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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Thermal-Hydraulic Performance of the TREAT Multi-SERTTA for Reactivity Initiated Accident Experiments

Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operating pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

A robust alloy design (RAD) strategy for next-generation (IV) nuclear fission reactors

Next-generation nuclear reactors demand structural materials capable of withstanding extreme conditions, including high temperatures, intense neutron flux, and corrosive environments. Multi-Principal Element Alloys (MPEAs) have emerged as promising candidates due to their exceptional radiation tolerance, thermal stability, and compositional flexibility. This study introduces a versatile and customizable Robust Alloy Design (RAD) strategy for systematically designing MPEAs for GEN-IV reactor fuel cladding. The RAD framework integrates nuclear-relevant selection criteria, empirical parameter assessments, and high-throughput CALPHAD simulations to efficiently narrow compositional space and identify stable alloys. A unified RAD score developed for the first time, combines key performance metrics, including fuel-clad chemical interaction (FCCI), neutron absorption cross-section (NAC), valence electron configuration (VEC), and melting point factor (MPF), into a flexible ranking system adaptable to reactor-specific priorities. Among 724 candidates, V555(5Al–5Cr–5Fe–85V) emerged as the top alloy, validated experimentally with a homogeneous single-phase BCC microstructure and superior mechanical properties (nano-indentation: 3.389 ± 0.258 GPa; Vickers hardness: 240 ± 6.7 HV), significantly outperforming Zircaloy-4 and V-4Cr-4Ti. Importantly, the RAD strategy is not limited to nuclear applications; its customizable weighting system enables scalability to other extreme environments. This adaptability positions RAD strategy as a versatile tool for advanced materials design across multiple industries.

Alloy design↗
DOE OSTI2022

Mechanical response and microstructure evolution from multiaxial stress relaxation in textured zircaloy nuclear cladding

Modern energy markets provide motivation for nuclear reactors to increase operational flexibility and load following capacity of existing reactor designs. The operational power changes result in fuel pellet expansion and contraction resulting in pellet-cladding mechanical interaction (PCMI) and high cladding stresses that are strongly related to stressed corrosion cracking (SCC) which can lead to fuel rod failure. This work aims to quantify the stress fields in cladding with a focus on multiaxial stress relaxation response from an imposed strain and the resulting microstructure changes in textured zircaloy claddings. Stress relaxation tests resulting in multiaxial stress fields are performed to measure the material response of Zircaloy-4 nuclear cladding. Multiple experiments using samples from the same piece of cladding are performed so samples with the same processing history may be removed at strategic points throughout mechanical testing for microstructure characterization. The microstructure evolution is analyzed with mechanical response to evaluate the static recovery impact on crystallographic texture and grain morphology. Experiments are performed at Idaho National Laboratory using an experimental device capable of applying internal pressure and independent axial load to commercial nuclear cladding at elevated temperatures with stress- or strain-controlled experimental capability. X-ray diffraction and electron backscatter diffraction are used for microstructure analysis. Rheological models are used to visualize the multiaxial stress-strain relationships and isolate elastic, plastic, and viscoplastic material properties for analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1975

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

Bowles, K. J.↗
DOE OSTI2018

Design and Thermal Analysis for Irradiation of Silicon Carbide Joint Specimens in the High Flux Isotope Reactor

This report provides a summary of the irradiation vehicle design and thermal analysis of SiC joint specimens planned for irradiation in the flux trap of the High Flux Isotope Reactor (HFIR). Two different capsule designs will be used to accommodate the two different specimen geometries: a small torsion joint specimen geometry to measure mechanical and thermal properties, and a tube and end plug joint representative cladding geometry to demonstrate strength and hermeticity retention. The capsule designs, with target temperatures of 350° and 750°C, will accommodate either sixteen torsion joint specimens or one tube end plug joint specimen. Three joint variations will be studied in each capsule design: a hybrid SiC (preceramic polymer with chemical vapor deposition (CVD) SiC), a transient eutectic phase (TEP) process, and an oxide process.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Hydride Microstructure at the Metal-Oxide Interface of a Zircaloy-4 Fuel Clad from the H.B. Robinson Nuclear Reactor

During nuclear reactor operation, some of the hydrogen that is produced by the water side corrosion is uptaken by cladding. When the uptaken hydrogen content exceeds its solubility limit in the cladding at a specific temperature, it precipitates as hydride particles in the cladding. This process occurs immediately at the metal-oxide interface forming a hydride rim because of the temperature gradient in the cladding. The presence of hydride rim and its microstructure severely reduces the cladding strength and can induce crack initiation and propagation during design basis accidents, transportation, and wet/dry storage. Because hydrides platelets are one of the potent crack initiation sites, microstructural characterization of hydride rim becomes essential to the safe operation of nuclear reactor and spent nuclear fuel retrievability. Therefore, we investigate the hydride rim microstructure at the metal oxide interface using scanning transmission electron microscopy (S/TEM) with Energy Dispersive X-ray Spectroscopy (EDS) that allows for the microstructural and local chemistry characterization of hydride precipitates to be determined, and enables a relationship to be derived on how the precipitate sites are affected by the oxide and their complex stacking behavior of the precipitates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Developing a Macro-scale SiC-cladding Behavior Model Based on Localized Mechanical and Thermal Property Evaluation on Pre- and Post-Irradiation SiC-SiC Composites (Final Report)

Silicon carbide is being investigated for accident tolerant fuel cladding applications due to its good neutronic performance, high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to Zircaloy under accident conditions. The development and investigation of these materials is particularly important in the light of the Fukushima event and subsequent emphasis in DOE on accident tolerant fuel (ATF) concepts. In this work, we propose to develop and improve upon existing small-scale mechanical and thermal characterization methods, including micro-cantilever bend tests and fiber push-out tests, and time-domain thermo-reflectance measurements. These techniques will be applied to evaluate micro-scale properties of SiC-SiC composite constituents (matrix, fibers, interphase). The results will be coupled with meso-scale fiber and void structural information and be used as input to develop a comprehensive, finite element-based model based on constituent properties. The goal of this multi-scale model is to be used to predict anisotropic mechanical and thermal cladding behaviors. The model will be benchmarked against measurements of bulk properties and validated for radiation-damaged materials by utilizing irradiated SiC-SiC composites available from General Atomics. The characterization approach developed under this program will have particular application in the property prediction of reactor irradiated materials, as only small volumes need to be tested, minimizing associated hazards and costs. The micro-scale test methods developed here will be used along with meso-scale structural data to inform a finite element model of the full SiC-based cladding system. This model will capture fabrication and irradiation effects and importantly include for the first-time separate material data for each system component. The incorporation of these micro-scale effects will lead to a more accurate model of SiC-SiC composite behavior, enabling further advancements in the development and design of SiC-based materials for improved fuel cladding performance. The aim of this work is to develop advanced localized material characterization techniques to directly measure mechanical and thermal properties of the individual constituents of SiC-based claddings at the relevant micro-scale. SiC-SiC composites will be evaluated before and after irradiation and these results will be coupled with macro-scale properties and microstructural information in order to provide the input parameters for a comprehensive finite element model, which will be developed in this program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Off-Gas Design and Testing from Advanced Chlorination Process

Chlorination of Zr alloy cladding with sulfur-chloride reagents is being considered as a means of separating the main metal from alloying elements, imbedded U, actinides, and fission and activation products to purify the Zr. This would allow the Zr to be disposed of as low-level nuclear waste or possibly be recycled. A multistage process was developed that involves several steps and waste streams. Recycling of sulfur-chloride reagents S 2 Cl 2 and SOCl 2 is planned to minimize the waste stream. The generation of off-gas and entrainment of volatiles is anticipated to be minimized. However, given the complex nature of spent nuclear fuel (SNF), some volatile components will be generated, including the solvents themselves; Cl 2 used for the regeneration of S 2 Cl 2 ; volatile fission product chlorides, such as Sb and 3 H; and chlorinated alloying components, such as Sn, Nb, Fe, and Mo. This report discusses the chemical phenomena that might give rise to volatile species and their planned capture. The volatility of the transition metals will depend on their oxidation state and the compounds that are formed, such as Zr tritide. Loading radionuclides in the process off-gas will depend on the inputs to the process. For instance, processing separated Zr to reduce waste will have a much lower loading of radionuclides than if alternative chlorination were used for decladding SNF. Much of the volatile process streams will be part of the solvent recycling, which may involve online scrubbing. Generally, a caustic scrubber should be included in the off-gas to remove acidic vapors (e.g., Cl 2 , SOCl 2 ). The decontamination procedures for removing these species could involve pretreatment (i.e., heating) options or reduction by contact with H 2 . This report discusses these options along with a framework for testing these alternatives in future experiments. Technologies for off-gas capture will leverage those developed in the Materials Recovery and Waste Forms program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Post-irradiation Examinations of Accident Tolerant Fuel Candidates with Iron-alloy Cladding

A collaborative effort of the Advanced Fuels Campaign together with industry consortia is focused on the development of enhanced Accident Tolerant Fuels (ATF) for Light Water Reactors (LWRs). The scoping studies, referred to as ATF 1 irradiations, are being performed in the Advanced Test Reactor of Idaho National Laboratory using drop-in style irradiations. The Post Irradiation Examinations (PIEs) of ATF-1 began in 2017 and continued in 2019. This report describes the results of examinations on ATF-1 irradiations designed to test the performance of iron-based cladding. Two rodlets with commercial iron-based cladding and UO2 pellets have been irradiated to low burnup (i.e., < 20 GWd/tHM). The first rodlet had an Alloy33® cladding, while the other had a Kanthal APMT® cladding. The performed PIEs were focused on the determination of fuel microstructure, evaluation of fuel cladding interaction and irradiation-induced variations of cladding mechanical properties. For both rodlets, no significant changes in the cladding hardness were measured, and the cladding hoop strain remained limited. Gamma tomography revealed no Cs axial migration as expected for LWR fuel, but significant cesium radial migration, which was induced by thermal migration consequent to high irradiation temperatures compared to standard LWR irradiation temperatures. Microstructure analyses revealed formation of secondary phases along the pellet rim of samples from both rodlets. No wastage or extensive chemical interaction between the fuel and the internal wall of the cladding was observed, suggesting that the secondary phases accumulated at the pellet rim might be due to formation of Cs-U-O compounds. However, further advanced PIEs are necessary to confirm this hypothesis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Scalable waveguide design for three-level operation in Neodymium doped fiber laser

We have constructed a double clad neodymium doped fiber laser operating on the three-level 4 F 3/2 → 4 I 9/2 transition. The laser has produced 11.5 W at 925 nm with 55% slope efficiency when pumped at 808 nm, comparable to the best previous results for a double-clad fiber configuration on this transition. Higher power pumping with both 808 nm and 880 nm sources resulted in an output of 27 W, albeit at lower slope efficiency. In both cases, output power was limited by available pump, indicating the potential for further power scaling. To suppress the stronger four-level 4 F 3/2 → 4 I 11/2 transition we developed a waveguide that provides spectral filtering distributed along the length of the fiber, based on an all-solid micro-structured optical fiber design, with resonant inclusions creating a leakage path to the cladding. Furthermore, the waveguide supports large mode areas and provides strong suppression at selectable wavelength bands, thus easing the restrictions on core and cladding sizes that limited power scaling of previous approaches.

42 ENGINEERING↗
NASA NTRS1976

Lightweight Long Life Heat Exchanger

A shuttle orbiter flight configuration aluminum heat exchanger was designed, fabricated, and tested. The heat exchanger utilized aluminum clad titanium composite parting sheets for protection against parting sheet pin hole corrosion. The heat exchanger, which is fully interchangeable with the shuttle condensing heat exchanger, includes slurpers (a means for removing condensed water from the downstream face of the heat exchanger), and both the core air passes and slurpers were hydrophilic coated to enhance wettability. The test program included performance tests which demonstrated the adequacy of the design and confirmed the predicted weight savings.

Moore, E. K.↗
DOE OSTI2022

Analyzing the effect of pressure on the properties of point defects in $\gamma\mathrm{U–Mo}$ through atomistic simulations

Uranium–molybdenum (U–Mo) alloys in monolithic fuel foil are the primary candidate for the conversion of high-performance research reactors in the USA. Monolithic fuel is utilized in a plate-type design with a zirconium diffusion barrier and aluminum cladding. These fuel types are unique in that they contain no plenum for the release of fission gases, which, in conjunction with the aluminum cladding, can lead to large stress states within the fuel. The nature of how fundamental processes of radiation damage, including the evolution of point defects, under such stresses occur is unknown. In this work, we present molecular dynamics simulations of the formation energy of point defects under applied stress. We report this work will allow for the implementation of stress-dependent microstructural evolution models of nuclear fuels, including those for both fission gas bubble growth and creep, which are critical to ensure the stable and predictable behavior of research reactor fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Initial assessment of clad/base metal systems for MSR

In support of the development of new design and construction rules and analysis methods for cladded structure for use under the anticipated service lifetime and cyclic service conditions in Molten Salt Reactors (MSR), research work has been initiated at ORNL for assessment of physical, metallurgical and mechanical properties of clad/base metal systems. This report summarizes the results for the nickel cladded SS316H system in FY18.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Second Target Station High-Fidelity Target Activation Comparison

The development of the Second Target Station (STS) target system at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) is well underway. The target system at STS consists of a rotating target disk that contains 21 segments of tungsten clad in tantalum clad in steel. A key aspect of the design of the target system is to account for the delayed heating and material damage caused by the delayed dose from decaying radionuclides. These radionuclides are a product of either spallation reactions or transmutation of the nuclei in the target system. These radionuclides build up in the target system components over the lifetime of the facility, and the radiation that is emitted can deposit energy in the components causing significant component heating and material damage. Monte Carlo N-Particle (MCNP) Version 6.2 transports the various particle species and calculates the spallation products and neutron fluxes throughout the target system. These spallation products and neutron fluxes along with the material definition of each component are relayed to the CINDER2008 transmutation code to calculate the radionuclide inventories and the corresponding decay gamma emission spectra. MCNP6.2 coupled with CINDER2008 is the computational method-of-choice for the analysis discussed in the following sections of this report. The analysis focuses on validating major assumptions in calculating the radionuclide inventory in the STS target system: all of the target segments are fresh, unirradiated material when the protons are incident on the segment, the average of the 21 segments of the target is sufficient to represent a single segment, and that averaging the proton pulse structure over time does not significantly affect the radionuclide inventory. The position-averaged, high-fidelity, and single-tally computational methods are used to validate the assumptions and provide a point of comparison to evaluate how the assumptions impact the radionuclide inventories. A more detailed explanation of the three computational methods is provided in Section 2. The position-averaged and single-tally methods are less computationally expensive when compared with the high-fidelity method where 54,000 individual calculations are needed to calculate 1 hr of STS operation. Section 3 details the comparison of the three methods to show that the assumptions made in the position-averaged method do not significantly impact the radionuclide inventory after 1 hr of operation. The discussions and results in this report are for 1 hr of operation. Due to the computational cost associated with calculating the transmutation and activation using the high-fidelity method, only 1 hr of operation has been calculated. The discrepancies observed after 1 hr of operation are not extrapolated out to longer operational times, and this report does not address how the discrepancies between the computational methods may manifest for longer operational periods.

43 PARTICLE ACCELERATORS↗
NASA NTRS1980

Fabrication and evaluation of brazed titanium-clad borsic/aluminum skin-stringer panels

A successful brazing process was developed and evaluated for fabricating full-scale titanium-clad Borsic/aluminum skin-stringer panels. A panel design was developed consisting of a hybrid composite skin reinforced with capped honeycomb-core stringers. Six panels were fabricated for inclusion in the program which included laboratory testing of panels at ambient temperatures and 533 K (500 F) and flight service evaluation on the NASA Mach 3 YF-12 airplane. All panels tested met or exceeded stringent design requirements and no deleterious effects on panel properties were detected followng flight service evaluation on the YF-12 airplane.

Bales, T. T.↗
DOE OSTI2017

IN-PILE MEASUREMENTS OF FUEL ROD DIMENSIONAL CHANGES UTILIZING THE TEST REACTOR LOOP PRESSURE FOR MOTION

Different types of materials are being considered for fuel, cladding, and structures in existing and advanced nuclear reactors. These materials can undergo significant dimensional and physical changes during irradiation. Currently in the US, such changes are measured with the ‘cook and look’ method (i.e. repeatedly irradiating a specimen for a specified period of time and then removing it from the reactor for evaluation). The time and labor to remove, examine, and return irradiated samples for each measurement makes this approach expensive. In addition, such techniques provide limited data, and handling may disturb the phenomena of interest. Therefore, in-pile detection of changes in geometry is sorely needed to understand real-time behavior during irradiation testing of fuels and materials in high flux US Material and Test Reactors (MTRs). This paper presents development results of an advanced Linear Variable Differential Transformer (LVDT)-based test rig capable of detecting real-time changes in diameter or length of fuel rods or other material samples during irradiation in US MTRs. This test rig is being developed at the Idaho National Laboratory (INL), and will provide experimenters with a unique capability to measure dimensional changes associated with fuel and clad swelling, pellet-clad interaction, and crud buildup. This novel LVDT design uses the MTR’s loop pressure to drive the test specimen at a constant rate. This leads to a simplified construction, that is, the use of fewer components than other diameter gauge test rigs. This simplified design only requires three pressure boundary penetrations. Further, a key component of this test rig is a leak tight hydraulic ram capable of actuating under low differential pressure, and is able to survive the high irradiation tests at Pressurized Water Reactor (PWR) conditions. The sensor has been proven on a benchtop scale, and shows repeatability and reliability. Measurement accuracies of ±3 µm for diameter and ±90 µm for length have been achieved.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Neutronic Consideration of TREAT Facility Fuel SiC Recladding

The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. Benchmark efforts have demonstrated sensitivity of TREAT neutronics models to graphite cross section data and key neutron poisons found in the fuel assemblies [2]. During fabrication of TREAT fuel assemblies, a small concentration of boron was inadvertently introduced with variable concentration into the aggregate of fuel blocks prepared for TREAT use. The concentration of boron, and its respective uncertainty, represent the most sensitive component in a neutronics study; unfortunately, nothing can be done to reduce or otherwise mitigate its impact. Hydrogen in the fuel blocks reduces the mean free path of the core and its true quantity and uncertainty is approximated. The hydrogen remains from pitch binder used in the fuel fabrication process and was not completely released due to incomplete fuel graphitization. Destructive fuel testing is necessary to quantify the actual hydrogen content and to improve benchmark model quality. A third neutron poison of some significance is the hafnium impurity in the Zircaloy-3 (Zry-3) assembly cladding. The quantity of hafnium is expected to be within nominal manufacturing limits for nuclear grade zircaloy. Recladding the fuel assemblies with material that is less neutron absorbing is expected to benefit reactor neutron economy. Moreover, TREAT’s total transient energy is restricted based on the rate of Zry-3 oxidation in TREAT’s air coolant environment. Hence, cladding with more benign air reaction kinetics could dramatically uprate the facility’s experimental capability. Silicon carbide (SiC) has been proposed as a candidate material for recladding TREAT assemblies, as SiC/SiC demonstrate favorable properties within nuclear environments [3]. The current operational limits of TREAT are restricted by the oxidation of Zry-3. Change of the cladding to SiC could feasibly double the current temperature limitation. Recladding opportunities allow for previously damaged fuel assemblies to be returned to service, as Zry-3 is not commercially available. The initial effort required to consider proceeding with any recladding effort is to neutronically characterize key reactor parameters while replacing Zry-3 with SiC. A characterization summary, with respective results, is provided herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗