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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 19 records
All sourcesNASA NTRSDOE OSTI
NASA NTRS2007

Analysis and Down Select of Flow Passages for Thermal Hydraulic Testing of a SNAP Derived Reactor

As past of the Vision for Space Exploration, man will return to the moon. To enable safe and productive time on the lunar surface will require adequate power resources. To provide the needed power and to give mission planners all landing site possibilities, including a permanently dark crater, a nuclear reactor provides the most options. Designed to be l00kWt providing approx. 25kWe this power plants would be very effective in delivering dependable, site non-specific power to crews or robotic missions on the lunar surface. An affordable reference reactor based upon the successful SNAP program of the 1960's and early 1970's has been designed by Los Alamos National Laboratory that will meet such a requirement. Considering current funding, environmental, and schedule limitations this lunar surface power reactor will be tested using non-nuclear simulators to simulate the heat from fission reactions. Currently a 25kWe surface power SNAP derivative reactor is in the early process of design and testing with collaboration between Los Alamos National Laboratory, Idaho National Laboratory, Glenn Research Center, Marshall Space Flight Center, and Sandia National Laboratory to ensure that this new design is affordable and can be tested using non-nuclear methods as have proven so effective in the past. This paper will discuss the study and down selection of a flow passage concept for a approx. 25kWe lunar surface power reactor. Several different flow passages designs were evaluated using computational fluid dynamics to determine pressure drop and a structural assessment to consider thermal and stress of the passage walls. The reactor design basis conditions are discussed followed by passage problem setup and results for each concept. A recommendation for passage design is made with rationale for selection.

Godfroy, T. J.↗
NASA NTRS2007

Assembly and Thermal Hydraulic Test of a Stainless Steel Sodium-Potassium Circuit

Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the NASA Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system was originally built for use with lithium, but due to a shift in focus, it was redesigned for use with a eutectic mixture of sodium potassium (NaK). Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a full design) was selected for fabrication and test. This paper summarizes the first fill and checkout testing of the Stainless Steel NaK-Cooled Circuit (SNaKC).

Garber, A.↗
DOE OSTI2025

Thermal Hydraulic Experimental Test Article: Second Year of Testing with Secondary Sodium System (Fiscal Year 2025 Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is currently installed in the Mechanisms Engineering Test Loop (METL) 28” test vessel #4. Both the primary and secondary sodium systems remain online to facilitate continued testing. This fiscal year, work was performed using a COMSOL Multiphysics magnetohydrodynamic model to characterize flow more accurately in the secondary electromagnetic flowmeters. Experimental campaigns were then performed to study the thermal hydraulic differences between sodium and water as a surrogate fluid in the THETA geometry as well as a study to better characterize and understand temperature oscillations that exist at the outlet of the core to the hot pool. A peer-reviewed article was published in the ASME Journal of Nuclear Engineering and Radiation Science detailing the THETA facility and providing an overview of a test that was performed with the primary and secondary system online [1]. Work continues to develop a database to house experimental THETA data to better facilitate collaboration with industry and laboratory partners for their use of the data for code benchmarking/validation. THETA remains fully operational and is positioned for continued testing in fiscal year 2026.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1999

Results of a First Generation Propellant Energy Source Module Testing: Non-Nuclear Testing of Fission System

The use of resistance heaters to simulate heat from fission allows extensive development of fission systems to be performed in non-nuclear test facilities, saving time and money. Resistance heated tests on the Module Unfueled Thermal- hydraulic Test (MUTT) article has been performed at the Marshall Space Flight Center. This paper discusses the results of these experiments to date, and describes the additional testing that will be performed. Recommendations related to the design of testable space fission power and propulsion systems are made.

VanDyke, Melissa↗
DOE OSTI2024

Development of an Improved RELAP5-3D Model for the High Temperature Test Facility

High-temperature gas-cooled reactors (HTGRs) are rapidly approaching deployment. Confidence in transient analysis of these systems for design, optimization, and licensing calculations requires modeling and simulation tools that have been validated against data relevant to HTGR conditions. The High Temperature Test Facility (HTTF) is an integral effects thermal hydraulics test facility for prismatic HTGRs. In spring and summer of 2019, HTTF was used for a series of experiments that now serve as the basis for the OECD/NEA Thermal Hydraulic Code Validation Benchmark for High Temperature Gas-Cooled Reactors using HTTF Data (HTGR T/H Benchmark). This benchmark contains problems for systems code, computational fluid dynamics (CFD), and coupled systems code/CFD modeling representing lower plenum mixing and both the depressurized and pressurized conduction cooldown (DCC and PCC respectively) transients. Benchmark problems include exercises for code-to-code and code-to-data comparisons as well as an exercise for error scaling between HTTF and the Modular High Temperature Gas-Cooled Reactor, which serves as the basis for the HTTF design. Previous analysis as part of the HTGR T/H benchmark used a RELAP5-3D model developed at Idaho National Laboratory (INL) and demonstrated an ability to reproduce trends in the measured data but difficulties reproducing experimental values within their uncertainty. These difficulties were largely attributed to assumptions made during the development of the initial RELAP5-3D model, which predated the HTTF experiments. A significant cause of difficulty reproducing the measured temperatures may be the radial nodalization of the previous RELAP5-3D model. The new model provides a finer nodalization to assess the impact of radial nodalization and allows for asymmetric heating within the core, which was a feature of multiple HTTF experiments. In this paper, we present the new RELAP5-3D model of HTTF. In addition to describing the new model, this paper compares the new and old models and provides results for a full-power steady state, a DCC, and a PCC in HTTF. These analyses are based on the code-to-code comparison exercises for the DCC and PCC problems of the HTGR T/H benchmark. We present the results of these exercises from the new model and compare them to the results of the old model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Development of an Improved RELAP5-3D Model for the High Temperature Test Facility

High-temperature gas-cooled reactors (HTGRs) are rapidly approaching deployment. Confidence in transient analysis of these systems requires modeling and simulation tools that have been validated against data relevant to HTGR conditions. The High Temperature Test Facility (HTTF) is an integral effects thermal hydraulics test facility for prismatic HTGRs. In spring and summer of 2019, HTTF was used for a series of experiments that now serve as the basis for the Organization of Economic Cooperation and Development / Nuclear Energy Agency Thermal Hydraulic Code Validation Benchmark for High Temperature Gas-Cooled Reactors using HTTF Data (HTGR T/H Benchmark). Previous analyses as part of the HTGR T/H benchmark used a RELAP5-3D model developed at Idaho National Laboratory (INL) and demonstrated an ability to reproduce trends in the measured data but difficulties reproducing experimental values within their uncertainty. These difficulties were largely attributed to assumptions made during the development of the initial RELAP5-3D model, which predated the HTTF experiments. A significant cause of difficulty reproducing the measured temperatures may be the radial nodalization of the previous RELAP5-3D model. In this paper, we present a new RELAP5-3D model of HTTF with finer radial nodalization built to assess the impact of radial heat transfer. We describe the new model and compare it against the old one at full-power steady state and for the pressurized conduction cooldown (PCC) transient. These analyses are based on the code-to-code comparison exercise for the PCC problem of the HTGR T/H benchmark. We compare maximum block temperature as the primary figure of merit and include discussion on intracore natural circulation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2000

Development and Results of a First Generation Least Expensive Approach to Fission: Module Tests and Results

The use of resistance heaters to simulate heat from fission allows extensive development of fission systems to be performed in non-nuclear test facilities, saving time and money. Resistance heated tests on the Module Unfueled Thermal-hydraulic Test (MUTT) article has been performed at the Marshall Space Flight Center. This paper discusses the results of these experiments and identifies future tests to be performed.

Houts, Mike↗
DOE OSTI2025

Proposed New STELLA-2 Tests

In 2025, KAERI and Argonne initiated a collaboration based on software validation using data from tests performed at KAERI’s STELLA-2 facility. STELLA-2 is a large-scale sodium thermal-hydraulic test facility that was originally developed to support KAERI’s development of the PGSFR design concept. KAERI has performed a large number of tests at STELLA-2, providing valuable data to support sodium fast reactor code validation. KAERI will be conducting new tests in 2026, targeting system conditions that were not achieved during previous test campaigns. KAERI has requested from Argonne proposed tests that could be performed during the upcoming test campaign. This report documents Argonne’s proposed tests in fulfillment of KAERI’s request.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Flow instabilities in helical-coil steam generators for small modular reactors: A review

Here, this study covers the research and discoveries in two-phase flow-boiling instabilities available in the literature—specifically for a helical-coil steam generator (HCSG), including experimental findings, theoretical research, computational models, and system code analyses—supporting research and development of representative small modular reactors (SMRs). Like other new and advanced reactor systems, water-cooled SMRs require experimental data from both integral and separate thermal-hydraulics test facilities for the verification and validation (V&V) of the computational models and computer codes in order to design and obtain regulatory approval. The complex dynamics of two-phase flow-boiling instabilities includes flow regimes physics phenomena, flow-channel geometries, heat-transfer behavior, and interactions among the solid–liquid-gas within the system boundary, all of which are pivotal for understanding the design and operational challenges of SMRs. This study focuses on identifying the relevant knowledge gaps on boiling instabilities—specifically for a HCSG—and provides insights about future research direction optimizing the transport of thermal energy, mass-flow rates, and boundary conditions that ensure the adequate heat-transfer performance, operational stability, and safety associated with SMR systems.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2025

Mechanisms Engineering Test Loop (METL) Operations, Maintenance, and Testing - FY2025

This report documents the operations, maintenance, and improvements that were performed at the Mechanisms Engineering Test Loop (METL) and its supporting infrastructure during FY2024. The METL facility had a very successful seventh year of operations while supporting the testing of multiple test article experiments in the facility. The METL facility continued supporting the Gear Test Assembly (GTA) testing and the Thermal Hydraulic Experimental Test Article (THETA) with the full testing with both the primary and secondary systems. Work to accommodate two additional experiments, a flow sensor test article (F-STAr) gripper test and a fuel handling gripper test article (GrTA) continued as they are expected to undergo testing in METL in FY2025. In addition, a new 18” test article, the Sample Testing Basket (STB) was used a few times to provide screening tests for sodium service materials. A fifth test vessel was installed in the location of Test Vessel 6, and a wet vapor nitrogen sodium processing system was developed and initially tested.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2007

High Fidelity Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Results

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronie response provides a bridge between electrically heated testing and fueled nuclear testing, providing a better assessment of system integration issues, characterization of integrated system response times and response characteristics, and assessment of potential design improvements' at a relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design can developed. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Bragg-Sitton, Shannon M.↗
NASA NTRS2007

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Bragg-Sitton, Shannon M.↗
NASA NTRS2010

Non-Nuclear Testing of Space Nuclear Systems at NASA MSFC

Highly realistic non-nuclear testing can be used to investigate and resolve potential issues with space nuclear power and propulsion systems. Non-nuclear testing is particularly useful for systems designed with fuels and materials operating within their demonstrated nuclear performance envelope. Non-nuclear testing allows thermal hydraulic, heat transfer, structural, integration, safety, operational, performance, and other potential issues to be investigated and resolved with a greater degree of flexibility and at reduced cost and schedule compared to nuclear testing. The primary limit of non-nuclear testing is that nuclear characteristics and potential nuclear issues cannot be directly investigated. However, non-nuclear testing can be used to augment the potential benefit from any nuclear testing that may be required for space nuclear system design and development. This paper describes previous and ongoing non-nuclear testing related to space nuclear systems at NASA's Marshall Space Flight Center (MSFC).

Houts, Michael G.↗
NASA NTRS2011

Non-Nuclear Testing of Fission Technologies at NASA MSFC

Highly realistic non-nuclear testing can be used to investigate and resolve potential issues with space nuclear power and propulsion systems. Non-nuclear testing is particularly useful for systems designed with fuels and materials operating within their demonstrated nuclear performance envelope. Non-nuclear testing also provides an excellent way for screening potential advanced fuels and materials prior to nuclear testing, and for investigating innovative geometries and operating regimes. Non-nuclear testing allows thermal hydraulic, heat transfer, structural, integration, safety, operational, performance, and other potential issues to be investigated and resolved with a greater degree of flexibility and at reduced cost and schedule compared to nuclear testing. The primary limit of non-nuclear testing is that nuclear characteristics and potential nuclear issues cannot be directly investigated. However, non-nuclear testing can be used to augment the potential benefit from any nuclear testing that may be required for space nuclear system design and development. This paper describes previous and ongoing non-nuclear testing related to space nuclear systems at NASA s Marshall Space Flight Center (MSFC).

Houts, Robert G.↗
NASA NTRS2006

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

Bragg-Sitton, Shannon M.↗
NASA NTRS2007

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

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

Bragg-Sitton, S. M.↗
NASA NTRS2020

Europa Clipper Passive Thermal Control Valve Test and Analysis

Europa Clipper spacecraft utilizes passive thermal control valves (TCV) in an active thermal control system to maintain spacecraft component temperatures within their allowable limits. The single-phase mechanically pumped fluid loop recovers heat from the electronics and uses it to maintain propulsion system warm. When excess heat is dissipated the mixing valves direct the fluid to the radiator to reject this additional heat. The loop uses two mixing TCVs placed in series to modulate fluid flow and heat rejection rate from the radiator. This paper describes the thermal and hydraulic test results for Europa Clipper engineering model (EM) mixing valves. For the first time two heritage mixing valves previously developed for the Mars Science Laboratory (MSL) were placed in series to modulate flow to the radiator from 0.1% to 95% of full flow between temperature range of 4 oC to 24 oC. Individually, each mixing valve was able to modulate the flow from 2% to 96% of full flow between the temperatures of 4 oC and 24 oC. Pressure drop was measured between 1.5 to 3 psid depending on the temperature of the mixed fluid and the opening fraction of the valves. A methodology was developed, tested, and validated to adjust the setpoints with an accuracy of 0.1 oC. The measured flow splits were corrected for the effects of impedance due to the flow meters used to measure the flow rates. Finally, flow rate to the radiator increased linearly as a function of mixed temperature for individual valves, while it increased parabolically with temperature when two mixing valves were placed in series.

Mastropietro, AJ↗