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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

In-situ ion irradiation of a spent UO2 fuel: evolution of fission products and nanograins with radiation dose

This study investigates the behavior of fission gas bubbles, five metal precipitates (5MPs) (Mo, Ru, Rh, Tc, Pd) and nanograins in spent uranium dioxide (UO2) fuel under various ion irradiation doses, temperatures, and flux conditions. Utilizing in-situ ion irradiation and advanced transmission electron microscopy, we analyzed the evolution of fission gas bubbles, 5MPs and nanograins in UO2 samples from the Belgium Reactor 3 (BR-3). Our findings reveal significant shrinkage of fission gas bubbles and 5MPs with increasing irradiation dose, accompanied by a decrease in pair density. The study also reveals nanograin coarsening under irradiation. We demonstrate that ion irradiation induces a homogeneous re-solution process where individual atoms are ejected from bubbles and precipitates, leading to their dissolution and subsequent re-precipitation in the matrix. This study provides critical insights into the dynamic behavior of fission products and nanograins under irradiation, facilitating the development of predictive models and contributing to the optimization of nuclear fuel performance and safety.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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.↗
DOE OSTI2024

Multi-principal element alloys for fast reactor cladding applications

Given the extensive list of multi-principal element alloys (MPEAs) within literature and the overwhelming number of alloys that can be made from only a handful of elements, this article proposes a methodology for prioritizing alloys for use as cladding within advanced reactors. This paper applies neutronic, mechanical, and chemical assessments to a collection of MPEAs available within literature for use as advanced reactor cladding. The results are compared to a reference design of HT9, a ferritic/martensitic steel, for employed in sodium-cooled fast reactor. Mechanical assessments determined pressure limits for a thin-walled tube pressure boundary, thus relating the material's mechanical properties to a minimum acceptable wall thickness. Neutronic analyses reveal a maximum allowable wall thickness that a given material must meet to provide a level of neutronic economy that is equivalent to than that of HT9. Lastly, each alloying element is compared to typical fission products found in a fast reactor fuels to mitigate deleterious phenomena such as fuel-cladding chemical interactions (FCCI). These analyses indicate that Mo-Nb-Ti-V-based alloys are likely to be advantageous and that the inclusion of elements with a high neutronic penalty (e.g., Hf) could be considered with minimal consequences.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1974

Physics and potentials of fissioning plasmas for space power and propulsion

A brief description of two gas core reactor concepts devised for propulsion applications is presented and recent research on the emission of radiation from nonequilibrium fissioning gases is considered. The principles of UF6 and plasma core reactor experiments are discussed. It is expected that the program of nuclear gas core reactor experiments can lead in several years to reactor operation at pressures, temperatures, and power levels of technological significance to terrestrial applications.

Thom, K.↗
NASA NTRS2016

Pressure Loss Predictions of the Reactor Simulator Subsystem at NASA Glenn Research Center

Testing of the Fission Power System (FPS) Technology Demonstration Unit (TDU) is being conducted at NASA Glenn Research Center. The TDU consists of three subsystems: the reactor simulator (RxSim), the Stirling Power Conversion Unit (PCU), and the heat exchanger manifold (HXM). An annular linear induction pump (ALIP) is used to drive the working fluid. A preliminary version of the TDU system (which excludes the PCU for now) is referred to as the "RxSim subsystem" and was used to conduct flow tests in Vacuum Facility 6 (VF 6). In parallel, a computational model of the RxSim subsystem was created based on the computer-aided-design (CAD) model and was used to predict loop pressure losses over a range of mass flows. This was done to assess the ability of the pump to meet the design intent mass flow demand. Measured data indicates that the pump can produce 2.333 kg/sec of flow, which is enough to supply the RxSim subsystem with a nominal flow of 1.75 kg/sec. Computational predictions indicated that the pump could provide 2.157 kg/sec (using the Spalart-Allmaras (S‒A) turbulence model) and 2.223 kg/sec (using the k- turbulence model). The computational error of the predictions for the available mass flow is ‒0.176 kg/sec (with the S-A turbulence model) and -0.110 kg/sec (with the k- turbulence model) when compared to measured data.

Mass Flow↗
DOE OSTI2024

A Molten Salt Aerosol Spectroscopy Approach to Online Process Monitoring in Molten Salt Reactors

Recent research and development in molten salt reactors (MSRs) has led to many advanced reactor designs, many of which have demonstrations planned in the next decade. However, a significant challenge that must be overcome before liquid fueled MSRs become viable globally is nuclear material accountancy (NMA) and safeguards. Liquid-fueled MSRs are problematic to safeguards because the nuclear material is in the bulk phase, inventories are constantly changing, and there are potentially many outlet pathways that must be monitored, especially in designs with online chemical processing. At the Idaho National Laboratory, we have developed a process monitoring tool to circulate molten salt through a nebulizer and optical cell, where laser-induced breakdown spectroscopy (LIBS) and ultraviolet-visible (UV-VIS) are applied to track the elemental and oxidation states of the constituents in the salt aerosol. This approach has the potential to provide information on the actinides, fission products, and corrosion in the reactor or process. Preliminary testing in aqueous praseodymium chloride (PrCl3) and neodymium chloride (NdCl3) showed good sensitivity detection limits n developed calibration curves. Additional tests and analysis or ongoing in aqueous solutions while equipment is being scaled up for high temperature molten salt testing. Results for continued testing as well as a status of the molten salt design and anticipated commissioning will be presented.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Evaluation of an Accident Tolerant Fuel Leak in the Advanced Test Reactor

Accident Tolerant Fuels (ATF), which are nuclear fuel sources designed to withstand operational irregularities and incidents, have been a topic of interest in the nuclear industry for several decades. Interest in ATF technology surged following the 2011 accident at Fukushima Daiichi in Japan. At the Advanced Test Reactor (ATR), one of Idaho National Laboratory’s (INL) four operating nuclear reactors, the ATF program is a collaborative effort between the national laboratory and various stakeholders within the nuclear industry. This program focuses on the research and development of novel fuel compositions, cladding, and component materials with enhanced accident-resistant properties. During one of ATR’s 60-day operating cycles in 2024, the reactor experienced five unplanned shutdowns. Following the fifth shutdown, radiation monitors detected an increase in radiation levels coming from the loop piping. Subsequent water samples confirmed the cause was a leak of fission products from the ATF experiment, designated as ATF-2C. The source of the leak was identified as the instrumented section of the test train. The primary discussions in this paper are 1) the design of the ATF test train, 2) the operating parameters leading up to and following the detection of the leak, and 3) the quantification and characterization of the released fission products.

Accident Tolerant Fuels↗
DOE OSTI2025

Evaluation of an Accident Tolerant Fuel Leak in the Advanced Test Reactor

Accident Tolerant Fuels (ATF), which are nuclear fuel sources designed to withstand operational irregularities and incidents, have been a topic of interest in the nuclear industry for several decades. Interest in ATF technology surged following the 2011 accident at Fukushima Daiichi in Japan. At the Advanced Test Reactor (ATR), one of Idaho National Laboratory’s (INL) four operating nuclear reactors, the ATF program is a collaborative effort between the national laboratory and various stakeholders within the nuclear industry. This program focuses on the research and development of novel fuel compositions, cladding, and component materials with enhanced accident-resistant properties. During one of ATR’s 60-day operating cycles in 2024, the reactor experienced five unplanned shutdowns. Following the fifth shutdown, radiation monitors detected an increase in radiation levels coming from the loop piping. Subsequent water samples confirmed the cause was a leak of fission products from the ATF experiment, designated as ATF-2C. The source of the leak was identified as the instrumented section of the test train. The primary discussions in this presentation are 1) the design of the ATF test train, 2) the operating parameters leading up to and following the detection of the leak, and 3) the quantification and characterization of the released fission products.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Implementation of a Drift Flux Model into SAM with Development of a Verification and Validation Test Suite for Modeling of Noncondensable Gas Mixtures

The advanced thermal-hydraulic system code, System Analysis Module (SAM), was originally developed for the modeling of single-phase flow in advanced reactors. It has since been expanded to include a four-equation drift flux model for the modeling of two-phase flows containing a noncondensable gas. The model was expanded to support the modeling of molten salt reactor (MSR) designs in which the fuel is directly dissolved in the circulating coolant. These designs have shown that circulating gas bubbles can play an important role in the management of fission products and the operational behavior of the reactor. A drift flux model was implemented to more accurately capture the localized behavior of the void in the core and its impact on the mass transfer of fission products. A thorough assessment of the new model was performed by developing a verification and validation test suite. Verification problems were designed to test all major terms in the new governing equations. The new model converged to the correct solution at the expected order of accuracy for all verification cases. The validation cases included a wide range of flow and void conditions in different pipe geometries. Although higher void experiments show a slight underprediction of void by the drift flux model, experiments that aim to reproduce Molten Salt Reactor Experiment (MSRE) experimental conditions show good agreement with the model. The gas transport model was activated for a SAM model of the MSRE to demonstrate that it can be used in a more complex model. Finally, this gas transport model will be used along with an interfacial area transport equation being implemented in SAM for the prediction of mass transport behavior in MSR conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Development and Validation of Two-Phase Flow Models in MOOSE and Application to Molten Salt Reactors

Two-phase flow in Molten Salt Reactors (MSRs) is important as it impacts reactivity evolution, reactor transient response, and the removal of species dissolved in the molten salt through gas phase transfer. Therefore, accurately predicting the gas distribution and the associated liquid-gas interface area in MSRs is essential for their design and operation. Recently, we integrated a new two-phase model into Idaho National Laboratory (INL)’s Multiphysics Object-Oriented Simulation Environment (MOOSE): a multi-D generalization of a mixture drift-flux model. It provides greater computational efficiency, which is typically preferred for modeling reactor transients. However, the mixture model’s accuracy in capturing void distribution and interfacial area in MSRs still needs to be assessed. This article begins with a description of the mathematical framework for the two-phase model implemented in MOOSE. It then presents validation of these models against relevant experimental data. Finally, the model is applied to the Molten Salt Reactor Experiment case study, analyzing various operational conditions such as different rates of fission product volatilization and diverse cover gas entrainment scenarios at the reactor pump. The article concludes by assessing the suitability of the mixture drift-flux model for capturing the two-phase flow dynamics critical to MSR operations

42 - ENGINEERING↗
DOE OSTI2025

Development and Validation of Two-Phase Flow Models in MOOSE for Molten Salt Reactor Application

Two-phase flow in Molten Salt Reactors (MSRs) is important as it impacts reactivity evolution, reactor transient response, and the removal of species dissolved in the molten salt through gas phase transfer. Therefore, accurately predicting the gas distribution and the associated liquid-gas interface area in MSRs is essential for their design and operation. Recently, we integrated two new models into Idaho National Laboratory (INL)’s Multiphysics Object-Oriented Simulation Environment (MOOSE): a multi-D generalization of a mixture drift-flux model and a Euler-Euler model. The Euler-Euler model offers higher fidelity, while the mixture drift-flux model provides greater computational efficiency, which is typically preferred for modeling reactor transients. However, the mixture model's accuracy in capturing void distribution and interfacial area in MSRs still needs to be assessed. This article begins with a description of the mathematical framework for the two-phase models implemented in MOOSE. It then presents validation of these models against relevant experimental data. Finally, both models are applied to the Molten Salt Reactor Experiment case study, analyzing various operational conditions such as different rates of fission product volatilization and diverse cover gas entrainment scenarios at the reactor pump. The article concludes by assessing the suitability of both models for capturing the two-phase flow dynamics critical to MSR operations.

42 - ENGINEERING↗
DOE OSTI2025

Development and Validation of Two-Phase Flow Models in MOOSE and Application to Molten Salt Reactors

Two-phase flow in Molten Salt Reactors (MSRs) is important as it impacts reactivity evolution, reactor transient response, and the removal of species dissolved in the molten salt through gas phase transfer. Therefore, accurately predicting the gas distribution and the associated liquid-gas interface area in MSRs is essential for their design and operation. Recently, we integrated two new models into Idaho National Laboratory (INL)’s Multiphysics Object-Oriented Simulation Environment (MOOSE): a multi-D generalization of a mixture drift-flux model and a Euler-Euler model. The Euler-Euler model offers higher fidelity, while the mixture drift-flux model provides greater computational efficiency, which is typically preferred for modeling reactor transients. However, the mixture model's accuracy in capturing void distribution and interfacial area in MSRs still needs to be assessed. This article begins with a description of the mathematical framework for the two-phase models implemented in MOOSE. It then presents validation of these models against relevant experimental data. Finally, both models are applied to the Molten Salt Reactor Experiment case study, analyzing various operational conditions such as different rates of fission product volatilization and diverse cover gas entrainment scenarios at the reactor pump. The article concludes by assessing the suitability of both models for capturing the two-phase flow dynamics critical to MSR operations.

42 - ENGINEERING↗
DOE OSTI2024

Radiation induced athermal diffusivity in uranium mononitride

Uranium mononitride (UN) is one of the ceramic nuclear fuel alternatives to oxide fuel considered for light water reactors and advanced reactor designs. Properties like self- and fission gas diffusivity need to be better understood, given that they influence key fuel performance phenomena such as fission gas swelling and release. In particular, the radiation induced athermal (D 3 ) diffusivity remains challenging to accurately predict and has only been sparsely characterized in UN, despite its importance as it likely governs diffusion at the low temperatures this high-thermal-conductivity fuel form may operate. Molecular Dynamics simulations are used to estimate the mean square displacement induced by a primary knock-on atom (PKA) with a given kinetic energy. These results are combined with the PKA energy distributions obtained from binary collision approximation calculations to obtain the displacement due to a particular fission fragment. Finally, this is combined with experimental fission fragment yields to determine the displacement due to an average fission event and, thus, express the athermal diffusivity as a function of the fission rate density. These results are in excellent agreement with available experimental data. In conclusion, a particular importance is given to the understanding and the quantification of the variability of these results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Decay studies of the 𝛽-delayed neutron emitters 87 Br and 88 Br measured by means of the Modular Total Absorption Spectrometer at ORNL HRIBF

The β decays of 87 Br, 88 Br, and 87 Kr were measured with the Modular Total Absorption Spectrometer (MTAS) at Oak Ridge National Laboratory’s Holifield Radioactive Ion Beam Facility (HRIBF). Both bromine isotopes are β-delayed neutron emitters that have large cumulative fission yields and were identified as top-priority cases for total absorption study by the Nuclear Energy Agency in 2007. Our investigations corroborate that the decay schemes of 87 Br and 88 Br suffer from the so-called pandemonium effect. Unique MTAS properties enable direct neutron measurements. Here, we present MTAS-derived β-delayed neutron spectra, β-delayed neutron emission probabilities of P n ( 87 Br) = 2.36(24)%, and P n ( 88 Br) = 6.4(6)%, and the β-delayed neutron transitions intensity of 4(2)% to the first excited 87 Kr state, populated in the β-neutron decay of 88Br. Incorporating new data into calculations of the electromagnetic decay heat component emitted during thermal neutron fission of 235 U and 239 Pu improves agreement with experimental data up to approximately 80 s after fission. The estimation of the nuclear reactor $\bar{ν}_e$ flux results in changes of up to 1% in the expected $\bar{ν}_e$ interactions with the detector material for 235 U, 238 U, 239Pu, and 241 Pu.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Calculation Of Neutrino Flux

The CONFLUX software framework, Calculation Of Neutrino FLUX, is built with the goal to simplify and standardize the calculation. CONFLUX packages three methods to calculate neutrinos generated from reactor neutrinos or individual beta decays with common nuclear data and beta theories for direct cross-method comparison. The software prepacked the latest nuclear database, including ENDF.B/VIII, JEFF 3.3, and ENSDF, as well as methods to process the uncertainties. It also allows customized nuclear data and beta theories and user generated time dependent reactor models for convenient adjustment of fission products, theoretical corrections.

Zhang, Xianyi [Lawrence Livermore National Laborat↗
DOE OSTI2024

Metallic Coating of Cerium Oxide Microspheres

The ability to remove heat is paramount to nuclear fuel performance and longevity. Retaining fission product and separating fuel from reactor coolant and the environment is also necessary to prevent radiological contamination. Conventional nuclear fuel for commercial light water reactors and radioisotope power systems (RPS) is composed of oxide powders pressed into a pellet (cm-scale) and then sealed into a metal cladding to confine the fuel. What typical fuels lack is a method to surround each particle of nuclear fuel in metal, thus providing a more intimate protection layer for accident tolerance and boosting the thermal extraction from the fuel element. In such a way, metal-coated fuel particles increase heat extraction efficiency over clad-pellet designs while increasing the accident tolerance of the fuel. Metal oxide microspheres have wide-ranging applications, including the realm of fuels for nuclear reactors and RPS. Microspheres of uranium oxide/uranium carbide, mixed uranium/plutonium oxides, transuranics, and thorium fuels have been extensively studied. Pacific Northwest National Laboratory has also demonstrated the production of 238 PuO 2 microspheres for RPS applications. Metal-coated oxide microsphere fuels may also be attractive for other applications such as nuclear thermal rockets, future nuclear reactor designs, and catalysts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗