impact of a multi-module operation environment on the task performance time of human operators – An explanatory study

The worldwide demand for Small Modular Reactors (SMRs) has surged in recent years due to their enhanced safety and versatility in supporting diverse industrial sectors. A unique feature of SMR operation is that a single human operator is responsible for managing multiple modules. Therefore, securing a sufficient amount of human performance data pertaining to this new environment is essential for the safe operation of SMRs. In this explanatory study, a series of experiments were conducted using the NuScale simulator, a representative SMR design, with student operators. A total of 12 student operators were assigned two types of off-normal events and asked to cope with them using paper-based procedures. Subsequently, their task performance times were compared with those of student operators responsible for a single unit based on the Task Complexity (TACOM) measure. Results indicate that the performance of student operators under the experimental conditions of this study degraded by a factor of 2 to 3, depending on the characteristics of the off-normal events.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Advanced nuclear reactor integration opportunities for the pulp and paper industry in the U.S. context: Technical perspectives, gap analysis, and preliminary technoeconomic assessment

Pulp and paper (P&P) manufacturing requires a large amount of low-pressure (LP) steam to digest, wash wood fibers and dry pulp into paper. Most of the LP steam is extracted from backpressure turbines that produce power from high-pressure (HP) steam. This HP steam is generated from burning wood waste material; bark is burned in hog boilers, and lignin is boiled in a black liquor recovery boiler. In a typical integrated P&P mill, 50–100% of the steam is produced from these sources, while additional steam is produced in natural gas (NG), fuel oil, or coal boilers. The other energy-intensive process in the plant is the chemical-recovery section (e.g., lime kiln), which requires high-temperature processing from NG combustion to retrieve and recirculate spent chemicals. This paper assesses the energy and heat demand and material balances of a typical generic kraft pulp mill, along with the nuclear heat, steam, and power integration opportunities to replace conventional combustion systems. The paper also addresses steam and electricity generation through a comprehensive technical and engineering gap analysis of five different nuclear-integration opportunities and their process economics, thus enabling the lignin and bark to be further processed into biobased chemicals or fuels, as well as the potential to reduce overall emissions from kraft pulping. Preliminary findings have shown that the P&P industry could achieve technological benefits by integrating their current manufacturing process with small modular nuclear reactors (SMNRs) on a national level. This research aims to set the path forward for a cleaner and more resilient P&P industry.

Advanced nuclear reactor↗
DOE OSTI2021

Application of sensitivity analysis in DYMOND/Dakota to fuel cycle transition scenarios

The ability to perform sensitivity analysis has been enabled for the nuclear fuel cycle simulator DYMOND through its coupling with the design and analysis toolkit Dakota. To test and demonstrate these new capabilities, a transition scenario and multi-parameter study were devised. The transition scenario represents a partial transition from the US nuclear fleet to a closed fuel cycle with small modular LWRs and fast reactors fueled by reprocessed used nuclear fuel. Four uncertain parameters in this transition were studied – start date of reprocessing, total reprocessing capacity, the nuclear energy demand growth, and the rate at which the fast reactors are deployed – with respect to their impact on four response metrics. The responses – total natural uranium consumed, maximum annual enrichment capacity required, total disposed mass, and total cost of the nuclear fuel cycle – were chosen based on measures known to be of interest in transition scenarios and to be significantly impacted by the varying parameters. Furthermore, analysis of this study was performed both from the direct sampling and through surrogate models developed in Dakota to calculate the global sensitivity measures Sobol’ indices. This example application of this new capability showed that the most consequential parameter to most metrics was the share of new build capacity that is fast reactors. However, for the cost metric, the scaling factor of the energy demand growth was significant and had synergistic behavior with the fast reactor new build share.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

A compact 10 kW, 476 MHz solid state radio frequency amplifier for pre-buncher cavity of free electron laser injector linear accelerator

A 10 kW, 476 MHz, 0.1% duty cycle solid state RF amplifier system for driving sub-harmonic, pre-buncher cavity of IR-FEL injector LINAC, has been developed at RRCAT. The 10 kW power is achieved by combining output of eight 1400 W amplifier modules using 8-way planar corporate combiner. The solid state amplifier modules have been developed using 50 V RF LDMOS transistors which although meant for push-pull operation are being used in single ended configuration with matching circuit developed on a thin (25 mils), high dielectric constant (9.7), low loss microwave laminate with an aim to have a compact structure. Ease of fabrication, modularity, small size, and low cost are the important features of this design which could be used as a template for low duty cycle medium to high pulsed power UHF amplifier system.

43 PARTICLE ACCELERATORS↗
DOE OSTI2017

Stand-alone polarization-modulation infrared reflection absorption spectroscopy instrument optimized for the study of catalytic processes at elevated pressures

This article describes the design and construction of a compact, “user-friendly” polarization-modulation infrared reflection absorption spectroscopy (PM-IRRAS) instrument at the Center for Functional Nanomaterials (CFN) of Brookhaven National Laboratory, which allows studying surfaces at pressures ranging from ultra-high vacuum to 100 Torr. Surface infrared spectroscopy is ideally suited for studying these processes as the vibrational frequencies of the IR chromophores are sensitive to the nature of the bonding environment on the surface. Relying on the surface selection rules, by modulating the polarization of incident light, it is possible to separate the contributions from the isotropic gas or solution phase, from the surface bound species. A spectral frequency range between 1000 cm -1 and 4000 cm -1 can be acquired. While typical spectra with a good signal to noise ratio can be obtained at elevated pressures of gases in ~2 min at 4 cm -1 resolution, we have also acquired higher resolution spectra at 0.25 cm -1 with longer acquisition times. By way of verification, CO uptake on a heavily oxidized Ru(0001) sample was studied. As part of this test study, the presence of CO adsorbed on Ru bridge sites was confirmed, in agreement with previous ambient pressure X ray photoelectron spectroscopy studies. In terms of instrument performance, it was also determined that the gas phase contribution from CO could be completely removed even up to pressures close to 100 Torr. A second test study demonstrated the use of the technique for studying morphological properties of a spin coated polymer on a conductive surface. Note that this is a novel application of this technique. In this experiment, the polarization of incident light was modulated manually (vs. through a photoelastic modulator). It was demonstrated, in good agreement with the literature, that the polymer chains preferentially lie parallel with the surface. This PM-IRRAS system is small, modular, and easily reconfigurable. It also features a “vacuum suitcase” that allows for the integration of the PM-IRRAS system with the rest of the suite of instrumentation at our laboratory available to external users through the CFN user proposal system.

36 MATERIALS SCIENCE↗
DOE OSTI2018

A roadmap for nuclear energy technology

The prospects for the future use of nuclear energy worldwide can best be understood within the context of global population growth, urbanization, rising energy need and associated pollution concerns. As the world continues to urbanize, sustainable development challenges are expected to be concentrated in cities of the lower-middle-income countries where the pace of urbanization is fastest. As these countries continue their trajectory of economic development, their energy need will also outpace their population growth adding to the increased demand for electricity. OECD IEA’s energy system deployment pathway foresees doubling of the current global nuclear capacity by 2050 to reduce the impact of rapid urbanization. The pending “retirement cliff” of the existing U.S. nuclear fleet, representing over 60 percent of the nation’s emission-free electricity, also poses a large economic and environmental challenge. To meet the challenge, the U.S. DOE has developed the vision and strategy for development and deployment of advanced reactors. As part of that vision, the U.S. government pursues programs that aim to expand the use of nuclear power by supporting sustainability of the existing nuclear fleet, deploying new water-cooled large and small modular reactors to enable nuclear energy to help meet the energy security and climate change goals, conducting R&D for advanced reactor technologies with alternative coolants, and developing sustainable nuclear fuel cycle strategies. Since the current path relying heavily on water-cooled reactors and “once-through” fuel cycle is not sustainable, next generation nuclear energy systems under consideration aim for significant advances over existing and evolutionary water-cooled reactors. Among the spectrum of advanced reactor options, closed-fuel-cycle systems using reactors with fast-neutron spectrum to meet the sustainability goals offer the most attractive alternatives. However, given these circumstances, unless the new public-private partnership models emerge to tackle the licensing and demonstration challenges for these advanced reactor concepts, realization of their enormous potential is not likely, at least in the U.S.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2023

Modeling and control of nuclear–renewable integrated energy systems: Dynamic system model for green electricity and hydrogen production

The need for decarbonization and diversification of energy resources has led to the development of integrated energy systems (IESs), where multiple resources supply more than one energy sector. Here, one such IES with small modular nuclear reactors and renewables (wind and solar) as generating resources, catering to the demand of the electric grid while producing hydrogen for industries, is modeled in this paper. The physics-based component models are represented using the Modelica language and interconnected to form the IES. The control and coordination of the overall system are ensured by designing a suitable control architecture composed of individual subsystem-level controls and supervisory control. The dynamic performance and the load-following capability of the IES are evaluated, while satisfying the safe operational limits of the components. Different configurations and modes of IES operation are considered, where the adaptability of the control system in the presence of varying demands and renewable generations is validated. The simulation results indicate that hydrogen as a flexible load facilitates the supply of varying grid demand. Additionally, the renewables are also accommodated into the IES owing to the flexibility of the balance of plant associated with the nuclear reactors.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2019

Numerical Simulation of Isothermal Flow Across Slant Five-Tube Bundle with Spectral Element Method Code Nek5000

The helical coil steam generator (HCSG) is a specific type of shell-and-tube heat exchanger known for having a higher heat transfer coefficient than many other designs. For this reason, they are considered in small modular reactor and high temperature reactor designs. Investigation of flow behavior in HCSGs is important for better design. In this paper we present our study for modeling the coolant flow in the primary side of the HCSG. We used Nek5000, an open source, high-order spectral element computational fluid dynamics code developed in Argonne National Laboratory. Nek5000 accepts only hexahedral mesh, which makes the meshing process for the complicated HCSG geometry very challenging. A tetrahedral-to-hexahedral meshing strategy was applied to bypass the geometric complexities. In this study large eddy simulation (LES) was performed at the Reynolds number of 9000 based on the inlet velocity and the tube diameter. The employed subgrid-scale model for LES relies on explicit filtering. First- and second-order statistics were compared to available experimental data. We report overall velocity and turbulent kinetic energy showed good agreement with particle image velocimetry data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Simulation of Boiling Two-Phase Flow in a Helical Coil Steam Generator Using the Spectral Element Code Nek-2P

To enable the design of a light water small modular reactor, the boiling flow inside a helical coil steam generator has been simulated with the two-fluid model in Nek-2P. Nek-2P is the multiphase branch of the spectral element code Nek5000. Details of the implementation of the two-fluid model and the included closure models are discussed. The presented closure models include interactions for momentum, heat, and mass transfer between phases. Models for the drag, lift, and turbulent dispersion forces are included. The complete model is fully consistent in the limits of both phasic volume fractions approaching zero and is able to simulate flows of dispersed vapor, continuous liquid, dispersed liquid, continuous vapor, or any combination thereof. The closure models and their implementation in Nek-2P have been validated by comparing to experimental data for a boiling flow, demonstrating excellent agreement. Results from the simulation of the helical coil indicate strong phasic separation driven by the effects of buoyancy and inertia. Finally, significant differences were observed in the results compared to simulations performed using Star-CCM+, although these differences were somewhat expected.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Severe Accident Phenomena: A Comparison Among the NuScale SMR, Other Advanced LWR Designs, and Operating LWRs

During a severe accident in a nuclear reactor, there are a number of phenomenological events which can present a challenge to containment integrity. These include the generation and combustion of hydrogen, energetic fuel-coolant interactions, thermal attack of fission product barriers, core-concrete interactions, direct containment heating, and gradual overpressurization. The advanced design of the NuScale small modular reactor has resulted in the reduced likelihood and severity of severe accident challenges to containment. This paper discusses the features of the NuScale design that reduce the likelihood of occurrence of these severe accident phenomena, and discusses the ability of containment to survive in the unlikely event that they do occur. Here, the impact of severe accident phenomena for the NuScale design is compared and contrasted against other advanced light water reactors, such as AP1000 and ESBWR, as well as the existing fleet, using information from publicly available documents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Impact of Radial Reflector Fidelity on Neutronics and Vessel Fluence Simulations

The Consortium for Advanced Simulation of Light Water Reactors is developing the Virtual Environment for Reactor Applications (VERA), and the MPACT code, which is the primary deterministic neutron transport solver in VERA, provides sub-pin level flux and power distributions as part of full-scale cycle depletion and analysis. In such calculations, an important aspect is the radial reflector treatment. To improve the fidelity of the radial reflector treatment, MPACT was extended to approximate the modeling of the reactor’s structural components such as the core shroud, barrel, neutron pads, and vessel. This work explores several modeling configurations with varying levels of fidelity and computational burden and assesses the importance of modeling fidelity on the eigenvalue and pin power distribution. Two two-dimensional (2-D) problems were analyzed to assess the impact on eigenvalue and pin power distributions with low-fidelity, coarse square cell reflector representations: (1) a Watts Bar Nuclear Plant Unit 1 (WBN1) quarter-core slice with depletion and (2) an AP1000 quarter-core slice. In this work, the analyses showed that the effect on eigenvalue is fairly small, but the effect on pin power is more pronounced, especially locally in the assemblies closest to the periphery, where the maximum pin power difference is nearly 3.5% in the AP1000 case. Two additional 2-D problems were used to assess the comparison between the low-fidelity coarse square cell treatment and a high-fidelity geometric representation that uses subpin material specification: (1) the same WBN1 quarter-core slice and (2) a representative model of the NuScale small modular reactor (SMR), which features a solid reflector design with moderator holes. These results demonstrate that even a coarse, low-fidelity representation adequately captures the necessary simulation characteristics. Last, these capabilities were applied to the 2-D WBN1 quarter-core depletion to assess the impact on vessel fluence using VeraShift. From adjoint calculations, pins along the periphery were observed to be of highest importance for fluence calculation, so the impact of the reflector representation in MPACT could theoretically substantially affect the predicted result. However, it was observed that the change in pin powers along the periphery minimally impacts the maximum vessel fluence with a difference within the statistical uncertainty but provides terrific insight on the sensitivity of the peripheral pins.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors

Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.

TRISO↗
DOE OSTI2012

Calibration Monitoring for Sensor Calibration Interval Extension: Gaps in the Current Science Base

Currently in the United States, periodic sensor recalibration is required for all safety-related sensors, typically occurring at every refueling outage, and it has emerged as a critical path item for shortening outage duration in some plants. International application of calibration monitoring has shown that sensors may operate for longer periods within calibration tolerances. This issue is expected to also be important as the United States looks to the next generation of reactor designs (such as small modular reactors and advanced concepts), given the anticipated longer refueling cycles, proposed advanced sensors, and digital instrumentation and control systems. Online monitoring (OLM) can be employed to identify those sensors that require calibration, allowing for calibration of only those sensors that need it. The U.S. Nuclear Regulatory Commission (NRC) accepted the general concept of OLM for sensor calibration monitoring in 2000, but no U.S. plants have been granted the necessary license amendment to apply it. This paper summarizes a recent state-of-the-art assessment of online calibration monitoring in the nuclear power industry, including sensors, calibration practice, and OLM algorithms. This assessment identifies key research needs and gaps that prohibit integration of the NRC-approved online calibration monitoring system in the U.S. nuclear industry. Several technical needs were identified, including an understanding of the impacts of sensor degradation on measurements for both conventional and emerging sensors; the quantification of uncertainty in online calibration assessment; determination of calibration acceptance criteria and quantification of the effect of acceptance criteria variability on system performance; and assessment of the feasibility of using virtual sensor estimates to replace identified faulty sensors in order to extend operation to the next convenient maintenance opportunity.

Calibration Interval Extension↗
DOE OSTI2013

Advanced Instrumentation for Extreme Environments

The Department of Energy (DOE) Office of Nuclear Energy (NE) is pursuing embedded instrumentation and controls (I&C) technology for next generation nuclear power generation applications. Embedded systems encompass a wide range of configurations and technologies; we define embedding in this instance as the integration of the sensors and the control system design into the component design using a systems engineering process. Embedded I&C systems are often an essential part of developing new capabilities, improving reliability, enhancing performance, and reducing operational costs. The new intrinsically safe, more efficient, and cost effective reactor technologies (Next Generation Nuclear Plant and Small Modular Reactors) require the development and application of new I&C technologies. These new designs raise extreme environmental challenges such as high temperatures (over 700 C) and material compatibility (e.g., molten salts). The desired reliability and functionality requires measurements in these extreme conditions including high radiation environments which were not previously monitored in real time. The DOE/NE Nuclear Energy Enabling Technologies (NEET) program currently has several projects investigating I&C technologies necessary to make these reactor designs realizable. The project described in this paper has the specific goal of investigating embedded I&C with the following objectives: 1.Explore and quantify the potential gains from embedded I&C improved reliability, increased performance, and reduced cost 2.Identify practical control, sensing, and measurement techniques for the extreme environments found in high-temperature reactors 3.Design and fabricate a functional prototype high-temperature cooling pump for molten salts represents target demonstration of improved performance, reliability, and widespread usage There are many engineering challenges in the design of a high-temperature liquid salt cooling pump. The pump and motor are in direct contact with molten fluoride salt at 700 C (1,292 F) as part of a reactor cooling loop. The motor-pump combination during normal operation would be red-hot (Figure 1). This environment challenges every facet of the design including seals, wiring, magnetic materials, and sensors. In this paper, we discuss the challenges of sensor design in extreme environments and specifically the sensor design for a high-temperature fluoride salt coolant pump. This pump will be used as a test-bed for embedded I&C development and validation in extreme environments.

Extreme Environments↗