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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Current Status of the High Field Cable Test Facility at Fermilab

Fermi National Accelerator Laboratory (FNAL) and Lawrence Berkeley National Laboratory (LBNL) are collaborating to construct a new High Field Vertical Magnet Test Facility (HFVMTF) designed for testing superconducting cables in high magnetic fields. This state-of-the-art facility will be situated at Fermilab and will provide capabilities comparable to EDIPO at PSI and FRESCA2 at CERN. The HFVMTF’s background magnetic field, reaching 15 T, will be generated by a magnet supplied by LBNL. The HFVMTF is a collaborative effort supported by the US DOE Offices of Science, High Energy Physics, and Fusion Energy Sciences. It will serve as a vital testbed for superconducting HTS cables, subjecting them to high magnetic fields and a wide range of temperatures, benefiting both scientific communities. Additionally, this facility will play a key role in testing high-field superconducting magnet models and demonstrators, including hybrid magnets, developed by the US Magnet Development Pro gram (MDP). These hybrid magnets, utilizing both LTS and HTS superconductors, are significant advancements toward achieving 18+ T dipoles for future hadron-hadron colliders. The presentation outlines the current status of the facility, covering aspects such as construction progress, cryostat designs, top and lambda plates, and systems for powering, quench protection, and monitoring.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2024

Molten Salt Loop Operational Experience and Test Campaigns in FY24

The Facility to Alleviate Salt Technology Risks (FASTR) at the US Department of Energy (DOE) Oak Ridge National Laboratory (ORNL) was developed to demonstrate technology for high-temperature chloride salt systems (Figure 1). FASTR is primarily constructed using alloy C-276 and is designed to operate at temperatures of up to 725°C. The facility is loaded with 250 kg of NaCl-KCl-MgCl 2 salt. This salt provides a relevant test environment for de-risking technology while avoiding the costs and hazards associated with beryllium-based or uranium-bearing salts. The facility’s major components include a centrifugal pump for salt circulation, an air-based heat exchanger to reject heat, a suite of instrumentation, and trace heating to prevent salt freezing. The salt was purified in 2020 and 2022, and the pumped loop first operated in December 2022. FASTR is a unique US capability for high-temperature molten halide salt testing. FASTR’s scale, co located purification system, and relatively large power (465 kW) differentiates it from other testing systems. Furthermore, access to the DOE-supported facility and efficient communication of results— which are generally disseminated publicly—distinguish FASTR as being broadly significant throughout the molten salt reactor community. FASTR is similar to ORNL’s Liquid Salt Test Loop (LSTL), although FASTR contains chloride-based salt instead of the fluoride-based salt (LiF-NaF-KF) found in LSTL. Furthermore, FASTR is approximately 2× larger than LSTL in terms of pipe size and length, power, salt volume, flow rate, and number of thermocouples. The LSTL first operated in 2016. At the end of FY23, there was a suspected gas leak in the LSTL that halted operation. At the start of FY24, a leak in the LSTL pump’s tank gas space was confirmed. Because the gas-space leak prevented operation of LSTL, FY24 efforts were focused on operation of FASTR. This report summarizes the progress made during FY24 in support of the DOE Office of Nuclear Energy (DOE-NE) work package, AT-24OR070202 Salt Loop and Capability for Testing Sensors and Off Gas Components.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Using optical fibers to examine thermal mixing of liquid sodium in a pool-type geometry

Sodium fast reactors (SFR) are poised to be a leading candidate for the next generation of commercial nuclear reactor deployment. Companies are commissioning designs for SFR increasing the need for experimental and computational analysis to improve the economics and safety of these reactors. Analysis of SFR behavior during reactor transients directly informs the understanding of reactor safety. Particularly, investigating the thermal transients associated with the reactor loss of flow transient allows for informed design decisions to be made regarding reactor safety. During loss of flow transients the flow rate of the coolant is significantly reduced, but not necessarily stopped. The reduction in flow rate results in coolant temperatures exiting the reactor core that can be drastically different than the coolant in upper pool. When these different temperature fluids interact, thermally stratified layers can form causing cyclic thermal fatigue on the reactor vessel potentially leading to pre-material failure. Past experimental studies have lacked high resolution temperature measurements. This research focuses on demonstrating the reliability of novel techniques for acquiring high resolution temperature measurements in a scaled liquid sodium facility. Tests are conducted that simulate postulated loss of flow transients and provide unprecedented spatial temperature distribution measurements.

42 ENGINEERING↗
DOE OSTI2022

HTS Cable Conductor for Compact Fusion Tokamak Solenoids

Significant progress has been made recently in the US fusion community to develop a strategic plan to enable engineering design and construction of a fusion pilot plant (FPP). Princeton Plasma Physics Laboratory (PPPL) is working on developing high current density HTS conductors for next fusion experiments. Partnering with the US industry, we are evaluating feasibility and affordability of Conductor on Round Core (CORC ® ) developed by Advanced Conductor Technologies (ACT) for the next compact fusion tokamak facility. High current density achieved by a CORC ® cable based on a four-layer model coil recently tested at National High Magnetic Field Laboratory (NHMFL) motivated its consideration for low cost, reduced size fusion magnet application. This is of interest to PPPL because of its scalability to tokamak central solenoid (CS) coils in terms of required flux swings for plasma startup operations. Furthermore, partnering with ACT, we designed and built a two-layer model coil solenoid directly wound with CORC to demonstrate its applicability for compact solenoids such as that used in the national spherical torus experiment (NSTX), NSTX-upgrade (NSTX-U) and the US sustained high-power density test facility (SHPD). The ~160 mm diameter solenoid wound by a two-layer CORC is being tested in early 2022 under electromagnetic cyclic loading at NHMFL in a unique 160 mm bore, 14 T background field magnet facility.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2022

Upgrade of Core Conduction Cooldown Test Facility Control System

The CCCTF was designed and installed at ORNL in the 1990s for high-temperature gas-cooled reactor accident simulation. It is located within a modular hot cell housed in the IFEL facility. In 2008 and 2009, modifications were made to the CCCTF in preparation for AGR safety testing. A new airlock was designed and installed to better support periodic exchange of the fission product deposition cups while maintaining the fuel at safety test temperature, and modifications were made to add routines to the existing user interface software for computer-controlled automation of the He sweep gas flow and liquid nitrogen supply to the 85Kr cold traps. However, the furnace control hardware and software remained essentially unchanged. At the conclusion of the AGR-1 and AGR-2 PIE campaigns, continued use of the CCCTF was in question because of the gradual obsolescence of the hardware and software used to control the furnace and ancillary systems that make up the CCCTF. Therefore, an effort was undertaken to replace all the obsolete components with new hardware and update the software to increase system reliability and ensure this capability’s maintenance for future research and fuel qualification efforts. The modifications included creating a new National Instruments (NI) LabVIEW-based control system with a graphical user interface (GUI).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Progress Towards Developing Neutron Tolerant Magnetostrictive and Piezoelectric Transducers

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is not feasible to obtain high resolution/microscale data with the current state of instrumentation technology. However, ultrasound-based sensors offer the ability to obtain such data if it is demonstrated that these sensors and their associated transducers are resistant to high neutron flux, high gamma radiation, and high temperature. To address this need, the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) is funding an irradiation, led by PSU, at the Massachusetts Institute of Technology Research Reactor to test the survivability of ultrasound transducers. As part of this effort, PSU and collaborators have designed, fabricated, and provided piezoelectric and magnetostrictive transducers that are optimized to perform in harsh, high flux, environments. Four piezoelectric transducers were fabricated with either aluminum nitride, zinc oxide, or bismuth titanate as the active element that were coupled to either Kovar or aluminum waveguides and two magnetostrictive transducers were fabricated with Remendur or Galfenol as the active elements. Pulse-echo ultrasonic measurements of these transducers are made in-situ. This paper will present an overview of the test design including selection criteria for candidate materials and optimization of test assembly parameters, data obtained from both out-of-pile and in-pile testing at elevated temperatures, and an assessment based on initial data of the expected performance of ultrasonic devices in irradiation conditions

Reinhardt1, Brian↗
DOE OSTI2024

Development and demonstration of a BISON–Griffin modeling framework for the design of targeted TRISO transient experiments in the Transient Reactor Test Facility

Uranium oxycarbide (UCO)-bearing tri-structural isotropic (TRISO) particle fuels are expected to be used in numerous U.S. commercial reactor applications within the next decade. Here, in this work, we reviewed historical particle fuel transient experiments to identify gaps in TRISO fuel performance transient testing. A BISON–Griffin modeling framework was then developed to conduct preliminary TRISO transient analyses and begin to address these gaps. The framework was demonstrated using limiting-case transient conditions from a prototypic high-temperature gas-cooled reactor (HTGR). It was then applied to develop a matrix of experiments that could be performed in the Transient Reactor Test Facility (TREAT) to (1) evaluate UCO-fueled particle performance at moderate and high heat rates, (2) assess whether historical testing involving UO 2 -fueled particles is applicable to modern UCO-fueled particles, (3) deconvolute the impacts of temperature and heat rate on particle transient response, and (4) collect the data needed for fuel performance model validation and/or further development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

NMDQi Nuclear Materials Discovery and Qualification Initiative Conference Overview

The Nuclear Materials Discovery and Qualification Initiative (NMDQi) is designed to accelerate nuclear materials qualification to fulfill the promises of early and advanced reactor technologies as a safe, clean, and low-cost baseload energy. Materials development and qualification in the nuclear industry is by definition challenging due to stringent safety requirements, limited availability of specialized facilities for materials irradiation and testing, and the challenging high-temperature, high-radiation environment. NMDQi will establish tools and capabilities that will greatly accelerate the nuclear fuels and materials development process. These tools will provide both computationally informed insights and high-throughput infrastructure with the goal of completing qualification in a single pass. In this approach, materials must be fabricated with a range of properties of interest so that materials performance can be examined in parallel, rather than with multiple discrete specimens. Advanced manufacturing (AM) techniques, which can produce complex component geometries, microstructures, and compositions, are ideal. The improved process monitoring and control that AM can provide are ideal for ensuring and reducing material variability, which will be particularly important as standards committees pursue regulations for AM components. However, gaining these benefits requires overcoming the substantial barrier of qualifying AM processes and the resulting materials and components for supporting research campaigns and eventually nuclear service.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Overview of Nuclear Materials Discovery & Qualification initiative (NMDQi)

The Nuclear Materials Discovery and Qualification Initiative (NMDQi) is a program launched in 2020 to address the need for development of new nuclear materials on shorter timelines. The NMDQi has a goal to accelerate nuclear materials qualification to fulfill the promises of early and advanced reactor technologies as a safe, clean, and low-cost baseload energy. Materials development and qualification in the nuclear industry is by definition challenging due to stringent safety requirements, limited availability of specialized facilities for materials irradiation and testing, and the challenging high-temperature, high-radiation environment. Outside the nuclear industry, options exist for increasing the efficiency and rate of materials discovery for accelerated technology commercialization, which can also be applied for our use through the NMDQi. Focus will be on generating tools and capabilities that integrate experimental and computational techniques, allowing materials to be selected prior to fabrication, and providing crucial data to improve upon modeling.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Design and Construction of a High Field Cable Test Facility at Fermilab

Fermi National Accelerator Laboratory, together with Lawrence Berkeley National Laboratory, is building a new High Field Vertical Magnet Test Facility (HFVMTF) to be situated in the Magnet Test Facility at Fermilab. The HFVMTF is jointly funded by the US DOE Offices of Science, High Energy Physics, and Fusion Energy Sciences, and will serve as a superconducting cable test facility for both communities. The background magnetic field for test samples is 15 T and will be produced by a magnet provided by LBNL operating at 1.9 K in superfluid helium. The samples will be placed in the background magnetic field, cooled to between 4.5 K and a user-specified upper limit, and will be powered with a super-conducting transformer at up to 100 kA. Additionally, this facility will be used to test high-field superconducting magnet models and demonstrators, including hybrid magnets, produced by the US Magnet Development Program. Presently, the various tasks of the project are at different stages of execution, from conceptual to ready-for-construction designs. Here, this paper describes the parameters and design status of the pit construction, cryostat, heat exchanger, lambda plate, power system, and quench protection and monitoring systems of the facility.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Pre-Test Feasibility Study Of Reactivity Transient Testing On TRISO Fuel In The TREAT Reactor

Uranium oxycarbide (UCO)-bearing tri-structural isotropic (TRISO) particle fuels are expected to be used in numerous US commercial reactor applications within the next decade. In this work, we reviewed historical particle fuel transient experiments to identify gaps in the TRISO fuel performance transient testing. A BISON--Griffin modeling framework was then developed to conduct preliminary TRISO transient analyses and begin to address these gaps. The framework was demonstrated using limiting-case transient conditions from a prototypic high-temperature gas-cooled reactor. It was then applied to develop a matrix of experiments that could be performed in the Transient Reactor Test Facility (TREAT) to (1) evaluate UCO-fueled particle performance at moderate and high heat rates, (2) assess whether historical testing involving UO$_2$-fueled particles is applicable to modern UCO-fueled particles, (3) deconvolute the impacts of temperature and heat rate on particle transient response, and (4) collect the data needed for fuel performance model validation and/or further development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Development of a Test-Bed for Testing and Refining EarthEn’s Supercritical CO 2 Based Energy Storage System

EarthEn’s energy storage concept leverages supercritical carbon dioxide (sCO 2 ) as a working fluid and relies on compact, high-performance components operating at elevated pressures and temperatures. To accelerate component development and reduce technical risk prior to larger-scale demonstrations, Oak Ridge National Laboratory (ORNL) developed a 100 kW-scale sCO 2 test-bed under a Cooperative Research and Development Agreement with EarthEn (CRADA NO. NFE-24-10050). The objective of the work was to design and construct a flexible experimental facility capable of reproducing key thermodynamic state points and heat-transfer conditions relevant to EarthEn’s thermal energy storage (TES) cycle, with particular emphasis on enabling development and evaluation of next-generation heat exchangers and TES concepts. The test-bed consists of a closed-loop sCO 2 circulation system housed within an open-topped enclosure. In its as-installed configuration, dense-phase sCO 2 is recirculated through a printed circuit recuperator, an electrically heated section, a throttling device used to simulate turbine expansion, and a water-cooled printed circuit heat exchanger that rejects heat to the building chilled-water system before returning to the pump. The pump is driven by a variable frequency drive, enabling controlled adjustment of flow and operating point. A comprehensive instrumentation suite was integrated to support both safe operation and high-quality data collection. Installed sensors include Coriolis flow meters for sCO 2 flow rate and density, resistance temperature detectors and thermocouples distributed throughout the loop (including the heated section and key heat exchanger ports), and pressure transducers for absolute and differential pressure measurements. The facility was designed to support high-pressure (19 MPa nominal) and high-temperature (575°C nominal) operation with credited overpressure protection provided by a rupture disk. Nominal operating conditions were selected to support 100 kW-class testing while maintaining flexibility for non-heated and heated shakedown, control development, and future integration of advanced TES test sections. In parallel with facility development, a system-level thermal-hydraulic model was created using Modelica-based tools to support component sizing, anticipate performance over targeted test conditions, and establish a framework for future model calibration against experimental data. At the conclusion of the project performance period, the facility was in final assembly, and the pressure boundary was nearly completed. However, several practical challenges associated with high-pressure/high-temperature systems and specialized component procurement impacted schedule and prevented initial pump-driven operation and full commissioning within the available resources. This report documents the as-built design, operating capabilities, and instrumentation, and it summarizes key lessons learned related to heater fabrication and testing, first-of-a-kind assembly factors, specialty flange supply constraints, and fill pump corrective actions. Finally, it outlines a phased plan for future commissioning and experimental campaigns, including control and instrumentation shakedown, heater characterization, model calibration, and testing at state points representative of EarthEn’s TES cycle.

25 ENERGY STORAGE↗
DOE OSTI2023

Molten Salt Loop Testing of Sensors and Off-Gas Components: FY23 progress

The Liquid Salt Test Loop (LSTL) at the US Department of Energy’s (DOE’s) Oak Ridge National Laboratory (ORNL) was developed to demonstrate technology for high-temperature fluoride salt systems (Figure 1). The LSTL is primarily constructed using Inconel 600 alloy and operates at temperatures of up to 700°C. The facility is loaded with 165 kg of LiF-NaF-KF salt (FLiNaK). This salt provides a relevant test environment for de-risking technology while avoiding the costs and hazards associated with beryllium-based or uranium-bearing salts. FLiNaK is also an advantageous salt for the secondary side of molten salt reactors. The facility’s major components include a centrifugal pump for salt circulation, an air-based heat exchanger to reject heat, a suite of instrumentation, and trace heating to prevent salt freezing. Additional heating is available through an induction heater rated at 200 kW. The relatively large heating and cooling capability enables the formation of a temperature gradient across the loop (i.e., a hot and a cold side), which is important for chemistry and corrosion studies. The LSTL is a unique US capability for high-temperature molten halide salt testing. Although some efforts are underway at universities, the LSTL’s scale, co-located purification system, and relatively large power differentiates it from other testing systems. Furthermore, unlike efforts within industry, access to the DOE-supported facility and communication of results, which are generally disseminated publicly, result in a broad significance in the molten salt reactor community.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

An Investigation of Liquefaction in Irradiated TRIGA Fuel Exposed to Relatively High Temperatures

This report presents the findings of an investigation into high temperature fuel cladding chemical interactions (FCCI) in Training, Research, Isotopes, General Atomics (TRIGA) fuel rods. A TRIGA fuel-rod core or meat is principally composed of uranium (U) particles dispersed in a zirconium hydride (Zr H) matrix. The fuel is clad in sealed 304SS or Incoloy 800 tubes. At high temperatures, the fuel will interact with the cladding, resulting in FCCI. To investigate the FCCI (in this case, liquefaction), irradiated TRIGA fuel rods were exposed to relatively high temperatures in furnace tests. The tests were performed at different temperatures using segments from an irradiated TRIGA fuel rod at the Hot Fuel Examination Facility located at Idaho National Laboratory (INL). The data from this study is important for developing a better understanding of the performance of U-ZrH fuel type during transient events (e.g., a Loss of Coolant Accident (LOCA) event) that may result in high temperatures while operating a TRIGA reactor. Furnace tests were run for 6 hours at 730, 800, 900, 950, and 1,000°C. One test was run for 12 hours at 950°C. Post-test microstructural characterization was performed on the heat-treated samples using optical metallography to look for evidence of liquefaction. Additionally, microstructural characterization was performed using scanning electron microscopy to investigate the microstructure of the as-irradiated fuel before heat treatment and the microstructural changes that occurred after heat treating for the 730 and 950°C heat treatments for 6 hours. Results of the analysis were compared to those produced during diffusion studies performed using unirradiated TRIGA fuel meat and type 304 stainless steel. A gap is typically present in as-irradiated fuel that limits any possibility for chemical interaction between the fuel and cladding. Optical metallography (OM) was used to identify areas in the specimens heat treated at 950 and 1000°C where fuel and cladding contact was satisfactory for the FCCI to transpire. The bulk of the interaction zones that form due to FCCI developed in the fuel meat and not as much in the cladding. Even in the sample tested at 1,000°C, no evidence was found that gross fuel meat or cladding melting had occurred. Only limited porosity was found in the fuel meat that possibly could have been due to melting of a particular precipitate phase or phases that had formed during FCCI. These results agreed with those reported for the unirradiated diffusion couple studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

UHT-CAMANCHE: Ultra-High Temperature Ceramic Additively Manufactured Compact Heat Exchangers

The conceptual basis for this project is the convergence of advanced ultra-high temperature ceramic materials and additive manufacturing technologies to produce compact ceramic heat exchangers with complex internal flow path geometries. Task areas were broadly divided into materials and manufacturing development, heat exchanger design, component testing, and techno-economic analysis. Technical challenges included the design and commissioning of new test facilities, improving feature resolution and deposition rate of ceramic additive manufacturing techniques, establishing process-structure-property relationships in additively manufactured ultra-high temperature ceramics, and assessing high temperature materials compatibility in CO 2 environments. The primary candidate material evaluated in this work is a composite comprising zirconium diboride (ZrB2) with 30 vol. % silicon carbide (SiC) which was selected based on its desirable combination of high temperature mechanical properties, high thermal and electrical conductivities, and oxidation resistance. High solids loaded ZrB2-SiC pastes suitable for extrusion-based additive manufacturing were developed for the first time as part of this work. Materials compatibility studies indicate this material oxidizes in CO 2 to form a protective borosilicate scale which transforms to pure silica above 1000°C. Parts made by additive manufacturing displayed enlarged grain sizes produced by pressureless sintering as compared to hot-press sintering. Increases in microstructural coarseness have outsized effect on oxidation performance up to 1400°C due to incomplete oxidation of coarse large diameter SiC particles resulting in lower amounts of silica that apparently inhibit protective scale formation. Additive manufacturing as a forming technique did not appear to significantly affect thermal conductivity, hardness, or elastic modulus, though flexural strength was reduced by half or more as compared to traditionally hot-pressed materials. This effect was attributed to the presence of strength-limiting flaws (ca. 40 microns in size) originating from extrudate inhomogeneities that could potentially be eliminated with further process improvements. Attempts to attain economies of scale for production of multi-kilowatt scale heat exchangers by ceramic additive manufacturing proved difficult. Lack of automation and a modest extrusion rate while retaining fine feature resolution made the overall process labor intensive and limited experimental throughput. A number of full-scale components were taken through post-process heat treatments including drying, binder burnout, and sintering; however, none survived without significant flaws or cracks. Therefore, no operational data from a newly installed heat exchanger test loop were able to be obtained during the performance period. Continued research and development is recommended to improve economic feasibility of ceramic additive manufacturing by standardizing the use of advanced sensors, artificial intelligence, and automation tools to reduce associated labor costs and accelerate production rates. The materials and manufacturing techniques demonstrated in this work are likely to find applications in defense and energy applications.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2022

Investigations of Advanced Measurement Methods for High-Temperature Heat-Pipe Experiments

Various types of microreactor designs are under consideration, and among those, the heat-pipe cooled reactor designs are currently being actively investigated at Idaho National Laboratory (INL). In order to support non-nuclear testing and demonstration of technology under the Department of Energy (DOE) Microreactor Program (MRP), Single Primary Heat Extraction and Removal Emulator (SPHERE) was designed and constructed to perform high-temperature heat pipe (HP) experiments at INL. Based on the SPHERE experimental setup, this summary paper conducts the preliminary investigations among the available literature and summarizes some potential advanced measurement techniques for operando high-temperature heat pipe which can be implemented to support MRP. The primary objective for the extensive literature research is to investigate the possibilities and feasibilities for designing and construct an advanced experimental test facility with the aim of producing high-fidelity high-resolution heat pipe data during its operations.

42 ENGINEERING↗
DOE OSTI2025

Thermal Hydraulics Validation Activities in the ART-GCR Campaign

The Advanced Reactor Technologies - Gas Cooled Reactor Campaign (ART-GCR) is part of the United States Department of Energy - Office of Nuclear Energy's Advanced Reactor Technologies program aimed at developing a High Temperature Gas-cooled Reactor (HTGR) which will offer enhancements in safety and efficiency. The Design, Methods, and Validation arm of the ART-GCR program provides experience and advanced tools for HTGR design and analysis. The Methods and Validation arm is developing data to validate predictions of decay heat removal through an extensive experimental campaign the Natural Convection Shutdown Heat Removal Test Facility at Argonne National Laboratory. In-core thermal hydraulics validation is being facilitated through the OECD/NEA's Thermal Hydraulic Code Validation Benchmark for High Temperature Gas-Cooled Reactors using HTTF Data. That benchmark contains code-to-code and code-to-data comparisons for 3 sets of HTGR thermal hydraulics phenomena. This ongoing benchmark activity has 16 participants from 9 countries developing leading insights into HTGR thermal hydraulics code validation. The ART-GCR campaign is also working on the development of state-of-the-art MOOSE-based models of the High Temperature Engineering Test Reactor (HTTR). Through the OECD/NEA's HTTR Loss of Forced Cooling (LOFC) program, the ART-GCR campaign is developing a multiphysics transient code validation benchmark based on 3 HTTR experiments that will provide an opportunity for international collaboration on multiphysics code validation.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

High-Fluence Active Irradiation and Combined Effects Testing of Sapphire Optical Fiber Distributed Temperature Sensors

The goal of this work was to investigate the in-core performance of sapphire optical fiber temperature sensors and to develop clad sapphire optical fibers for in-core instrumentation. We fabricated clad sapphire optical fibers and evaluated the distributed sensing performance of these sensors via optical backscatter reflectometry under high fluence and combined radiation and temperature effects. A series of irradiations was completed to evaluate the effect of irradiation on sapphire optical fiber temperature sensors and to determine the operational limits of these sensors. (1) Objective 1: Fabricate sapphire optical fiber sensors. (2) Objective 2: Evaluate the clad sapphire fiber to verify single-mode behavior and determine and characterize the light modes supported by optical fibers. (3) Objective 3: Characterize the in-core temperature sensing of sapphire optical fiber, as well as the combined temperature and irradiation effects. (4) Objective 4: Evaluate the lifetime and performance of the sensor under irradiation to high neutron fluence. Objectives 1, 2, and 3 were completed during the first 2 years of the project. Due to the Covid pandemic, Objective 4, a high-fluence irradiation performed at the Massachusetts Institute of Technology Research Reactor (MITR), was delayed, as partner facilities were subject to mandatory shutdowns and required a 1 year, no-cost extension. This irradiation was eventually completed on December 12, 2022. This work indicates that sapphire optical fiber sensors may be a solution for ultra-high-temperature applications in which traditional silica optical fibers are prone to fail. Sapphire sensors are potentially suitable for experiments featuring temperatures above 700°C for long periods of time, or for any length of time above 1000°C. Experiments featuring a low total fluence, such as irradiations conducted in the Transient Reactor Test (TREAT) facility, also represent good applications for sapphire optical sensors. Additional work is required to characterize the sapphire fiber cladding performance, which falls outside the scope of this project, as well as the effects of high temperatures on the response of the fiber. A comprehensive material study is recommended as future work to evaluate the attenuation in sapphire under irradiation, and how that attenuation changes with irradiation temperature. The drift and attenuation in the fiber at temperatures of up to 1600°C and a total fluence of up to 2.9 x 10 17 n/cm 2 was minimal, and the fibers returned to baseline after being heated to 1600°C under irradiation. This is promising for the future use of sapphire optical fibers in advanced reactors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗