SEARCH · Engineering Papers

Results for “THERMOCOUPLES”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 793 records · Page 44

All sources NASA NTRS DOE OSTI

DOE OSTI2016

Cryostat system for investigation on new neutron moderator materials at reactor TRIGA PUSPATI

A simple continuous flow (SCF) cryostat was designed to investigate the neutron moderation of alumina in high temperature co-ceramic (HTCC) and polymeric materials such as Teflon under TRIGA neutron environment using a reflected neutron beam from a monochromator. Cooling of the cryostat will be carried out using liquid nitrogen. The cryostat will be built with an aluminum holder for moderator within stainless steel cylinder pipe. A copper thermocouple will be used as the temperature sensor to monitor the moderator temperature inside the cryostat holder. Initial measurements of neutron spectrum after neutron passing through the moderating materials have been carried out using a neutron spectrometer.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2016

Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N 2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that “solves” the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional–integral–derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Lower hybrid wave edge power loss quantification on the Alcator C-Mod tokamak

For the first time, the power deposition of lower hybrid RF waves into the edge plasma of a diverted tokamak has been systematically quantified. Edge deposition represents a parasitic loss of power that can greatly impact the use and efficiency of Lower Hybrid Current Drive (LHCD) at reactor-relevant densities. Through the use of a unique set of fast time resolution edge diagnostics, including innovative fast-thermocouples, an extensive set of Langmuir probes, and a Lyα ionization camera, the toroidal, poloidal, and radial structure of the power deposition has been simultaneously determined. Power modulation was used to directly isolate the RF effects due to the prompt (t<τ E ) response of the scrape-off-layer (SOL) plasma to Lower Hybrid Radiofrequency (LHRF) power. LHRF power was found to absorb more strongly in the edge at higher densities. It is found that a majority of this edge-deposited power is promptly conducted to the divertor. This correlates with the loss of current drive efficiency at high density previously observed on Alcator C-Mod, and displaying characteristics that contrast with the local RF edge absorption seen on other tokamaks. Measurements of ionization in the active divertor show dramatic changes due to LHRF power, implying that divertor region can be a key for the LHRF edge power deposition physics. These observations support the existence of a loss mechanism near the edge for LHRF at high density (n e >1.0×10 20 (m -3 )). Results will be shown addressing the distribution of power within the SOL, including the toroidal symmetry and radial distribution. Finally, these characteristics are important for deducing the cause of the reduced LHCD efficiency at high density and motivate the tailoring of wave propagation to minimize SOL interaction, for example, through the use of high-field-side launch.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2018

A divertor scraper observation system for the Wendelstein 7-X stellarator

Two graphite divertor elements called scrapers have been installed on the Wendelstein 7-X stellarator in the throat of the magnetic island divertor. To diagnose one, we have designed, built, calibrated, and installed a new infrared/visible imaging endoscope system to enable detailed observations of the plasma interactions and heat loads at one of the scrapers and the nearby divertor surfaces. The new system uses a shuttered pinhole-protected pair of 90° off-axis 228 mm focal length aluminum parabolic mirrors, and two flat turning metal mirrors, to send light to a sapphire vacuum window 1.6 meters away, beyond which we have co-located telephoto lens-based infrared and visible cameras. The back-to-back off-axis parabolas serve to cancel out most aberrations, enabling the use of off-the-shelf commercial optics outside of the vessel. For the infrared, we use a 3-5 μm 1-megapixel FLIR SC8303HD camera and for the visible, a 5-megapixel CMOS PCO 5.5 edge camera. A short 1-m quartz pickoff fiber is used to send 200-1100 nm light to a compact spectrometer, also located in the same iron shield box as the cameras. The camera field of view covers the 700 mm length of the scraper, and includes locations monitored by thermocouples and Langmuir probes embedded in some of the scraper tiles. Predicted and actual optical test performances of the overall system are then compared.

47 OTHER INSTRUMENTATION↗

DOE OSTI2020

A simple vacuum suitcase for enabling plasma facing component characterization in fusion devices

We have demonstrated a vacuum suitcase to transport samples in vacuo to a surface analysis station for characterization of tokamak plasma facing components (PFCs). This technique enables surface analysis at powerful, dedicated stations that are not encumbered by design constraints imposed on them by a tokamak. The vacuum suitcase is an alternative solution to characterizing PFCs using diagnostics that are designed and built around a tokamak. The vacuum suitcase, called the Sample Exposure Probe (SEP), features mobile ultra-high vacuum pumping. Active pumping under high vacuum enables sample transfer between the Lithium Tokamak eXperiment-β (LTX-β) and a high resolution X-ray Photoelectron Spectroscopy (XPS) system that is situated close by. A thermocouple inserted in the back of the sample head measures heat flux from the plasma during exposure, and together with a button heater, allows the sample to match the LTX-β PFCs in high temperature operations. As vacuum conditions are better during transfer and analysis than in the tokamak, less contamination is introduced to the samples. XPS scans on a dedicated analysis station enable peak identification due to higher resolution and signal to noise ratio. A similar probe could be implemented for other fusion devices. As a result, the SEP is the first vacuum suitcase implementation for fusion applications that incorporates active pumping.

47 OTHER INSTRUMENTATION↗

DOE OSTI2020

Radiation furnace for synchrotron dark-field x-ray microscopy experiments

We present a multi-purpose radiation furnace designed for x-ray experiments at synchrotrons. The furnace is optimized specifically for dark-field x-ray microscopy (DFXM) of crystalline materials at beamline ID06 of the European Synchrotron Radiation Facility. The furnace can reach temperatures above 1200 °C with a thermal stability better than 10 °C, with heating and cooling rates up to 30 K/s. The non-contact heating design enables samples to be heated either in air or in a controlled atmosphere contained within a capillary tube. The temperature was calibrated via the thermal expansion of an α-iron grain. Temperature profiles in the y and z axes were measured by scanning a thermocouple through the focal spot of the radiation furnace. In the current configuration of the beamline, this furnace can be used for DFXM, near-field x-ray topography, bright-field x-ray nanotomography, high-resolution reciprocal space mapping, and limited powder diffraction experiments. As a first application, we present a DFXM case study on isothermal heating of a commercially pure single crystal of aluminum.

Yildirim, C. (ORCID:0000000249690685)↗

DOE OSTI2022

Reexamination of hydrodynamic phonon transport in thin graphite

The recent hydrodynamic phonon transport theory for graphitic materials has been supported by the measurements of the second sound at temperatures up to about 100 K. When boundary scattering becomes comparable to momentum-conserving normal phonon scattering processes that are responsible for phonon hydrodynamics, Poiseuille phonon flow phenomena can emerge to give rise to unique size-dependent thermal conductivity in thin graphite. Here, we examine the thickness range for the Poiseuille phonon flow to become observable in thin graphite with the use of both deviational Monte Carlo simulation of the Peierls-Boltzmann transport equation and four-probe thermal transport measurements. As the basal-plane thermal conductivity calculated by prior first-principles theories saturates to the graphite value when the thickness is increased to five graphene layers, the phonon dispersion of graphite is used in the current calculations of thin graphite of micrometer thickness and a 23-layer thick ultrathin graphite (UTG) sample. The calculations show that diffuse surface scattering by surface defects can lead to Poiseuille phonon flow at 50 K in thin graphite with the thickness close to several micrometers but not in the 65 μm thin graphite and 23-layer UTG, where phonon scattering by the top and bottom surfaces become, respectively, much less and more frequent than the normal processes. In addition, the calculation results with the bulk graphite dispersion and diffuse surface scattering show decreased basal-plane thermal conductivity with decreasing thickness, opposite to recent thermocouple measurements of thin graphite samples. In comparison, the calculation results reveal that partially diffuse surface defect scattering can yield the four-probe measurement results of UTG samples, which are prepared here with an improved process to minimize surface contamination.

36 MATERIALS SCIENCE↗

DOE OSTI2017

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s ability to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Development of Advanced Instrumentation for Transient Testing

A fuel safety research program centered on in-pile transient testing experiments is being developed to support assessment and qualification of advanced nuclear fuel systems using the recently restarted Transient Reactor Test (TREAT) facility at the Idaho National Laboratory (INL). While resumption of transient testing at TREAT is crucial to enable these programs, full recovery and cutting-edge transient testing capability also requires a well-coordinated and innovative instrumentation development and qualification program to support near-term and future objectives. This paper summarizes the experimental approach of transient testing to focus on measuring the response of nuclear fuel to off-normal (or power-cooling mismatch) conditions for modern and advanced reactor environments requiring capabilities extending over wide measurement and environment conditions. It also highlights unique attributes of transient testing of importance to in-pile instruments including relatively low total neutron fluence, high gamma heating, and the need for well-defined and possibly short time response. Historical approaches to instrumentation for transient testing are also reviewed to provide context to the modern instrument strategy. The instrumentation needs of a modern transient testing are detailed while summarizing several ongoing R&D activities are supporting the development of state-of-the-art and ‘advanced’ measurement technologies that will provide baseline capability for Light Water Reactor (LWR) and Sodium-Cooled Fast Reactor (SFR) experiment objectives, while extending to other advanced reactor needs and advanced sensing technology opportunities. Examples of specific sensors planned for near-term deployment with ongoing development includes prompt-response self-powered neutron detectors (SPND), miniature fission chambers, optical-fiber-coupled infrared pyrometer, cladding surface thermocouples, electrical-impedance based boiling detector, and Linear Variable Differential Transformer (LVDT) based sensors for fuel elongation and pressure measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Nonnuclear Experimental Capabilities to Support Design, Development, and Demonstration of Microreactors

This work provides a summary of selected experimental capabilities being developed to support nonnuclear testing and demonstration of technology in support of microreactors under the U.S. Department of Energy’s (DOE’s) Microreactor Program. Major capabilities include the Single Primary Heat Extraction and Removal Emulator (SPHERE) and the Microreactor Agile Non-nuclear Experimental Test Bed (MAGNET). The SPHERE facility allows for controlled testing of the steady-state and transient heat rejection capabilities of a single heat pipe using electrical heaters that simulate nuclear heating. The facility is capable of monitoring axial temperature profiles along the heat pipe and surrounding test articles during startup, steady-state operation, and transients. Instrumentation includes noncontact infrared thermal imaging, surface thermocouples, spatially distributed fiber optic temperature and strain sensors, electrical power meters, and a water-cooled, gas-gap calorimeter for quantifying heat rejection from the heat pipe. The facility can be operated under both vacuum and inert-gas conditions. The MAGNET facility is a large-scale, 250-kW electrically heated microreactor test bed to enable nonnuclear experimental evaluation of a variety of microreactor concepts. It can be supplied to electrically heat a scaled section of a microreactor and further test the capabilities of heat rejection systems. The initial MAGNET experiments will support technology maturation and reduce uncertainty and risk associated with the design, operation, and deployment of monolithic heat pipe–based reactors. However, this test bed can broadly be applied to multiple microreactor concepts to evaluate a wide range of thermal-hydraulic and structural phenomena such as interface coupling with power conversion units and other collocated systems. MAGNET can evaluate integral thermomechanical effects during electrical heating of an array of heat pipes in a larger test article. Examples of initial testing will include thermal stresses in the monolith and the impact of debonding of a heat pipe from the core block and how that failure could impact surrounding heat pipes, i.e., understanding the potential for cascading failure. This work also discusses some modeling capabilities that can support experiment design, analysis, and interpretation, including the heat pipe code Sockeye and a comparison of thermal-structural simulations performed using ABAQUS and STAR-CCM+.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Sensitivity Analysis of Transient Critical Heat Flux by RIA Under High-Pressure Flow Boiling Conditions in TRTL

A reactivity-initiated accident (RIA) is a design-basis accident under which critical heat flux (CHF) is likely to be exceeded. The operational margin for RIAs is currently determined using steady-state CHF lookup tables, which provide conservative estimates relative to transient CHF phenomena. The Transient Reactor Test Loop (TRTL) facility at Oregon State University is capable of performing out-of-pile rapid heating experiments representative of a RIA at conditions representative of a pressurized water reactor (PWR). Here, to further our understanding of and ability to predict transient CHF under PWR conditions, we performed a sensitivity analysis on a RELAP5-3D model of the TRTL facility coupled to the RAVEN code framework to define a proposed experimental test matrix to be performed at the TRTL facility. We then implemented a flow boiling CHF correlation into RELAP5-3D and performed a secondary sensitivity analysis inspecting the impact of the built-in RELAP5-3D CHF and heat transfer multipliers on both the prediction of CHF and key safety parameters, such as peak cladding temperature and heat flux. The results show that the multiplier with the highest influence toward the prediction of CHF occurrence and the safety parameters is the transient CHF multiplier. Operational performance envelopes have been developed for each of the test matrix cases and will be used for validation once the experiments are performed. The TRTL facility is currently performing shakedown testing to verify system performance prior to proceeding with the experimental campaign. Restart testing results include pump curve restart testing, pressure tests, and heater rod thermocouple transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Application of a Physics-Informed Convolutional Neural Network for Monitoring the Temperature Fields in High-Temperature Gas Reactors

Here, this work presents current advances in applying a physics-informed convolutional neural network (CNN) to evaluate temperature distributions in advanced reactors. Our goal is to demonstrate that the CNN can reconstruct temperature fields within the solid region of a prismatic fuel assembly in a high-temperature gas reactor (HTGR) with sensor data available in only a few cooling channels. Before that, we showcase the superior performance of the physics-informed CNN in comparison to a purely data-driven multilayer perceptron (MLP), considering a canonical heated channel setup. This analysis shows the advantages of our approach and justifies its choice. The datasets employed here are obtained upon numerical simulations performed with codes under the Nuclear Energy Advanced Modeling and Simulation program. This work is important, as industry experience indicates that the assembly material in HTGR concepts is prone to large thermal-mechanical loads nearing operational limits. This makes it crucial to characterize peak temperatures and their distributions near hot spots. Modern thermocouples are unreliable in these types of harsh environments because of the high neutron fluxes and elevated temperatures involved. The CNN-based field reconstruction represents an attractive solution, enabling sensor arrays in less aggressive locations and augmenting indirect predictions for less accessible regions. The results show that the CNN reduces prediction errors by orders of magnitude in comparison to the MLP, considering the simple yet well-representative heated channel case. In the case of the HTGR fuel assembly, the CNN can successfully reconstruct temperature fields over various cooling regimes. Furthermore, we also explore the algorithm’s ability to detect abnormalities. Interestingly, the CNN proves it has the capacity to detect blockage in one of the noninstrumented cooling channels.

Machine learning↗

DOE OSTI2024

Experimental analysis of conductive drying of paper and board

The primary heat transfer source in the conventional multi-cylinder drying of paper and board is conduction. Conduction is facilitated by high-tension contact with steam-heated cylinders, while convective drying, the main mass transfer source, operates as heated air flows over the paper web in the pockets. This conduction process occurs at elevated temperatures and contact pressures to ensure effective contact between the wet paper web and the heated cylinders. The contact pressures and temperatures of the hot surface significantly influence the conductive drying characteristics. This research paper presents an experimental study that involves designing a simple lab-scale setup with in-situ and continuous sensing capabilities for various commercially available grades of paper and board. Embedded thermocouples measure the temperatures of the heated platen and the sheet, allowing the collection of flux data to determine heat transfer characteristics. Here, the goal of this study is to ascertain the instantaneous contact coefficients for different basis weights as a function of moisture content. The acquired data will provide valuable insights and information towards process development, design and simulation of the paper drying process.

42 ENGINEERING↗

DOE OSTI2018

Supersonic Gas Injector for Plasma Fueling in the National Spherical Torus Experiment

A supersonic gas injector (SGI) has been developed for fueling and diagnostic applications on the National Spherical Torus Experiment (NSTX). It is comprised of a graphite converging-diverging Laval nozzle and a commercial piezoelectric gas valve mounted on a movable probe at a low-field-side midplane port location. Also mounted on the probe is a diagnostic package: a Langmuir probe, two thermocouples, and five pick-up coils for measuring toroidal, radial, vertical magnetic field components and magnetic fluctuations at the location of the SGI tip. The SGI flow rate is up to 33.25 Pa m 3 / (1.75 × 10 22 euterium particles/s), comparable to conventional NSTX gas injectors. The nozzle operates in a pulsed regime at room temperature and a reservoir gas pressure up to 665 kPa (5000 Torr). The deuterium jet Mach number of about 4 and the divergence half-angle of 5 to 25 deg have been measured in laboratory experiments simulating the NSTX environment. Reliable operation of the SGI and all mounted diagnostics at distances 0.01 to 0.20 m from the plasma separatrix has been demonstrated in NSTX experiments. The SGI has been used for fueling of ohmic and 2- to 4-MW neutral beam injection–heated L- and H-mode plasmas. Fueling efficiency in the range 0.1 to 0.3 has been obtained from the plasma electron inventory analysis. In conclusion, the SGI-fueling–based plasma discharge scenarios enabling better density control have been developed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Thermal Hydraulics of Helium-Cooled Finger-Type Divertors at Higher Incident Heat Fluxes

Over the last decade, a number of studies at the Georgia Institute of Technology (GT) have evaluated the thermal hydraulics of the design of the helium-cooled modular divertor with multiple jets (HEMJ) originally developed at the Karlsruhe Institute of Technology. Using the GT helium loop, a test section of a single HEMJ finger heated by a radio-frequency (rf) induction heater was studied at near prototypical condition at pressures of ~10 MPa, maximum mass flow rates of 8 g/s, and maximum helium inlet temperatures Ti of 425°C. The area-averaged cooled surface temperature was estimated from embedded thermocouple measurements. This, together with the average incident heat flux q ‾ ′′ , was used to determine the average heat transfer coefficient and the corresponding Nusselt number N u ‾ over the cooled surface. The normalized pressure loss coefficient KL was determined from the pressure drop measured across the test section. The helium loop was modified last year by enclosing the test section and heater within an argon-filled stainless steel chamber to minimize oxidation of the tungsten-alloy test section. Initial results, when extrapolated to prototypical conditions, suggested that N u ‾ was about 20% higher than our previous results. However, the maximum heat flux for these results was less than 3 MW/m2 due to rf coupling with the steel chamber walls. The chamber was then recently upgraded to a glass–stainless steel enclosure with modified feedthroughs for the induction heater connections to minimize this coupling. With this upgrade, a maximum incident heat flux q ‾ ′′ = 8.1 MW/m 2 was achieved. This work presents experimental estimates and correlations for N u ‾ and K L at higher heat fluxes. These results provide greater confidence when estimating the maximum heat flux that can be accommodated by the HEMJ at fully prototypical conditions. Lastly, preliminary metrology results for the test section used to experimentally study the simplified flat design variant of the HEMJ are presented as part of an effort to resolve recently reported discrepancies between experimentally estimated and numerically simulated N u ‾ .

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Predicted Heat Flux Performance of Actively Cooled Tungsten-Armored Graphitic Foam Monoblocks

Tungsten (W)–armored graphitic foam monoblocks were developed for applications requiring high-Z plasma-facing material in long-pulse fusion experiments and ultimately deuterium-tritium fusion reactors. The monoblocks are an integrated material system combining the advantages of a chemical vapor deposited (CVD) W coating with a high-conductivity graphitic foam. The W is a high-melting-point, high-Z material with low tritium retention. The graphitic foam coupled to a swirl tube serves as a high-thermal-conductivity heat sink that cannot melt, although it can sublime at much higher temperatures than copper melts. Together, they comprise a robust plasma-facing component (PFC) weighing roughly 5% of an all-W component or 17% of a traditional W-coated copper heat sink.A single-channel mock-up consisting of four graphitic foam monoblocks equipped with a water-cooled swirl tube was fabricated for eventual testing in the 60-kW, EB-60, rastered electron beam at the Applied Research Laboratory of The Pennsylvania State University. Two monoblocks have a thin 50-μm-thick coating of pure W chemically vapor deposited over NbC and pure Nb interlayers. Two others have a 2-mm-thick pure W coating CVD on graphitic monoblocks using the same interlayers. The mock-up will be cooled with available 10 m/s, 0.7 MPa water with a 22°C inlet temperature and subjected to varying uniform heat loads up to 20 MW/m 2 . It is equipped with type-K thermocouples at various depths, and calibrated infrared thermography and spot pyrometry will be used to characterize the heated surface. Real-time water calorimetry will be used to ascertain the absorbed steady-state power and infer the heat flux during testing. Since testing cannot be done under prototypic divertor flow conditions, it is necessary to predict the thermal response of this novel PFC system and investigate the power sharing between radiation and convection at divertor heat flux levels and its inherent ability to avoid critical heat flux. Results are reported for predictions obtained from computational fluid dynamics models up to 30 MW/m 2 of steady-state uniform heat flux. Leading-edge heat loads of 30 MW/m 2 on a 2-mm-wide side strip were also investigated to ascertain if coating delamination is likely.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

CFD Simulation of Helium Flow Loop Test Section

A helium flow loop is being assembled at Oak Ridge National Laboratory to analyze heat transfer enhancement for systems such as blanket and divertor components. To efficiently identify optimum geometries for heat transfer enhancement in these applications, simulation work is performed to optimize test section designs that are built and tested in the helium flow loop that operates at 4 MPa and a mass flow rate of 100 g/s. Different ribbed geometries that examine rib shape, rib height, rib orientation, rib spacing, and three dimensional orientation are modeled and simulated in STAR-CCM+ to compare their ability to remove heat and mitigate pressure drop. Following the simulations, models are selected and manufactured for the helium flow loop tests. Simulations initially focus on a hydrodynamic study to determine the appropriate mesh and physics models and then add a heat flux to analyze the heat transfer abilities of the models. The simulations are run in steady state and use a Reynolds-averaged Navier-Stokes k-ε turbulence model. The helium is modeled as an ideal gas. The simulation explores models of geometries that enhance the heat transfer and decrease pressure drop with an overall goal of increasing fluid collision with the wall. Enhanced geometries are simulated to select appropriate designs for manufacturing, and preliminary experimental results are used to validate the simulations. Furthermore, the factors that are being analyzed in the comparison between the experimental and the simulated results include matching thermocouple temperatures, pressure drop, roughness, and fluid velocity.

42 ENGINEERING↗

DOE OSTI2022

Thermo-Mechanical Distortion of Tungsten-Coated Steel During High Heat Flux Testing Using Plasma Arc Lamps

An experimental setup and a test section were designed and fabricated for high heat flux testing (HHFT) of neutron-irradiated specimens using water-wall plasma arc lamps. Because of the radiological considerations and limitations of reactor irradiation, the size of the test articles was limited to disks less than 10 mm in diameter. The specimen was clamped onto an actively cooled block, and clamping allowed the insertion of several thermocouples on the back surface of the specimen through a copper (Cu) block. Five vacuum plasma sprayed tungsten (W)–coated F82H steel specimens were subjected to HHFT. Surface profilometry measurements, which were conducted after HHFT, revealed central bowing of the top W surface. This type of residual distortion occurred for all of the specimens, and the larger the specimens were, the larger was the distortion.In an attempt to understand specimen distortion and address the science questions related to the testing of subsize specimens during HHFT, a simplified thermo-mechanical model was developed. By using a measured temperature in the Cu as an isothermal boundary condition, the model eliminated the need for coupling cooling fluid flow models with stress models, greatly simplifying the analysis. The main variable in the proposed model is hC, i.e., the thermal contact conductance between the F82H and the Cu washer. Inelastic properties, including hardening properties, were considered for F82H steel and Cu. Numerical simulation results demonstrated a buildup of residual deformation during HHFT and a very complex state of stress and deformation during typical heat flux (HF) cycling. Additionally, hoop stress evolution during a high heat flux cycle reveals that F82H at an interface with W would be mainly in compression during HF application and experienced a transition to a tension state during cooldown. Also, specimen distortion evolves during each HF cycle, as the specimen bows downward during HF application and upward during the cooldown period between HF cycles. The final specimen distortion, i.e., upward bowing of the specimen center, was qualitatively predicted for hC values of 4000 to 5000 W/(m 2 ·K). This hC range of values, for which bulging is obtained, is at the lower spectrum of the range of values for hC, consistent with the low thermal contact conductance expected from the unpolished F82H surface.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗