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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Implementation of an Experimental Setup Based on the Three-Omega Method for Thermal Conductivity Measurements of Molten Salts

Molten salt reactors (MSRs) are one of the most promising next-generation nuclear reactor designs which use molten salts as the fuel and coolant. To design, model, predict, license, and operate MSRs, the behavior of the fuel and coolant must be understood as a function of temperature and irradiation, i.e., burnup. Of particular importance and complexity is the reliable and reproducible measurement of thermal conductivity of molten salts at high temperatures (> 500 °C). Thermal conductivity provides information on the rate at which heat passes through a specified material and is dependent on chemical composition. Currently, in the national laboratory complex, several methods for measuring thermal conductivity on molten salt systems are under development, however, to date, consistent data have not been produced. Typically, thermal conductivity measurements of high-temperature melts have been conducted by using the Transient Hot Wire (THW) method where thermal conductivity is determined by the change in temperature over time at a certain distance from a heating wire. Another time-dependent technique is the laser flash method, commonly used for solid samples. Laser flash requires prior measurements of density and heat capacity to calculate the thermal conductivity of the sample, introducing additional measurement errors. Steady state and time-dependent measurements such as THW and laser flash have errors from convection and ambient temperature fluctuations (heat radiation). In order to produce high fidelity thermal conductivity data, a measurement technique needs to be developed that will reduce the effect of convection at high temperatures by measuring in a very fast time window. In addition, to measure on actinide and irradiated salts the sample size should be small. Therefore, this project aims to develop a frequency-domain technique based on the three-omega (3?) method, previously proven to work on solids, nanofluids and low temperature solar salts (25 to 250 °C). If the 3?-device can be designed with materials that can withstand high temperatures (>500 °C) and can be configured to reduce errors due to convection, then it would prove to be a valuable technique for ensuring thermal conductivity on molten salt systems needed for MSR development. For this proposal, the 3? instrument will be designed, assembled, and tested on non-radioactive salts before being used on actinide bearing salts. In addition, a thermal model will be developed to determine the thermal conductivity of the sample from the measured thermal oscillation of the sensor. Because 3? is a direct method that requires a small volume of sample and works in a very fast time window, it will aid in suppressing the interference due to convection which is ideal when working with high temperature liquid samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

FY20 Report for Instrumentation Development for the Transient Testing Program

This yearly report provides a summary of accomplishments carried out under funding from the Department of Energy (DOE) Nuclear Technology Research and Development (NTRD) program for in-pile instrumentation supporting the transient testing program in FY20. These activities were performed in support of cross-cutting transient testing experiment objective. In this year, specific instrument R&D has been a lesser priority to years past, with complementary instrumentation R&D being ramped up under the DOE Advanced Sensors and Instrumentation Program. The primary focus of the work in FY20 pertain to defining a process and system, developed over the past several years, for developing, qualifying, and deploying instrumentation in in-pile experiments. The purpose of this report is to provide a summary of these outcomes and be a reference for researchers that are interested in developing in-pile instrumentation, with a particular emphasis on the Transient Reactor Test (TREAT) facility for deployment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Assessment of Condition Monitoring Methods and Technologies for Inservice Inspection and Testing of Nuclear Power Plant Components

This report was prepared for the U.S. Nuclear Regulatory Commission (NRC) to explore the application of advanced technologies toward meeting the current and future regulatory requirements for maintenance and condition monitoring of structures, systems, and components. The advanced technologies considered in this work are advanced sensors and instrumentation, data analytics, machine learning and artificial intelligence (ML/AI), physics-based models, and digital twins (DT). The interest in the application of advanced technologies for condition monitoring in nuclear power plants continues to grow, and current and future licensees are expected to implement advanced technologies as part of their inservice inspection (ISI) and inservice testing (IST) programs. This report delineates the outcomes of an exploratory investigation into the implementation of advanced condition monitoring technologies to address ISI and IST requirements. A thorough review was conducted of the existing regulatory requirements for ISI and IST, along with an analysis of associated industry practices. Additionally, a state-of-the-art assessment was performed on advanced condition monitoring technologies frequently employed in non-nuclear sectors. This research incorporated two nuclear-specific case studies to illustrate the application of these technologies within the current nuclear fleet. The report provides an exhaustive discussion on the technical challenges, considerations, and opportunities associated with the deployment of advanced condition monitoring technologies. The following are key considerations in the application of advanced technologies for the ISI and IST of nuclear power plant components: • Developing adequate verification and validation procedures to confirm the functional and non-functional requirements, • Developing technical capabilities to conduct real-time asset condition monitoring, • Establishing guidance and protocol for modeling and simulation tools to continuously meet regulatory requirements, • Addressing trustworthiness, explainability, and interpretability of ML/AI methods, • Evaluating maintenance activities to maintain an adequate safety margin and avoid undesirable conditions, • Establishing cybersecure condition monitoring programs associated with a computer-based software system, and • Establishing standardized evaluation metrics for advanced condition monitoring programs. Interest in the use of advanced technologies for condition monitoring in ISI and IST programs continues to grow, and the technology is expected to experience rapid and wide industry adoption in the near future. Adoption of advanced technologies for condition monitoring could have novel and unique impacts on regulatory activities associated with ISI and IST programs. The NRC is continuing to explore the regulatory aspects of advanced technologies as part of ISI and IST programs by pursuing additional research in this technical area.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Improving nuclear power plant safety and operating margins through linear quadratic estimation - 290

Redundant measurement, trip, and actuation channels are required for reactor trips and engineered safety systems in nuclear power plants. There are three or four independent channels in existing plants, depending on the system. To prevent failures, consistent signals from any two channels together can trip the reactor or actuate safety systems. Measurements associated with safety trips and actuations include uncertainties related to the instruments making the measurements and to process conditions. Margins that include measurement uncertainties are set for each trip and actuation channel as though it were the only one taking action, thus preserving independence. There are sensors, however, that provide partially redundant information that could be used to decrease the uncertainty associated with measurements of plant conditions, while maintaining the three-channel redundancy important to plant safety. Reducing these uncertainties would both improve plant safety by providing more accurate knowledge of plant conditions and improve the operating margins that must take such uncertainties into account. This paper demonstrates how information that is already available to plant instrumentation and control systems and is already redundant can be used through software routines to reduce the error that must be taken into account in reactor trip and safety system actuation setpoints. Issues that would have to be addressed related to standards, licensing, and implementation are also discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

The effects of main-ion dilution on turbulence in low q 95 C-Mod ohmic plasmas, and comparisons with nonlinear GYRO

Recent experiments on C-mod seeding nitrogen into ohmic plasmas with q 95 = 3.4 found that the seeding greatly reduced long-wavelength (ITG-scale) turbulence. The long-wavelength turbulence that was reduced by the nitrogen seeding was localized to the region of r/a ≈ 0.85, where the turbulence is well above marginal stability (as evidenced by Q i /Q GB $\gg$ 1). The nonlinear gyrokinetic code GYRO was used to simulate the expected turbulence in these plasmas, and the simulated turbulent density fluctuations and turbulent energy fluxes quantitatively agreed with the experimental measurements both before and after the nitrogen seeding. Unexpectedly, the intrinsic rotation of the plasma was also found to be affected by the nitrogen seeding, in a manner apparently unrelated to a change in the electron-ion collisionality that was proposed by other experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Guidance for Developing Digital Twins for Online Condition Monitoring of Nuclear Power Plant Components

Online condition monitoring is an area of active research that may enable optimized scheduling, maintenance, and safety of nuclear power plant components, reducing unnecessary derates while simultaneously improving operational capacity. Digital twins (DTs) are one avenue to conduct online condition monitoring and are currently being explored by national laboratories and universities alike. DTs for online condition monitoring are, in essence, state concurrent models that emulate a physical process which predicts a parameter and compares it against a measured value. The promise of DT is that they may provide additional insights by combining and interpreting various sources of information and may be used for preventative maintenance scheduling optimization or early fault detection. DTs for condition monitoring are projected to be valuable for meeting requirements under 10 CFR 50.55a and 10 CFR 50.65. However, DT technologies are still under significant development and the process for developing a DT for condition monitoring has not been formalized. Therefore, in this work, we present an initial framework for developing a DT, discuss and review the various challenges and considerations for DT deployment, and identify the opportunities that a DT can improve. Here, the presented framework is intended to help developers formulate a strategy when approaching DT development for condition monitoring. A DT use case for a reactor coolant pump is presented to demonstrate the proposed framework.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Assessment of thermodynamic stability of sapphire in eutectic molten chloride environment

The continued development of molten salt reactors requires highly efficient and stable instruments that can efficiently track monitor conditions of the molten salt during long-term application operation in both the fuel and coolant/heat transfer fluid loops. Sapphire (Al 2 O 3 ) fibers have shown tremendous potential by enabling radiation resistance and a broader operational range of temperature . Thermodynamic modeling using the ThermoCalc and the SGTE Molten Salts (SALT1) and Pure Substances (Pure5) database is done performed to understand the compatibility of Al 2 O 3 fibers with NaCl-MgCl 2 eutectic molten salt in the temperature range of 600 to 1000°C. The thermodynamic calculations show that sapphire fibers are not expected to be compatible with long term exposure to molten chloride salts within this temperature range. Furthermore, various pure metallic elements were evaluated as potential cladding materials for part of the Al 2 O 3 fibers. Based on the thermodynamic analysis, molybdenum (Mo) and nickel (Ni) could be effective cladding materials to enhance the stability of Al 2 O 3 in NaCl and MgCl 2 chloride salt molten bath in the desired temperature range. The presence of a metallic cladding leads to the suppression of aluminum dissolution from Al 2 O 3 , thus improving its stability in chloride salts. Additionally, it is also shown that Al 2 O 3 remains stable up to 2300 K in the presence of pre-existing Al 2 MgO 4 and Al 2 NiO 4 in the eutectic molten chloride bath environment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Analyzing Mechanical integrity of Microwave resonant Cavity flowmeter

The microwave resonant cavity flow sensor has been proposed to address the difficulties of measuring high-temperature molten salt flow in reactor vessels. The flow sensor consists of a metallic cylindrical cavity coupled to a microwave waveguide. The cylindrical cavity has one of the flat surfaces thin enough to undergo microscopic deflection due to dynamic pressure of the exerted by the flowing fluid. Deflection of the membrane causes a shift in the microwave resonant frequency. Fluid velocity can be estimated from the calibration curve. The objective of this report is to investigate mechanical integrity of the flowmeter’s membrane. The membrane is a thin stainless-steel circular plate with a clamped edge, subjected to uniform load. By calculating the stress on the plate due to deflection and, comparing the stress to the material ultimate tensile strength and yield strength, it can be estimated if the plate will fail. If the maximum stress is greater than the yield strength, the plate will undergo plastic deformation, and the sensor will be disabled. A stress greater than the ultimate tensile strength will cause the plate to fracture. The stress on the plate was calculated with an analytic closed form solution model, and with COMSOL Structural Mechanics Module which does not involve any approximations. Stresses in both the analytic model and COMSOL numerical model were calculated for stainless steel 316 thin circular plates under room temperature conditions, where the fluid is water. The diameter of the circular plate is approximately one inch, and thickness is 10mil. Fluid flow velocity is in the range from 0.5m/s to 1m/s. The results from the two models are similar. Both the analytic model and COMSOL model showed that maximum stresses on the plate, which are at the radial boundary of the plate, are three orders of magnitude smaller than the yield strength and ultimate tensile strength. This indicates that the sensor is at a low risk of mechanical failure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2007

Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

Bragg-Sitton, S. M.↗
NASA NTRS2011

Fission Surface Power Systems (FSPS) Project Final Report for the Exploration Technology Development Program (ETDP): Fission Surface Power, Transition Face to Face

The Fission Surface Power Systems Project became part of the ETDP on October 1, 2008. Its goal was to demonstrate fission power system technology readiness in an operationally relevant environment, while providing data on fission system characteristics pertinent to the use of a fission power system on planetary surfaces. During fiscal years 08 to 10, the FSPS project activities were dominated by hardware demonstrations of component technologies, to verify their readiness for inclusion in the fission surface power system. These Pathfinders demonstrated multi-kWe Stirling power conversion operating with heat delivered via liquid metal NaK, composite Ti/H2O heat pipe radiator panel operations at 400 K input water temperature, no-moving-part electromagnetic liquid metal pump operation with NaK at flight-like temperatures, and subscale performance of an electric resistance reactor simulator capable of reproducing characteristics of a nuclear reactor for the purpose of system-level testing, and a longer list of component technologies included in the attached report. Based on the successful conclusion of Pathfinder testing, work began in 2010 on design and development of the Technology Demonstration Unit (TDU), a full-scale 1/4 power system-level non-nuclear assembly of a reactor simulator, power conversion, heat rejection, instrumentation and controls, and power management and distribution. The TDU will be developed and fabricated during fiscal years 11 and 12, culminating in initial testing with water cooling replacing the heat rejection system in 2012, and complete testing of the full TDU by the end of 2014. Due to its importance for Mars exploration, potential applicability to missions preceding Mars missions, and readiness for an early system-level demonstration, the Enabling Technology Development and Demonstration program is currently planning to continue the project as the Fission Power Systems project, including emphasis on the TDU completion and testing.

Palac, Donald T.↗
NASA NTRS2012

RESOLVE Mission Architecture for Lunar Resource Prospecting and Utilization

Design Reference Mission (DRM) evaluations were performed for The Regolith & Environment Science, and Oxygen & Lunar Volatile Extraction (RESOLVE) project to determine future flight mission feasibility and understand potential mission environment impacts on hardware requirements, science/resource assessment objectives, and mission planning. DRM version 2.2 (DRM 2.2) is presented for a notional flight of the RESOLVE payload for lunar resource ground truth and utilization (Figure 1) [1]. The rover/payload deploys on a 10 day surface mission to the Cabeus crater near the lunar south pole in May of 2016. A drill, four primary science instruments, and a high temperature chemical reactor will acquire and characterize water and other volatiles in the near sub-surface, and perform demonstrations of In-Situ Re-source Utilization (ISRU). DRM 2.2 is a reference point, and will be periodically revised to accommodate and incorporate changes to project approach or implementation, and to explore mission alternatives such as landing site or opportunity.

George, J. A.↗
DOE OSTI2017

Upcoming irradiation of ultrasonic and fiber optic sensors in the MITR - 361

Fiber optic and ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. These technologies, which have performed successfully in out-of-pile tests, have not been well qualified in a test reactor environment. The uncertainty of successful operation of instruments at this level of development has typically been a significant barrier to in-core deployment. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of these technologies to provide enhanced sensors for irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. The National Scientific User Facility (NSUF) is sponsoring a project to evaluate the performance of promising temperature sensors based on ultrasound and on fiber optics during an upcoming irradiation test in the Massachusetts Institute of Technology Research Reactor (MITR). Several types of thin-wire metallic waveguide ultrasonic thermometers will be tested along with fiber optic temperature sensors using regenerated Fiber-Bragg Grating technology. Additionally, samples of several new optical materials which are potentially much more radiation tolerant than commonly used fiber materials, such as silica, will be included and analyzed post-irradiation. The irradiation is a collaborative effort between researchers from Idaho National Laboratory, the French Alternative Energies and Atomic Energy Commission, the University of Pittsburgh, the Massachusetts Institute of Technology, and AFO Research, Inc. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Status Report for NEET In-Pile Sensor Irradiation Capabilities at TREAT

The recently-established in-pile sensor development project under the Nuclear Energy Enabling Technology (NEET) program is focused on the development of measurement technologies to enable acquisition of data focused on the performance, properties, and boundary conditions of nuclear fuels/materials under irradiation in nuclear reactors. These sensors can be used in steady state irradiation tests to measure parameters of interest under conditions representing normal nuclear plant operation, or in transient irradiation tests to determine specimen performance under postulated unplanned conditions. In either case, development of suitable sensor technologies must include instrument irradiations to help guide sensor material/design selection, verify instrument performance, and demonstrate deployment viability. The Transient Reactor Test facility (TREAT) while just a part of the overall capabilities portfolio needed to enable this sensor development mission, has unique capabilities and characteristics that make it well poised for a number of instrument irradiation data objectives including sensor survivability screening, response characterization, qualification tests, and in-situ properties measurements. TREAT’s brilliantly-basic design, when married with appropriate irradiation hardware and analytic capabilities, fills a crucial capability gap between out-of-pile testing and irradiation in steady state high power research reactors. Leveraging a recently developed approach to irradiation testing in TREAT, the present work demonstrates that cost-effective and scientifically-impactful instrument irradiation capabilities for TREAT are a well-placed capability in the NEET in-pile sensor development project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Autonomous continuous flow reactor synthesis for scalable atom-precision

With new instrumentation design, robotics, and in-operando hyphenated analytical tool automation, the intelligent discovery of synthesis pathways is becoming feasible. It can potentially bridge the gap for the scale-up of new materials. In this article, we review current progress and describe a new system that uses an autonomous continuous flow chemistry framework to translate high-quality lead molecules and materials to quantities that can meet scalability demands. At the core is a continuous flow synthesis platform that can design its viable synthesis pathway to a particular molecule or material and then autonomously carry it out. This is realized by integrating: (1) A workflow/architecture for multimode chemical/materials characterization in-line. The in-line characterization modes are NMR, ESR, IR, Raman, UV-Vis, GC-MS, and HPLC, along with ex-situ modes for X-Ray and neutron scattering; (2) Integration for feedback/analysis/data storage of the control variables; (3) A core software stack that includes deep learning and reinforcement learning alongside quantum chemistry and molecular dynamics; (4) On-demand compute architectures that parse calculations to compute resources needed which include light-weight edge, mid-level edge (NVIDA DGX-2), and high-performance computing. We demonstrate preliminary results on how this autonomous reactor system can enhance our ability to deliver deuterated materials, copolymers, and site-substituted molecules.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of Cambridge have been investigated. The rationale for the superior performance of the type N using a customized sheath developed at the University of Cambridge is explained in comparison with the behavior of conventional type N Inconel 600 sheathed thermocouples. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Liquid sodium testing of in-house phased array EMAT transducer for L-wave applications

This paper describes the development of an in-house phased array EMAT transducer for longitudinal wave inspection in liquid sodium. The work presented herein is part of an undergoing project aimed at improving in-service inspection techniques for the ASTRID reactor project. The design process of the phased array EMAT probe is briefly explained and followed by a review of experimental test results. We first present test results obtained in the laboratory while the last part of the paper describes the liquid sodium testing and the produced ultrasound images. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Identifying subassemblies by ultrasound to prevent fuel handling error in sodium fast reactors: First test performed in water

Identifying subassemblies by ultrasound is a method that is being considered to prevent handling errors in sodium fast reactors. It is based on the reading of a code (aligned notches) engraved on the subassembly head by an emitting/receiving ultrasonic sensor. This reading is carried out in sodium with high temperature transducers. The resulting one-dimensional C-scan can be likened to a binary code expressing the subassembly type and number. The first test performed in water investigated two parameters: width and depth of the notches. The code remained legible for notches as thin as 1.6 mm wide. The impact of the depth seems minor in the range under investigation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗