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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 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Pre-Transient Characterization of MLOF-1 Test Pin

This study focuses on the pre-transient characterization of U-10Zr test and sibling fuel pins for the THOR-M-LOF test series. Using neutron radiography, element contact profilometry (ECP), precise gamma scan (PGS), and gas assay, sampling, and recharge (GASR) analysis, it was confirmed that the fuel pins were intact and suitable for testing. Key fuel behaviors quantified include axial elongation, diametral strain, fluff structure geometry, axial isotope distribution, and fission gas release. Any deviations from historically expected behaviors were investigated and attributed to factors other than the irradiation behavior of the fuel pin. These pre-transient measurements establish a baseline for future post-transient analysis, which will be used to inform fuel performance models and safety criteria for sodium-cooled fast reactors (SFRs). The results will enhance understanding of transient fuel behavior and expand limited data on LOF scenarios.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Status Report for NEET In-Pile Sensor Irradiation Capabilities at TREAT

The recently-established in-pile sensor development project under the Nuclear Energy Enabling Technology (NEET) program is focused on the development of measurement technologies to enable acquisition of data focused on the performance, properties, and boundary conditions of nuclear fuels/materials under irradiation in nuclear reactors. These sensors can be used in steady state irradiation tests to measure parameters of interest under conditions representing normal nuclear plant operation, or in transient irradiation tests to determine specimen performance under postulated unplanned conditions. In either case, development of suitable sensor technologies must include instrument irradiations to help guide sensor material/design selection, verify instrument performance, and demonstrate deployment viability. The Transient Reactor Test facility (TREAT) while just a part of the overall capabilities portfolio needed to enable this sensor development mission, has unique capabilities and characteristics that make it well poised for a number of instrument irradiation data objectives including sensor survivability screening, response characterization, qualification tests, and in-situ properties measurements. TREAT’s brilliantly-basic design, when married with appropriate irradiation hardware and analytic capabilities, fills a crucial capability gap between out-of-pile testing and irradiation in steady state high power research reactors. Leveraging a recently developed approach to irradiation testing in TREAT, the present work demonstrates that cost-effective and scientifically-impactful instrument irradiation capabilities for TREAT are a well-placed capability in the NEET in-pile sensor development project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Testing of graphite thermal neutron scattering law data in support of treat neutronic analysis

The Transient Reactor Test Facility (TREAT) has historically provided the capability for transient fuel performance and safety testing in the United States. The operation of this facility was suspended in 1994. However, in recent years preparations for the restart of TREAT were initiated. TREAT is a graphite moderated and air cooled reactor. Its core is composed of a 19 x 19 array of fuel assemblies that contain UO{sub 2} (HEU) fuel particles dispersed in a graphite matrix, along with reflector assemblies filled with graphite only. Each fuel assembly is 8 feet tall with a fueled region that is 4 feet tall and upper and lower regions of graphite that are 2 feet tall. In addition, a 2 feet thick radial graphite reflector region is present. Finally, a 5 feet thick concrete biological shield surrounds the core. Previous work showed that the graphite thermal scattering cross sections can introduce variations exceeding 2000 pcm in the predicted eigenvalue of the TREAT core. Recognizing that the graphite in the assembly (and thereby the core) is a porous mixture of crystalline and amorphous carbon (with a lower density in comparison to ideal crystalline graphite), it becomes clear that TREAT specific thermal neutron scattering cross section libraries will be needed (at all temperatures of interest) for use within neutronic codes seeking to simulate TREAT. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1995

Transient and Steady-state Tests of the Space Power Research Engine with Resistive and Motor Loads

The NASA Lewis Research Center (LeRC) has been testing free-piston Stirling engine/linear alternators (FPSE/LA) to develop advanced power convertors for space-based electrical power generation. Tests reported herein were performed to evaluate the interaction and transient behavior of FPSE/LA-based power systems with typical user loads. Both resistive and small induction motor loads were tested with the space power research engine (SPRE) power system. Tests showed that the control system could maintain constant long term voltage and stable periodic operation over a large range of engine operating parameters and loads. Modest resistive load changes were shown to cause relatively large voltage and, therefore, piston and displacer amplitude excursions. Starting a typical small induction motor was shown to cause large and, in some cases, deleterious voltage transients. The tests identified the need for more effective controls, if FPSE/LAs are to be used for stand-alone power systems. The tests also generated a large body of transient dynamic data useful for analysis code validation.

Rauch, Jeffrey S.↗
NASA NTRS1980

Transient stress intensity factors for edge and corner cracks in quench-test specimens

The transient temperature and stress fields in a quenched rectangular bar were computed and used to determine the stress intensity factors for midside and corner-edge cracks. The stress intensity factors, KI, are presented as a function of time, flaw size, and aspect ratio. The variations of KI along the crack edge are given and the self-limiting growth of the crack depth is discussed in relation to the nonlimited surface growth.

Emery, A. F.↗
DOE OSTI2019

Suitability of Type-II Fiber Bragg Gratings in Silica Optical Fiber for Temperature Sensing in TREAT

Optical fibers provide a variety of options for instrumentation in reactor environments. Optical fibers can be used to measure multiple physical phenomena including, temperature, strain, pressure, and fluid level. In addition to the various sensing applications, optical fibers are immune to electromagnetic interference, have a small footprint (~100 μm), and a fast response. The Department of Energy and Idaho National Laboratory have considered optical fibers for use as in-pile instrumentation in the Transient Reactor Test Facility (TREAT). TREAT was designed to test reactor fuels under accident conditions by replicating accident conditions for a variety of reactor transients, such as those associated with a loss of coolant accident (LOCA). This work investigates silica fiber optic temperature sensors with inscribed type-II fiber Bragg gratings (FBGs) under conditions similar to those that would be experienced in a TREAT transient. Separate effects testing was used to test the sensors under high-temperature step transients and under irradiation up to a total fluence similar to that of TREAT. Specifically, this work investigates distributed temperature measurements, using the Optical Frequency Domain Reflectometry (OFDR) sensing technique, using a Luna Innovations Optical Backscatter Reflectometer (OBR) 4600, with silica optical fibers inscribed with type-II fiber Bragg gratings (FBGs). In conclusion, separate effects testing of type-II FBGs in silica optical fiber, to high temperature and to neutron fluences that are an order of magnitude larger than fluences that are anticipated for TREAT tests, demonstrate that type-II FBGs in silica optical fiber hold great promise for high-temperature reactor instrumentation in TREAT.

McCary, Kelly↗
DOE OSTI2016

Time Constants for Optical Fiber Temperature Sensing using a Thermal Resistance Approach

Optical fibers are immune to electromagnetic interference, have no moving parts, no applied voltages, no special environmental conditions, and they are extremely small (100's of microns in diameter). The U.S. Department of Energy is interested in extending this technology to extremely high temperature radiation environments for the purposes of developing in-pile instrumentation for the restart of TREAT (Transient Reactor Test Facility). TREAT is an air-cooled, thermal, heterogeneous test reactor facility that was designed to evaluate reactor fuels and structural materials under conditions that simulate various types of transient overpower and under-cooling conditions that may be postulated to occur in a nuclear reactor. Spatially distributed temperature measurements can be made using an Optical Backscatter Reflectometer (such as, model 4600, manufactured by Luna Innovations, Inc.), coupled to single-mode optical fiber that is placed in an in-pile nuclear reactor testing facility. In-pile temperature sensing for the TREAT reactor will need to have a fast response to temperature changes, since the reactor pulse will be on the order of 100 msec. The sensors must also be robust enough to be installed in the testing facility; this requires the fibers to be in a protective capillary tubing that is made of a tough material, such as stainless steel. The temperature response of the fiber will be delayed by the capillary tubing. This paper discusses the characterization and modeling of the thermal time constants of the sensors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Forward and inverse predictive transient models of TREAT using surrogate reactivity models

In this work, we present two novel approaches for predicting the power evolution and control rod height of the Transient Reactor Test Facility (TREAT) to support experiment modeling; specifically, transient analysis of the NASA-sponsored Sirius series of experiments. These approaches utilize steady-state Monte Carlo model, point kinetics model, and surrogate models to predict power evolution and control rod axial position during transient experiments. Both approaches were tested and validated against several Sirius experiments that were performed in the TREAT facility at different power levels. The validation test results show very good agreement with the experimental data, and the models were able to accurately predict the power evolution and the axial control rod position with an average error within 3.0%. Finally, this indicates that these approaches will help the reactor engineering team of the TREAT facility in preparing and predicting the power and temperature of the experiment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

SPC-71260 Rev 0 MARVEL Heat Extraction Subsystem Secondary Coolant Equipment (SCE) Design/Build

A. The Microreactor Applications Research Validation and Evaluation (MARVEL) reactor will offer experimental capabilities that are not currently available at DOE’s national laboratories. Idaho National Laboratory (INL), operated for the U.S. Department of Energy (DOE) by Battelle Energy Alliance, LLC (BEA) (Contractor hereafter) is procuring services for the design, analysis, fabrication, testing and delivery of a Secondary Coolant Equipment system (SCE). This specification contains the requirements for design, analysis, fabrication, testing and delivery of the SCE as described herein. The MARVEL reactor is a microreactor which uses eutectic sodium-potassium alloy (NaK) as a primary coolant. The primary coolant is circulated through four primary loops by natural convection of the coolant. In each loop is a closed well which will accommodate an intermediate heat exchanger (IHX) for extracting heat from the loop. These wells will be referred to in this specification as the “IHX wells.” It is intended for the IHX containment to also be filled with NaK. The MARVEL design team has determined that a Heat Extraction System (HES) using pumped NaK will be used to extract heat from the IHXs and deliver it to a downstream system for power generation or alternate process heat users. This Heat Extraction System will enable MARVEL operations including the ability to test, demonstrate, and address issues related to installation, startup, and operations. In addition, it will allow down-stream utilization of process heat for various uses. The objective of this specification is to develop the final design for the HES Secondary Coolant Equipment system (SCE) that will be used as the core of the HES. This system provides control of the NaK circulation between the MARVEL reactor and the subsequent process heat utilization systems. It does not include design of the Intermediate Heat Exchangers and piping inside the T-REXc pit in which the reactor is located. B. The MARVEL microreactor will be installed in the Transient Reactor Test Facility (TREAT) building in the Transient Reactor Test (TREAT) Micro-Reactor Experiment Cell (T-REXc) C. An INL Subcontractor has developed a conceptual design for this system per SPC-71145, referred to in that specification as the Process Heat Extraction System. SPC-71260 is based on the pumped NaK loop concept developed under SPC-71145. D. The SCE system design and (as option scope) fabrication shall be provided by the awardee of the subcontract (Subcontractor hereafter) pertaining to this Specification. Prior to shipment, the SCE will be fabricated, assembled, and tested at the Subcontractor’s facility. After successful completion of acceptance testing, the SCE and associated equipment will be shipped to the Materials and Fuels Complex (MFC) at the INL (Contractor’s Facility hereafter) to be installed by others in TREAT/T-REXc.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2018

An advanced experimental design for modified burst testing of nuclear fuel cladding materials during transient loading

Here an advanced tube-burst test system was designed to test samples of nuclear fuel cladding under conditions relevant to a postulated design-basis reactivity insertion accident (RIA) in light-water reactors. The system is based on the “driver tube” concept and allows for high-speed testing with internal pressure impulses of 10–1000 ms. To measure strain, the system was equipped with an ultra-high-speed video camera with a telecentric lens providing high focal depth, to compensate for specimen shift. A mirror system was developed to provide 360° view of the tube specimen into a single camera. Images taken during the test allow the use of a digital image correlation approach in strain evaluation. Several experiments were conducted to assess the performance of the experimental setup, and results are reported.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

OPTI-TWIST Test Report, FY2025

OPTI-TWIST is the out-of-pile testing prototype and instrumentation TWIST with analogous components without nuclear material. Fabrication, construction, and testing in FY2025 is described in this test report.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Code Benchmark of Depressurized Conduction Cooldown Transient in the High Temperature Test Facility

This paper presents results from modeling of a depressurized conduction cooldown (DCC) transient at the High Temperature Test Facility (HTTF) as part of the OECD-NEA Thermal Hydraulics Code Validation Benchmark for High-Temperature Gas-Cooled Reactors using HTTF Data . This paper briefly describes the benchmark and the models being used. It then presents a comparison of steady state and transient results based on the Problem 2 Exercise 1A and 1B definitions. We compare block and helium temperature distributions, mass flow distribution, and energy balance in steady state. All models show comparable mass flow distributions and energy balances. The temperatures within the core and outer regions are comparable in all models too, but inner reflector temperatures can vary significantly. Despite that, we find that the models are in good agreement for the full-power steady state. In the DCC, we look at block temperature at the core midplane and RCCS water exit temperature. The INL and ANL models are found to be in excellent agreement with one another on block temperature over time, while the agreement when the KAERI and NRG models are added into consideration is good. Differences in the transient heat removal from the RCCS cause the differences in block temperature over time in these models. The CNL models show similar trends to the INL, ANL, KAERI, and NRG models, but the temperatures are high because the volumes used in calculating the average temperature include the heater rods in the CNL models only. The HUN-REN model shows results that suggest significantly lower heat removal in the RCCS which merit further investigation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

FOR-868 Microreactor Applications Research Validation and Evaluation (MARVEL) Project

The Microreactor Applications Research Validation and Evaluation (MARVEL) reactor will offer experimental capabilities that are not currently available at DOE’s national laboratories. The test bed will perform R&D on the fundamental features, operations, and behaviors of microreactor technologies and help industry partners quickly test, develop, and demonstrate their technologies. The liquid- metal cooled microreactor will produce energy using small amounts of high- assay, low-enriched uranium (HALEU) from available research materials. Its design is primarily based on existing technology and will be built using off-the- shelf components to the extent practical. The reactor will be built inside the Transient Reactor Test (TREAT) Facility Micro-Reactor Experiment Cell (T- REXC), a separate project governed by the requirements of FOR-684, “Transient Reactor Test (TREAT) Facility Micro-Reactor Experiment Cell (T REXC),” and outside the scope of this document.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2007

Testing in Support of Space Fission System Development and Qualification

Extensive data would be required for the qualification of a fission surface power (FSP) system. The strategy for qualifying a FSP system could have a significant programmatic impact. This paper explores potential options that could be used for qualifying FSP systems, including cost-effective means for obtaining required data. three methods for obtaining qualification data are analysis, non-nuclear testing, and nuclear testing. It has been over 40 years since the US qualified a space reactor for launch. During that time, advances have been made related to all three methods. Perhaps the greatest advancement has occurred in the area of computational tools for design and analysis. Tools that have been developed, coupled with modem computers, would have a significant impact on a FSP qualification. This would be especially true for systems with materials and fuels operating well within temperature, irradiation damage, and burnup limits. The ability to perform highly realistic non-nuclear testing has also advanced throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modem FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.} and extensive data to be taken from the core region. Both steady-state and transient operation can be tested. For transient testing, reactivity feedback is calculated (or measured in cold/warm criticals) based on reactor temperature and/or dimensional changes. Pin power during a transient is then calculated based on the reactivity feedback that would occur given measured values of temperature and/or dimensional change. In this way nonnuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. Realistic non-nuclear testing is most useful for systems operating within known temperature, irradiation damage, and burnup capabilities.

Houts, Mike↗