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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 577 records · Page 32
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Simulation of CEFR neutronic start-up tests with FENNECS

This paper presents simulation results of selected Neutronic Start-up Tests of the China Experimental Fast Reactor (CEFR) obtained by the neutronics code FENNECS that have been performed within the frame of the IAEA Coordinated Research Program I31032. The Finite Element Neutronics code FENNECS is developed at GRS and solves the few-group steady-state and transient diffusion equation using a Galerkin-based finite element approach. Its main purpose is the safety assessment of Small Modular Reactors and Micro Reactors with complex geometry (e.g., rotating control drums) which gain increased interest internationally. Serpent has been applied to create reference models of the CEFR and cross-section libraries in 10 energy groups for FENNECS. Using these libraries, the following experiments have been simulated with FENNECS: net criticality, control rod integral and differential worth, void reactivity effect, subassembly exchange reactivity effects and reaction rate distribution. The obtained satisfactory agreements with the measurements represent a valuable contribution to the validation of FENNECS. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

A refined assessment of nuclear data target accuracy requirements for ALFRED

The interest in target accuracy assessment has been recently renewed within the OECD/NEA Working Party on International Nuclear Data Evaluation Cooperation with an update exercise, thanks to the invaluable initiative of Massimo Salvatores. This work, moving from evaluations on the main core integral parameters of the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED), and the identification of the associated target accuracies needed for improving its design, approaches the retrieval of target accuracy requirements for nuclear data, as the inverse problem of uncertainty quantification. The preliminary results obtained in a previous work are here refined by explicitly taking into account the total contribution due to correlations among variables in the optimization constraint, while its various components are evaluated individually. Moreover, three different sets of cost parameters are considered to introduce a bias in the solution in order to take in due account the relative difficulties in the execution of new differential experiments as needed for achieving the proposed target accuracies. By this, the obtained results are deemed usefully informative, and proposed for the update of a High Priority List which identifies the isotopes-reaction couples of most interest for future refinement experiments, especially when integrated with further analogous information collected on other advanced nuclear systems. Among the cross-sections whose refinement contributes the most to the reduction of the multiplication factor, the fission of Pu{sup 239} was found, which needs to be fixed below 1% in the range of interest for ALFRED (2 keV-4 MeV). Another contribution for the amelioration of the multiplication factor calculation accuracy, is due to the capture channel of Pu{sup 239}, for which the initial high uncertainty (around 15%) can be significantly reduced resorting to easily practicable experiments. The same occurs for the inelastic of Pb{sup 207}, for which the initial uncertainty value of around 50% can be drastically reduced up to target values between 4% and 7%.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Thermochemical Modeling of Radionuclide Vapor-Liquid Equilibria in Sodium Pools for SFR Mechanistic Source Term Analysis

Mechanistic source term (MST) analysis of sodium fast reactors (SFR) requires understanding of various radionuclide (RN) transport phenomena influencing potential releases from the fuel to the environment. One such phenomenon includes the retention or release of RNs from the sodium coolant pool, representing the step after possible fuel failures and influencing transport to the cover gas region. Thermodynamic vapor-liquid equilibria (VLE) calculations were performed on systems representing SFR sodium pools containing oxygen impurities and radionuclide (RN) inventories. First, an assessment and recreation of a previously developed thermodynamic database was completed, including updates to thermodynamic parameters. The RN inventories used in VLE calculations represented hypothetical source terms that might be released to the pool during previously analyzed fuel failure scenarios. The calculations were performed for all possible combinations of sodium pool size (i.e., total oxygen) and number of failed fuel pins (i.e., total RNs). In this way, multiple ratios of the RN relative to the oxygen impurity (RN:O) were compared for their impact on RN volatility, which is discussed in terms of the vapor fraction (VF), defined as the fraction of the RN that is calculated to exist in the vapor phase at equilibrium above condensed phases of that element. Similar trends in VLE behavior are seen in the equilibrium calculation results for elements of similar chemistry, and for some element types, it was found that the RN:O ratio can be important due to oxide formation, which typically exist in the condensed phase.

Shahbazi, Shayan↗
DOE OSTI2020

Examination of Legacy Metallic Fuel Pins (U-10Zr) Tested in FFTF

The MFF series of metallic fuel (U-10wt%Zr) tests performed in the Fast Fuel Test Facility (FFTF) were the beginning tests to qualify the fuel as a driver fuel for FFTF. They all performed very well, to relatively high burnup and with no pin breaches. Tests MFF-3 and MFF-5 were chosen to be destructively examined because they were run at high peak cladding temperatures, 643 and 649°C respectively, and to modest to high burnup (138 and 101 MWd/kgM respectively). They were the only sodium fast reactor (SFR) metallic tests, with long (91.4 cm) fuel columns and clad in HT9, to be operated at these high temperatures. Extensive operating condition reconstruction was performed to accompany the examination, providing detailed operating conditions axially along the pins. In addition, detailed ORIGEN calculations provided fuel burnup along the pin length. These calculations and detailed cross-section metallography allowed a PhD dissertation to be performed where the fuel/cladding chemical interaction (FCCI) was modeled using Fickian and Soret Effect driven diffusion of rare earth fission products to the cladding surface and interact with the cladding. The examination of the MFF-3 and MFF-5 pins included neutron radiography, axial profilometry including pin bow and length measurements, precision gamma scanning, plenum gas puncturing to measure fission gas release, detailed chemical/isotopic analyses of fuel samples to confirm burnup calculations, and metallography of pin cross-sections, including micro-hardness measurements of fuel and cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Molten Chloride Thermophysical Properties, Chemical Optimization, and Purification

Next-generation concentrating solar power (CSP) technology requires high-temperature heat-transfer fluids and thermal energy storage that can operate in the temperature range of 500°–720°C. This high temperature demands a new chemistry for the heat-transfer fluids that must have high-temperature thermal and chemical stabilities, high energy density, and low corrosion rate on metallic materials used in the heat exchangers, piping, and thermal energy storage tanks. A ternary MgCl 2 -KCl-NaCl chloride-salt system has been proposed to achieve such goals, given its high-temperature thermal stability. The chloride-salt-based systems have been used by the metallurgy industry as the electrolyte to electrolytically extract metallic magnesium. They have also attracted attentions from the nuclear industry for developing the molten salt fast reactors that can operate at close to atmospheric pressure and at higher temperature than traditional light-water reactors. However, the corrosion properties of the ternary salt are not well known. The corrosion mechanism is believed to be correlated to the presence of MgOHCl impurity at the operation temperature. MgOHCl is fundamentally caused by the hygroscopic nature of the MgCl 2 component. Therefore, the first key objective of this project is to understand the nature of the corrosiveness of the ternary chloride-salt system and its prevention by designing new lab-scale purification processes. Because of its importance, a working group called the Chloride Salt Collective across multiple U.S. Department of Energy national laboratories is established to gain experimental agreement on the corrosion behaviors. In addition, water generation during salt dehydration could lead to highly corrosive HCl/water mixture via a hydrolysis process between water and anhydrous MgCl 2 or MgCl 2 hydrates. MgOHCl will eventually be converted to MgO particulates regardless of the chemical path chosen (i.e., by thermal decomposition, chemical purification with an active metal, or electrochemical purification utilizing the reduction power of electron flows). Large MgO particles with high hardness can wear out bearing and bushing and even clog the delicate channels in the primary heat exchanger. Because of this, a second objective is to understand the chemistry behind salt purification and transfer the scientific understanding to a set of engineering know-how for large-scale salt melter design. Because there is limited information on the thermophysical properties and the sensitivity of these properties around the targeted salt composition, a third objective is to measure thermophysical properties relevant to system and component design and support Topic 1 Liquid Pathway development. For example, composition shift due to salt reaction with water can be a potential issue because it can lead to changes of thermophysical properties, such as (1) melting point (or freeze point), which can jeopardize safe operation of the supercritical CO 2 power cycle proposed for next-generation CSP, and (2) heat capacity and density, which can decrease the overall power-generation capability of the plants. The value of the thermal conductivity of the selected salt composition is also a critical criterion for the solar receiver design.

14 SOLAR ENERGY↗
DOE OSTI2020

Requirements and Conceptual Design of Off-gas Systems for the Reprocessing of Metallic Fuels

An assessment has been conducted to determine how key regulations regarding volatile radionuclide emissions to the atmosphere may apply to the off-gas streams associated with electrochemical reprocessing. The scope of this assessment was based upon a generic electrochemical reprocessing scheme with a throughput rate of 200 MTIHM/y applied to metallic fuel discharged from a sodium fast reactor (SFR), but the findings are able to be translated to other advanced nuclear scenarios as merited. Air dispersion modeling was performed using the EPA CAP-88 model and evaluated the uncontrolled decontamination factors (DFs) that would be required to achieve regulatory compliance with the dose-based limits set forth by EPA regulation 40 CFR 190.10(a). These DFs were compared to those required by fuel cycle–based limits set forth by EPA regulation 40 CFR 190.10(b). Two theoretical sites with disparate climatological conditions were selected for air dispersion modeling (Idaho and Tennessee). The radionuclides modeled included 3 H, 85 Kr, 129 I, and selected alpha-emitting transuranic isotopes (referred to here as 239 Pu-TRU <1y ). It was found that the fuel cycle-based limits in 40 CFR 190.10(b) are most restrictive for 85 Kr and 239 Pu-TRU <1y , with DFs of 3 and 6.1E+09, respectively. The dose-based limit as derived from 40 CFR 190.10(a) could require mitigation of tritium in some scenarios, with an estimated DF of about 3 for the reference scenarios. The fuel cycle-based limit for 129 I resulted in a DF of about 240 for the reference scenario. The need for iodine mitigation based on dose to the public depended upon the physical form of iodine as either particulate or vapor-phase species. Emission of iodine from the facility as a vapor necessitated DFs of about 2 but emission as a particulate would require DFs >6,000 to meet thyroid dose-based limits. Effects of physical form on needed iodine mitigation are significant, but the understanding of speciation of iodine both during electrochemical reprocessing and after release to the atmosphere is limited. The electrochemical processing unit operations were evaluated to identify potential release points for the volatile radionuclides and to assess the potential for retention of the radionuclides within the process (thus decreasing the need for mitigation). Mitigation strategies for 3 H, 85 Kr, 129 I, and 239 Pu-TRU <1y were identified. In all cases, there are reasonably achievable pathways to regulatory compliance, although in some cases additional R&D is merited to verify the chemical speciation of these isotopes and to develop and demonstrate potential treatment technologies for this application. Whether or not additional off-gas controls (beyond common operations such as HEPA filtration and oxygen and moisture control) are needed for any of these regulated or volatile radionuclides depends on the (a) type of facility (NRC-regulated or DOE), (b) used fuel process rate, (c) used fuel burnup and composition, (d) speciation and retention of volatile radionuclides in the process and in the cell gas cleanup system, (e) site-specific parameters such as location, meteorology, stack height, and site boundaries, and (f) levels of conservatism and safety factors used in assessing compliance to air emissions regulations. Performance of this assessment revealed several areas where information is lacking or additional research is required in order to better determine if or what kinds of off-gas control might be needed. First, and most significantly, the understanding of the chemical speciation and physical form and partitioning of iodine during electrochemical processing operations is lacking and prevents the ability to accurately assess the potential iodine mitigation requirements. Future research in this area should be multifaceted and include thermodynamic modeling of iodine speciation in different process steps, experiments to quantify the kinetics of vapor-phase and melt-phase transitions, bench-scale experiments to determine the potential chemical and physical form of iodine emissions from the electrorefining process, and verification of iodine behavior with experiments utilizing operational facilities. Similarly, an improved understanding of iodine behavior in the environment after release from the facility stack will be required to refine dose estimations, as particulate and vapor-phase emissions result in significantly different doses to the MEI.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Assessment of MOOSE-Based Tools for Calculating Radial Core Expansion

Radial core expansion in liquid-metal cooled fast reactor systems is a well-known phenomenon that produces strong reactivity feedback effects. An inherently safe reactor design takes advantage of negative reactivity feedback in accident scenarios by utilizing a core restraint system which produces a bowed shape that allows for radial expansion of the fuel regions. Detailed modeling and simulation of core radial expansion itself as well as subsequent reactivity feedback is a challenging task involving contact of many fuel assembly elements and physics feedback from neutronics, thermal hydraulics, and thermal mechanical response. A variety of physics codes have been developed to model aspects of radial core expansion but in general invoke geometrical or physics approximations. No code system currently exists which tightly and robustly couples these physics with enough detail to fully resolve the complex core radial expansion reactivity feedback effects. The future availability of such a code system is of vital importance to fully understanding the reactivity feedback effects that occur due to radial expansion, and consequently to optimizing the design of the core restraint system. A high-fidelity code will also be used to benchmark existing lower fidelity, faster running models to understand their benefits, limitations, and range of applications. A code development path using MOOSE-based tools is proposed in order to leverage the detailed geometry capabilities and natural tight coupling and robustness of MOOSE-based applications for modeling this complex phenomena. While simulation of the full phenomenon involves several physics, an assessment has been initiated on the capabilities and readiness of the currently available Tensor Mechanics module within MOOSE for calculation of the structural mechanical responses which occur within the reactor core. This report focuses on modeling the force-deformation response which mimics the physics of a duct contact deformation, as well as differential thermal expansion which produces a thermal bowed shaped for the fuel assemblies in a core. Simple examples were initially performed such as simple supported beam bending under load. The complexity of examples was progressively increased to better mimic the duct behavior by including a differential thermal example and inclusion of hexagonal cross-sections in the geometry. Further assessment of the structural mechanical response simulation capability is still required for modeling duct contact interactions and irradiation creep and swelling. Companion thermal hydraulic and neutronics assessments will also be required; these activities are planned for future years. Finally, integration of the multiple physics components through MOOSE is required to predict the core radial expansion and subsequent feedback effects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Report on FY 2020 Welding Parameters Optimization and the Fabrication of Qualified Alloy 709 Welds

Alloy 709 is an attractive candidate construction material for the Sodium Fast Reactor (SFR). As part of the Alloy 709 Intermediate Term Test Program, this work covers the development of the technical basis for weld fabrication and weld qualification of Alloy 709. The phase of research reported herein focused on the issue of weld solidification cracking recently experienced in both experimental heats and a commercial heat of Alloy 709 with relatively high phosphorous levels.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Report on FY 2021 Fabrication and Testing of Qualified Alloy 709 Welds at ORNL

Alloy 709 is a candidate construction material for the Sodium Fast Reactor (SFR). As part of the Alloy 709 ASME Code Case development team under the Advanced Reactor Technologies (ART) Program, this work covers the development of the technical basis for weld fabrication and weld qualification of Alloy 709. This report summarizes the research conducted in FY 2021 and relevant results from previous reporting periods, aimed at welding commercial heat of Alloy 709 plates of relatively high phosphorous (P) level in support of ASME code case data package. Welding, code qualification tests, and associated microstructure characterization were carried out in determining the effect of P levels of weld wire on the weldability of the high P commercial heat of Alloy 709. The Alloy 709 weld creep strength was evaluated though preliminary creep testing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Sodium Hydrogen Sensor and Acoustic Flow Sensor Development

This report documents the development of diffusion-type hydrogen meters (DTHMs) as part of a steam generator leak detection system (SGLDS) of sodium-cooled fast reactors (SFR) and the demonstration of the capabilities of acoustic flowmeters, including a commercial UTFM (Flexim FLUXUS F721) and an Argonne HT acoustic flowmeter running on shear-to-longitudinal mode. It also evaluates two Argonne permanent magnet flowmeters (PMFMs).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

SFR Safety Experiments in Support of PGSFR Licensing (Final CRADA Report)

Argonne National Laboratory and Korea Atomic Energy Research Institute entered into this CRADA to perform safety experiments and assessment of experimental results in the area of the metal fuel behavior in severe accident conditions to support the Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) licensing process. The goal was to utilize the existing sodium facilities at ANL to perform experiments aimed at providing validation data associated with the source term determination and metal fuel characteristics which enables subcriticality and coolable geometry in postulated severe accident scenarios.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

DIF3D-VARIANT Usage Guide

The DIF3D software is a key part of the legacy fast reactor analysis software package developed and maintained at Argonne National Laboratory. The DIF3D-VARIANT capability has been pushed to the forefront of usage in the last decade at ANL because of improved computing resources and the focus on reactors with significant leakage. The purpose of this document is to discuss when the DIF3D-VARIANT solver provides a solid benefit to analysis work and discuss the resource requirements and explain why they are so much greater than the other solver options in DIF3D. Additional discussion will be provided on solution accuracy and any deficiencies relative to other solvers available in the literature.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Online Lead/Water Heat Exchanger Sensor/System Feasibility-PNNL (Final Report)

On-line structural health corrosion monitoring in advanced lead fast reactor heat exchangers and molten salt reactor heat exchangers is desirable for detecting tube degradation prior to leaks that may allow mixing of heat exchanger fluids or release of radiological contamination beyond the design containment boundary. This program, On-Line Lead/Water Heat Exchanger Sensor/System Feasibility – PNNL 76092, demonstrates feasibility for an ultrasonic torsional wave mode sensor attached to the outside of a long (30.5-m) heat exchanger tube in the stagnant flow area where the tube joins a heat exchanger plenum and where it is possible to protect a sensor and cable from high-force flows. The sensor must be cable connected to a monitoring instrument near the heat exchanger. The sensor and cable management approach for periodic in-service inspection will be impractical to implement on existing heat exchangers; rather permanently mounted sensors must be installed in conjunction with heat exchanger fabrication. Previous work has shown low-temperature lead zirconate titanate (PZT) piezoceramic sensors are able to detect anomalies of interest in 3.0-m long tubes. These sensors have hereby been extended to a 30.5-m long tube more representative of commercial power heat exchanger designs. The program will continue to investigate higher temperature piezoelectric ceramics and long-term performance of high temperature adhesives and sealants and mechanical pressure coupling for 350-500 °C lead reactor environments and higher temperature (700 °C) molten salt environments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

ARDP Natrium Neutronic Methodology: Argonne Neutronic Assessment of ABR-1000

The Natrium Sodium-cooled Fast Reactor (SFR) concept developed by TerraPower, LLC was selected as one of the advanced reactor designs for demonstration under the Advanced Reactor Demonstration Program Natrium Demo project. In collaboration with TerraPower, the Argonne National Laboratory (ANL) team provide independent reviews of the neutronic methodology being used for the Natrium design. As part of this activity, both the TerraPower and ANL teams agreed to perform independent neutronics analyses of the ABR-1000 reactor problem as a demonstration on the usage of the neutronics methodologies. Comparison of the results from TerraPower and ANL would expose any methodological differences in the modeling approaches. This report describes and presents the Argonne analysis work on the ABR-1000, and will be provided to TerraPower, along with data files, to complete the comparison study. The comparison will cover the methodology to evaluate the reactivity coefficients for a safety analysis, the shutdown margin, and the reconstructed pin power distribution. The methods and models used to calculate the kinetic parameters and the reactivity coefficients (density, Doppler, axial/radial expansion, etc.) are described in detail with recommendations of specific method options and modelling techniques. The control system reactivity worth and the excess reactivity due to temperature defect are calculated for the shutdown margin analysis. The method to evaluate the excess reactivity is discussed in detail. The pin power reconstruction methodology used in SE2ANL, SE2RCT, and DASSH are discussed in this report. The pin power distribution evaluated by different methods are compared and discussed in the report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Documentation on How to Run the NEAMS Workbench GUI on Sawtooth

This document provides guidelines on how to run Moose-based applications from the NEAMS WorkbenchGUI on the HPC platform Sawtooth located at Idaho National Laboratory. The different steps are illustrated with a multi-app example modeling a sodium-cooled fast reactor (SFR) taken from the virtual test bed (VTB) website. The workflow consists of four steps to demonstrate the capabilities of the NEAMS Workbench GUI, that covers log into Sawtooth, editing and validating input files, submission of the job to the queue and visualization of the numerical solution and the geometry. Each step is illustrated with figures in the text, and a demonstration video is also available.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Progress Report on SFR Metallic Fuel Data Qualification

This report summarizes the progress of SFR metallic fuel qualification related activities, which are focused on providing quality assurance relevant information applicable to experiments irradiated during the Integral Fast Reactor (IFR) program. A background of metallic fuel performance data and the associated databases, including the EBR-II Fuels Irradiation & Physics Database (FIPD) and Out-of-Pile Transient Database (OPTD), is included. The legacy data in the databases, including as-built, post-irradiation examination (PIE), operating parameters, and out-of-pile experiment data are introduced. The SFR metallic fuel Quality Assurance Program Plan (QAPP) and its implementation to qualify these legacy data is described in detail. Important PIE data QA documents and the specifications of four types of PIE measurements (contact profilometry, laser profilometry, neutron radiography and gamma scan) are provided. Examples of the implementation of the QAPP to qualify each of those types of PIE data are provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Impact of Glass Irradiation on Laser-induced Breakdown Spectroscopy Diagnostics in the Visible and NIR Range

Increased absorption of optical materials arising from the exposure to ionizing radiation must be accounted for to accurately analyze the laser-induced breakdown spectroscopy (LIBS) data retrieved from high-radiation environments. We evaluate this effect on a specifc example that mimics the diagnostics placed within a gas-cooled fast reactor coolant stream. Analysis is performed on the LIBS data measured with 1% Xe gas in an ambient He environment and the measured optical absorption from the gamma- and neutron-irradiated low-OH fused silica and sapphire glasses. No significant change in the number of shots required to reach a 3σ detection level was observed up to 10 Mrad gamma dose and exposure to a 1.7×10 17 n/cm 2 neutron fluence. In contrast, the spectral dependence of absorption results in either a systematic overestimation or underestimation of line intensity ratios, depending on the line of interest selected for analysis. Moreover, if lines from different spectral regions are used to create Boltzmann plots, this attenuation also leads to statistically significant changes in the calculated temperatures for calculations using Xe II lines only, lowering them from 8000±610 K to 6800±810 K for the case of exposure to the 1.7×10 17 n/cm 2 flux, and broadening the temperature range required for a 95% confidence interval. In the case of measuring the Xe spectrum, these effects may be mitigated using only the longer-wavelength spectral region, where radiation-attenuation is relatively small, or through analysis using the iterative Saha-Boltzmann method.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2022

Flow Sensor Test Article (F-STAr) (Design and Fabrication Status Report)

The Flow Sensor Test Article (F-STAr) is a test article currently under construction for the Mechanisms Engineering Test Loop (METL). F-STAr was designed to provide high sodium flowrate capabilities for sensor calibration, component testing, and fluid studies. Figure 1 shows two solid model views of the new test article. F-STAr includes a high-capacity pump that can provide a nominal flowrate of 120 GPM; a test section support structure that can accommodate a wide array of sub-test articles and their instrumentation; and finally, a heating and cooling system to aid in controlling the testing environment. Initially, F-STAr will be configured to test liquid metal flow sensors, specifically field shift sensors like the Eddy Current Flow Sensors (ECFS) based on the RDT C4-7T standard. To test these sensors, two test sections were designed that attempt to model Sodium Fast Reactor (SFR) outlet conditions. Figure 2 shows models of the “Full-Scaled Test Section” (FSTS) that represents a generic SFR fuel handling socket and a “Pseudo-Scaled Test Section” (PSTS) that represents a generic array of scaled fuel handling sockets. While F-STAr will be configured to test ECFS’s, it can also be outfitted to meet other experimental needs. For example, F-STAr could be setup with a test section that includes a fluidic diode or a component test investigating the performance of hydrodynamic bearings. Other test sections include studies of sodium thermal hydraulics like thermal striping. Overall, F-STAr is a flexible test article designed to accommodate many needs. This report will provide a status update on the design and construction of F-STAr. Manufacturing of all components has commenced, and several components have already been completed. These components include the pump and test section assembly stand. Other components, such as the heater, have been completed but were rejected due to the vendor not meeting the requirements of the purchase order. Lastly, the submersible flowmeter and main flange components are under construction, and their status will be reviewed in this report. Overall, most of the F-STAr components and parts will be completed by the end of September 2022.

42 ENGINEERING↗