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DOE OSTI2024

Single Primary Heat Extraction and Removal Emulator (SPHERE) Long Duration Testing

For heat-pipe cooled microreactor development, it is essential to understand the characteristics of heat pipes and how they function under a wide range of operating conditions. An important process for heat-pipe cooled microreactor development is the passive heat removal and how the performance may change over long periods of time. Increased experimental data provides additional information to assess the operational lifetime of alkali metal heat pipes. Idaho National Laboratory has completed testing over a long duration of a high-performance sodium filled heat pipe while monitoring axial temperature profile, power supplied by the heaters, and heat removed by a gas-gap calorimeter. The results from this testing can aid in heat pipe validation efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Single Primary Heat Extraction and Removal Emulator (SPHERE) Long Duration Testing

For the development of heat-pipe cooled microreactors, it is crucial to thoroughly understand the characteristics and functioning of heat pipes across a wide spectrum of operating conditions. Passive heat removal and its long-term performance stability are critical factors in this context. Enhanced experimental data is vital for evaluating the operational lifespan of alkali metal heat pipes. Idaho National Laboratory (INL) has successfully conducted an extended duration test on a high-performance sodium-filled heat pipe, closely monitoring the axial temperature profile, power supplied by the heaters, and heat removed by a gas-gap calorimeter. The results from this testing provide valuable data that are instrumental in supporting heat pipe validation efforts. Specifically, this data aids in the development and validation of Sockeye, the Multiphysics Object-Oriented Simulation Environment (MOOSE) tool under the US-DOE NEAMS program designed for heat pipe modeling. By comparing experimental results with Sockeye’s predictions, the tool's accuracy and reliability can be assessed and improved, thereby enhancing its capability to simulate heat pipe operations under various conditions.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1993

Test results of a Stirling engine utilizing heat exchanger modules with an integral heat pipe

The Heat Pipe Stirling Engine (HP-1000), a free-piston Stirling engine incorporating three heat exchanger modules, each having a sodium filled heat pipe, has been tested at the NASA-Lewis Research Center as part of the Civil Space Technology Initiative (CSTI). The heat exchanger modules were designed to reduce the number of potential flow leak paths in the heat exchanger assembly and incorporate a heat pipe as the link between the heat source and the engine. An existing RE-1000 free-piston Stirling engine was modified to operate using the heat exchanger modules. This paper describes heat exchanger module and engine performance during baseline testing. Condenser temperature profiles, brake power, and efficiency are presented and discussed.

Skupinski, Robert C.↗

NASA NTRS1999

Stirling Space Engine Program: Appendixes A, B, C and D - Volume 2

The objective of this program was to develop the technology necessary for operating Stirling power converters in a space environment and to demonstrate this technology in full-scale engine tests. Volume 2 of the report includes the following appendices: Appendix A: Heater Head Development (Starfish Heater Head Program, 1/10th Segment and Full-Scale Heat Pipes, and Sodium Filling and Processing); Appendix B: Component Test Power Converter (CTPC) Component Development (High-temperature Organic Materials, Heat Exchanger Fabrication, Beryllium Issues, Sodium Issues, Wear Couple Tests, Pressure Boundary Penetrations, Heating System Heaters, and Cooler Flow Test); Appendix C: Udimet Testing (Selection of the Reference Material for the Space Stirling Engine Heater Head, Udimet 720LI Creep Test Result Update, Final Summary of Space Stirling Endurance Engine Udimet 720L1 Fatigue Testing Results, Udimet 720l1 Weld Development Summary, and Udimet 720L1 Creep Test Final Results Summary), and Appendix D: CTPC Component Development Photos.

Dhar, Manmohan↗

NASA NTRS1986

Heat pipe cooling for scramjet engines

Liquid metal heat pipe cooling systems have been investigated for the combustor liner and engine inlet leading edges of scramjet engines for a missile application. The combustor liner is cooled by a lithium-TZM molybdenum annular heat pipe, which incorporates a separate lithium reservoir. Heat is initially absorbed by the sensible thermal capacity of the heat pipe and liner, and subsequently by the vaporization and discharge of lithium to the atmosphere. The combustor liner temperature is maintained at 3400 F or less during steady-state cruise. The engine inlet leading edge is fabricated as a sodium-superalloy heat pipe. Cooling is accomplished by radiation of heat from the aft surface of the leading edge to the atmosphere. The leading edge temperature is limited to 1700 F or less. It is concluded that heat pipe cooling is a viable method for limiting scramjet combustor liner and engine inlet temperatures to levels at which structural integrity is greatly enhanced.

Silverstein, Calvin C.↗

DOE OSTI2023

High-Fidelity Modeling and Experiments to Inform Safety Analysis Codes for Heat Pipe Microreactors

Heat pipe microreactors are reactor designs that primarily use liquid-metal heat pipes to cool the core. The main interest in heat pipes is the fact that they can remove heat passively. This, along with the use of liquid metal, allows the reactor to operate at higher temperatures. Although the use of heat pipes in nuclear reactors is new, liquid-metal heat pipe technology is mature. Nevertheless, experimental data on heat pipes are scarce, and very little is known about their behavior during abnormal operations and close to their thermal limits. Therefore, new experiments and accurate heat pipe simulations are needed to develop reliable closure models. This work describes a joint experimental and numerical investigation into heat pipes that attempts an initial closure of this gap. Further, the numerical and experimental efforts are currently proceeding in parallel, aimed at different aspects of heat pipes. The numerical part is focused on gaps in local closures, and the experiments capture the overall heat pipe behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

A CONDUCTION-BASED HEAT PIPE MODEL FOR ANALYZING THE ENTIRE PROCESS OF LIQUID-METAL HEAT PIPE STARTUP

The Heat Pipe-cooled Microreactor (HPM) is one of the micro nuclear reactor designs under active study at the U.S. Idaho National Laboratory. Among the major concerns of HPM research is to understand the startup behavior of the heat pipe-cooled system associated with the startup of the liquid-metal heat pipes initialing from frozen state. The startup of liquid-metal heat pipes typically involves a number of nonlinear mass and heat transport processes including the phase change from solid to liquid and vapor. Hence, it is still a huge challenge to simulate the liquid-metal heat pipe startup using conventional CFD methods and software. The major difficulties of numerical CFD modeling come from the phase-change process, multiphase interaction, microporous wick flow, and compressible gas dynamics that occur during startup of the liquid-metal heat pipes. This paper proposes a simplified conduction-based method to provide practical insights into the entire startup process of the liquid-metal heat pipes while mitigating the challenges of addressing all the complex physics. We discuss the theoretical basis and modeling assumptions to analyze the liquid-metal heat pipe startup from frozen state based solely on heat-conduction equations. Then, the proposed model is implemented into the commercial CFD software to verify the model performance. The model prediction results are discussed via the comparison with the experimental data obtained from sodium heat-pipe startup experiments.

42 ENGINEERING↗

NASA NTRS1989

Initial characterization of a modular heat exchanger with an integral heat pipe

As part of the Civil Space Technology Initiative (CSTI) Advanced Technology program, a conceptual design of the Stirling space engine (SSE) was generated. The overall goal of the CSTI high capacity power element is to develop the technology base needed to meet the long duration, high capacity power requirements for future NASA space missions. The free-piston Stirling engine (FPSE) was chosen as the growth option in the CSTI program. A major goal during the conceptual design of the SSE was to reduce the number of critical joints. One area of concern was the heat exchanger assemblies that typically have the majority of critical joints. The solution proposed in the SSE conceptual design used 40 modular heat exchangers. Each module has its own integral heat pipe to transport heat from the heat source to the engine. A demonstration of the modular concept was undertaken before committing to the detailed design of the SSE heat exchangers. An existing FPSE was modified as a test bed for modular heat exchanger evaluation. The engine incorporated three heat exchanger modules, each having a sodium filled heat pipe. The thermal loading of these modules was intended to be similar to the conditions projected for the SSE modules. The engine was assembled and tests are underway. The design and fabrication of the heat exchanger modules and the engine used for these tests were described. Evaluation of the individual heat pipes before installation in the engine is described. The initial test results with the modules in operation on the engine were presented. Future tests involving the engine were outlined.

Schreiber, Jeffrey G.↗

DOE OSTI2025

Development of a conduction-based model for analyzing frozen startup of alkali-metal heat pipes

One key area of interest in heat pipe modeling/simulation is to analyze the startup behavior of the liquid-metal heat pipes (LMHPs) from a frozen state. This so-called ‘frozen startup’ process involves a complex set of nonlinear mass and heat transport phenomena, including phase transitions from solid to liquid and vapor, multiphase interactions, microporous wick flow, and compressible vapor dynamics. The complexity of these processes makes it challenging to simulate LMHP’s frozen startup using conventional numerical methods or commercial computational fluid dynamics (CFD) software. This paper presents a simplified conduction-based modeling approach that can provide practical insights into the entire LMHP frozen startup process, while alleviating the challenges of modeling its complex physics. The theoretical foundation and physical assumptions of the proposed model are based solely on heat-conduction equation, allowing for a more tractable simulation without sacrificing essential physical accuracy. The proposed model was implemented in a commercial CFD software, and its prediction was compared with the experimental data obtained from sodium heat-pipe startup experiments. The comparison highlights the proposed model's ability to capture the transient thermal behavior of LMHP during frozen startup. This study not only validates the conduction-based frozen startup modeling method but also shows its potential as a practical and efficient tool for understanding the startup performance of the LMHP systems.

Microreactor↗

NASA NTRS1992

Heat pipe fatigue test specimen: Metallurgical evaluation

An innovative creep/fatigue test was run to simulate the temperature, mechanical load, and sodium corrosion conditions expected in a heat pipe designed to supply thermal energy to a Stirling cycle power converter. A sodium-charged Inconel 718 heat pipe with a Nickel 200 screen wick was operated for 1090 hr at temperatures between 950 K (1250 F) and 1050 K (1430 F) while being subjected to creep and fatigue loads in a servo-hydraulic testing machine. After testing, the heat pipe was sectioned and examined using optical microscopy, scanning electron microscopy, and electron microprobe analysis with wavelength dispersive x-ray spectroscopy. The analysis concentrated on evaluating topographic, microstructural, and chemical changes in the sodium exposed surfaces of the heat pipe wall and wick. Surface changes in the evaporator, condenser, and adiabatic sections of the heat pipe were examined in an effort to correlate the changes with the expected sodium environment in the heat pipe. This report describes the setup, operating conditions, and analytical results of the sodium heat pipe fatigue test.

Walak, Steven E.↗

DOE OSTI2020

Neutronics Analysis of Cold Critical KRUSTY Experiments using MCNP and Serpent

Kilowatt Reactor Using Stirling Technology (KRUSTY) is a prototype designed as a proof of concept for NASA’s Kilo power program, which was funded in order to create a small reactor for various space applications such as providing power for a colony on Mars or the moon and eventually as a possible option for powering rockets on deep space missions. KRUSTY was tested at the National Criticality Experiments Research Center (NCERC) located at the Nevada National Security Site (NNSS) from November 2017, through March 2018. The experimental campaign was conducted in four phases: com ponent critical experiments, cold critical experiments, warm critical runs, and high temperature demonstration. The component critical and cold critical experiments were used to determine the worth of the beryllium oxide (BeO) reflector rings and the boron carbide (B4C) control rod disks under different conditions. These conditions were altered by the addition and subtraction of other components in the assembly. The component critical phase was performed at atmospheric pressure and had four solid support rods spanning the center core; whereas the cold critical phase had the center core placed in a vacuum chamber and eight sodium-filled heat pipes spanning the center core region. This summary discusses the static neutronic analysis completed on the cold critical phase of the campaign using MCNP ® and Serpent. These two codes were used to assess any potential differences or biases in the simulation results. Understanding these potential biases is a key step toward the final goal of being able to fully analyze the transient experiments performed during phases 3 and 4 of the campaign.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

SPHERE Gas Composition Test and SOCKEYE Validation

The single primary heat extraction and removal emulator (SPHERE) gas composition test was performed to obtain data on the temperature distribution of a 2-m, 0.625-in.-diameter sodium-filled heat pipe from Advanced Cooling Technologies (ACT) in a horizontal orientation filled with varying gas compositions to support NEAMS tool development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Thermal Hydraulic Experimental Test Article (FY2020 Status Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary and secondary system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, Figure 2. The design work for the facility has been completed. Since the FY19 THETA status report, the THETA primary system components have been assembled and tested using de-ionized water. Some results of this testing at select Richardson and Reynolds numbers are included herein. The custom THETA sodium flowmeter (permanent magnet submersible sodium flowmeter) has been built and calibrated using sodium in an auxiliary facility against a NIST traceable vortex shedder type flowmeter. The permanent magnet flowmeter showed excellent performance, following the theoretical equation for measured voltage quite well, with an error < 3.5% at flow rates ranging from 3-24 GPM at temperatures up to 400 °C. The secondary sodium system for heat rejection from intermediate heat exchanger has been analyzed according to ASME code. All secondary side piping, fittings and a U-stamped air-to-sodium heat exchanger have been procured.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Thermal Hydraulic Experimental Test Article - Fiscal Year 2023 (Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that is used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility simulates nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field may be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility was designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary coolant and secondary coolant system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. To date a test matrix has been completed utilizing the primary system of THETA. These tests, along with computational fluid dynamics and systems code models, determined the heat transfer across the core barrel and intermediate heat exchanger outlet was too great to effectively represent scaled thermal hydraulic phenomena of a liquid metal cooled reactor. Therefore, a significant effort was made to remove the primary system from the METL 28” test vessel #4 and clean the residual sodium from the primary system to facilitate upgrades. Thermal insulation was then incorporated in the core barrel and intermediate heat exchanger outlets. The primary system was then replaced, and a series of tests were performed to assess the performance of the thermal insulation. With primary system testing and upgrades complete, the secondary system could then be brought online. The tube side of the shell-and-tube intermediate heat exchanger was installed onto the primary system flange to begin installing the secondary system. The support structure for the secondary system was then erected on the METL mezzanine alongside the THETA primary system to facilitate installation of the secondary system components (sodium-to-air heat exchanger, flowmeter and pump). The piping and expansion tank were welded into the secondary system. Non-destructive examination of the secondary system welds was completed in order to satisfy ASME B31.3 pipe code for class M process fluids. The heating system and insulation were then added to prepare the system to be filled with sodium. The ancillary electrical equipment was installed which included the pump control box, blower VFD, pipe heater control system, etc. The secondary system will be filled, and a test matrix will be completed in early FY2024.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Microreactor Agile Nonnuclear Experimental Testbed Test Plan

Microreactors are an attractive technology option for kick-starting nuclear innovation if they can be operated at high temperature, yielding high power conversion thermal efficiencies comparable or better than in commercial light water reactors. Microreactors are currently the smallest variation of Small Modular Reactors (SMRs). SMRs are “newer generation reactors designed to generate electric power up to 300 MWe and whose components and systems can be shop-fabricated and then transported as modules to the sites for installation as demand arises.” (IAEA, 2016). Vendors are developing microreactor designs to provide an affordable, potentially mobile source of electricity - see Fig. 1 for an example of a microreactor on a semi-truck. Various microreactor designs are possible including heat pipe- and gascooled options, which are the focus of the nonnuclear testing described in this document. In heat pipe microreactors, high-temperature heat pipes using liquid sodium or potassium working fluid transport fission heat from the core to a heat removal section which in turn transfers heat to the power conversion system working fluid. In a gas-cooled design, He or other gas will flow through a solid monolith of material and transfer heat as the temperature of the gas increases through a heat exchanger to a power conversion unit. The logistics of all these processes will be examined and tested through a series of articles at the nonnuclear test bed at Idaho National Laboratory (INL), the Microreactor Agile Nonnuclear Experiment Testbed (MAGNET) facility. Microreactors designed to produce power of 0.1-20 MWt offer the potential for more affordable nuclear energy for a range of applications. In a heat pipe microreactor, heat pipes, fuel rods, and/or moderator are intermixed in the reactor core assembly. Heat pipes extend from the core region into the heat removal section where the power conversion unit working fluid flows through holes or channels, transferring heat from the heat pipes to the working fluid. In a gas-cooled microreactor, gas flows through the solid monolith region and up into the heat exchanger region, transferring heat to the working fluid. For initial testing, the heat removal working fluid can be a low pressure gas for testing that addresses thermal stresses. In the final application, heat addition to the power conversion working fluid typically occurs at high pressure, supporting operation of an air-Brayton, supercritical CO 2 (SCO2), or He-recuperated Brayton cycle. Various stages of the steps above will be demonstrated through the tests described in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Thermal Hydraulic Experimental Test Article - Status Report FY20 Q2

42 ENGINEERING↗

DOE OSTI2021

Thermal Hydraulic Experimental Test Article (FY21 Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary and secondary system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. Figure 1 illustrates the main components of the primary system. THETA has been installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, Figure 2 and Figure 3. Since the FY20 THETA status report, further water shakedown testing was performed, some final welding modifications were performed, all system components were sanitized, dry assembled, and commissioned. The THETA submersible electromagnetic flowmeter was welded closed to hermetically seal internal components from sodium followed by a helium leak check for the final time. Following installation of THETA into vessel #4, all data acquisition and control systems were brought online and commissioned at room temperature in the argon gas space of the empty vessel. The vessel zone 1 heaters were reinstalled and the flange was then insulated. The vessel 4 heaters were brought to a temperature of 100 °C and the THETA pump was jogged at 600 RPM to ensure it was functioning correctly. The secondary, AC Conduction based electromagnetic pump from CMI Novacast (product number CA-15) was ordered and will be delivered in late calendar year 2021. The custom designed permanent magnet based secondary flowmeter was designed, a drawing package and manufacturing specification created and quote requested from a machine shop. The intermediate heat exchanger design was completed and a drawing package is currently in progress. During the middle of calendar year 2021 the Building 308 sodium scrubber system was shut down in order to replaced some corroded piping and facilitate improvements on the system, thus requiring the sodium in METL to remain frozen. This has imposed a delay in filling vessel 4 with sodium to begin THETA testing. Sodium testing is expected to begin in early fiscal year 2022 as all THETA primary systems are operational.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS2005

Hardware Progress Made in the Early Flight Fission Test Facilities (EFF-TF) To Support Near-Term Space Fission Systems

The EFF-TF provides a facility to experimentally evaluate thermal hydraulic issues through the use of highly effective non-nuclear testing. These techniques provide a rapid, more cost effective method of evaluating designs and support development risk mitigation when concerns are associated with non-nuclear aspects of space nuclear systems. For many systems, electrical resistance thermal simulators can be used to closely mimic the heat deposition of the fission process, providing axial and radial profiles. A number of experimental and design programs were underway in 2004. Initial evaluation of the SAFE-100a (19 module stainless steel/sodium heat pipe reactor with integral gas neat exchanger) was performed with tests up to 17.5 kW of input power at core temperatures of 1000 K. A stainless steel sodium SAFE-100 heat pipe module was placed through repeated freeze/thaw cyclic testing accumulating over 200 restarts to a temperature of 1000 K. Additionally, the design of a 37-fuel pin stainless steel pumped sodium/potassium (NaK) loop was finalized and components procured. Ongoing testing at the EFF-TF is geared towards facilitating both research and development necessary to field a near term space nuclear system. Efforts are coordinated with DOE laboratories, industry, universities, and other NASA centers. This paper describes some of the 2004 efforts.

VanDyke, Melissa↗