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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 1,225 records · Page 68
All sourcesNASA NTRSDOE OSTI
NASA NTRS2005

SOLID SOLUTION CARBIDES ARE THE KEY FUELS FOR FUTURE NUCLEAR THERMAL PROPULSION

Nuclear thermal propulsion uses nuclear energy to directly heat a propellant (such as liquid hydrogen) to generate thrust for space transportation. In the 1960 s, the early Rover/Nuclear Engine for Rocket Propulsion Application (NERVA) program showed very encouraging test results for space nuclear propulsion but, in recent years, fuel research has been dismal. With NASA s renewed interest in long-term space exploration, fuel researchers are now revisiting the RoverMERVA findings, which indicated several problems with such fuels (such as erosion, chemical reaction of the fuel with propellant, fuel cracking, and cladding issues) that must be addressed. It is also well known that the higher the temperature reached by a propellant, the larger the thrust generated from the same weight of propellant. Better use of fuel and propellant requires development of fuels capable of reaching very high temperatures. Carbides have the highest melting points of any known material. Efforts are underway to develop carbide mixtures and solid solutions that contain uranium carbide, in order to achieve very high fuel temperatures. Binary solid solution carbides (U, Zr)C have proven to be very effective in this regard. Ternary carbides such as (U, Zr, X) carbides (where X represents Nb, Ta, W, and Hf) also hold great promise as fuel material, since the carbide mixtures in solid solution generate a very hard and tough compact material. This paper highlights past experience with early fuel materials and bi-carbides, technical problems associated with consolidation of the ingredients, and current techniques being developed to consolidate ternary carbides as fuel materials.

Panda, Binayak↗
DOE OSTI2019

Development of Advanced Instrumentation for Transient Testing

A fuel safety research program centered on in-pile transient testing experiments is being developed to support assessment and qualification of advanced nuclear fuel systems using the recently restarted Transient Reactor Test (TREAT) facility at the Idaho National Laboratory (INL). While resumption of transient testing at TREAT is crucial to enable these programs, full recovery and cutting-edge transient testing capability also requires a well-coordinated and innovative instrumentation development and qualification program to support near-term and future objectives. This paper summarizes the experimental approach of transient testing to focus on measuring the response of nuclear fuel to off-normal (or power-cooling mismatch) conditions for modern and advanced reactor environments requiring capabilities extending over wide measurement and environment conditions. It also highlights unique attributes of transient testing of importance to in-pile instruments including relatively low total neutron fluence, high gamma heating, and the need for well-defined and possibly short time response. Historical approaches to instrumentation for transient testing are also reviewed to provide context to the modern instrument strategy. The instrumentation needs of a modern transient testing are detailed while summarizing several ongoing R&D activities are supporting the development of state-of-the-art and ‘advanced’ measurement technologies that will provide baseline capability for Light Water Reactor (LWR) and Sodium-Cooled Fast Reactor (SFR) experiment objectives, while extending to other advanced reactor needs and advanced sensing technology opportunities. Examples of specific sensors planned for near-term deployment with ongoing development includes prompt-response self-powered neutron detectors (SPND), miniature fission chambers, optical-fiber-coupled infrared pyrometer, cladding surface thermocouples, electrical-impedance based boiling detector, and Linear Variable Differential Transformer (LVDT) based sensors for fuel elongation and pressure measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Role of Chiral Two-Body Currents in 6 Li Magnetic Properties in Light of a New Precision Measurement with the Relative Self-Absorption Technique

A direct measurement of the decay width of the excited $0^+_1$ state of 6 Li using the relative self-absorption technique is reported. Our value of $Γ_{γ,0^{+}_{1}} → 1^{+}_{1}$ = 8.17 (14) stat. (11) syst. eV provides sufficiently low experimental uncertainties to test modern theories of nuclear forces. The corresponding transition rate is compared to the results of ab initio calculations based on chiral effective field theory that take into account contributions to the magnetic dipole operator beyond leading order. This allows for a precision test of the impact of two-body currents that enter at next-to-leading order.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Corrosion testing needs and considerations for additively manufactured materials in nuclear reactors

Additive manufacturing (AM) technologies have developed rapidly in recent years, creating new opportunities and challenges for the nuclear industry; however, adoption requires that their corrosion performance be evaluated. Here, we discuss known reactor-specific corrosion issues for multiple reactor types and engineering concerns such as regulations and standards. A review of corrosion studies conducted on select AM alloys informs a discussion on key bulk and surface factors likely to impact corrosion behaviors. Recommendations to assess corrosion performance for AM materials are provided, including management of the unique nature of as-built AM surfaces and the inherent process variability that occurs for AM components.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2000

A Nuclear Interaction Model for Understanding Results of Single Event Testing with High Energy Protons

An internuclear cascade and evaporation model has been adapted to estimate the LET spectrum generated during testing with 200 MeV protons. The model-generated heavy ion LET spectrum is compared to the heavy ion LET spectrum seen on orbit. This comparison is the basis for predicting single event failure rates from heavy ions using results from a single proton test. Of equal importance, this spectra comparison also establishes an estimate of the risk of encountering a failure mode on orbit that was not detected during proton testing. Verification of the general results of the model is presented based on experiments, individual part test results, and flight data. Acceptance of this model and its estimate of remaining risk opens the hardware verification philosophy to the consideration of radiation testing with high energy protons at the board and box level instead of the more standard method of individual part testing with low energy heavy ions.

Culpepper, William X.↗
DOE OSTI2017

Overview of Alloy 617 Notched Specimen Testing

Prior to use in high temperature nuclear applications, Alloy 617, which is currently undergoing approval for inclusion in codes and regulations related to high temperature construction, must be approved by the Nuclear Regulatory Commission. Several concerns relating to the effects of geometric discontinuities, such as notches, and multi-axial stress states (as opposed to uniaxial stress states typically present in laboratory run creep tests) must first be addressed. This work has shown that at short time intervals (1000-2500 hours), 617 is notch strengthening and will fail preferentially at locations away from the notch, provided that diameters are similar (resulting in similar stresses). Additionally, specimens designed with semi-circular notches (or U-notches, as they are referred to in this report) to impose multi-axial stress states on the specimen during creep testing were found to either maintain similar creep rupture lives as uniaxial stress specimens tested at similar conditions, or have creep rupture lives that were significantly longer. This work, while not fully completed, shows that notches and multi-axial stress states do not appear to raise concerns with relation to the use of design rules created for Alloy 617 using smooth, straight gauge test specimens and uniaxial creep tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1972

Flowing gas, non-nuclear experiments on the gas core reactor

Flow tests were conducted on models of the gas core (cavity) reactor. Variations in cavity wall and injection configurations were aimed at establishing flow patterns that give a maximum of the nuclear criticality eigenvalue. Correlation with the nuclear effect was made using multigroup diffusion theory normalized by previous benchmark critical experiments. Air was used to simulate the hydrogen propellant in the flow tests, and smoked air, argon, or freon to simulate the central nuclear fuel gas. All tests were run in the down-firing direction so that gravitational effects simulated the acceleration effect of a rocket. Results show that acceptable flow patterns with high volume fraction for the simulated nuclear fuel gas and high flow rate ratios of propellant to fuel can be obtained. Using a point injector for the fuel, good flow patterns are obtained by directing the outer gas at high velocity along the cavity wall, using louvered or oblique-angle-honeycomb injection schemes.

Kunze, J. F.↗
DOE OSTI2022

Benchmark Specification for Select Experiments Conducted at the University of Wisconsin-Madison Thermal Stratification Test Facility

The Department of Energy (DOE)-Nuclear Energy University Programs (NEUP) supported the creation and operation of the Thermal Stratification Test Facility (TSTF) at the University of Wisconsin Madison (UWM) as part of a larger effort to understand thermal stratification behavior in liquid-metal-cooled reactors. The TSTF was designed to simulate transients in a reactor plenum that are known to cause thermal stratification. High-reliability and high-resolution measurements of the flow and temperature were collected for use as experimental benchmarks to support validation efforts for computational models. The results of these tests contribute to the greater understanding of thermal stratification behavior of liquid sodium under various configurations and operating conditions. The six TSTF tests selected for benchmarking are a set of forced circulation tests at a fixed flow rate with different Upper Internal Structure (UIS) configurations in the test section (no UIS, solid UIS, and a UIS with flow area of 4, 8, 12, and 100%). This report provides a complete description of the benchmark problems, including all key test facility details, descriptions of each test condition, and measured data for comparison with modeled results.

42 ENGINEERING↗
DOE OSTI2024

Advanced Test Reactor Long-Term (20-Year) Operational Strategy

The Department of Energy Office of Nuclear Energy (DOE-NE) and the Naval Nuclear Propulsion Program (NNPP) have identified the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) as a critical need for continued thermal irradiation testing capability to support and advance U.S. commercial and naval nuclear power systems. The ATR is currently the only suitable test reactor operating in the world that can provide these capabilities at volume. ATR must, then, plan to operate for at least the next 20 years. As of the beginning of 2024, ATR has operated for 57 years, with many of the original reactor structures, systems, and components installed and operated for 60 years. Some ATR Complex and reactor support equipment that is currently in use was installed 75 years ago with the startup of the Materials Testing Reactor (MTR).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Innovative Hot Lab Concept for Advanced D and D Application - 19175

Hot Labs are used in the nuclear industry for various purposes: maintenance of nuclear reactor, sampling and testing of fuel, inspection activities and decommissioning. Despite their vast use, their current designs and operations do not meet entirely today's needs and requirements of the operations nor tomorrow's. This paper will give an overview of the research and investigations performed around Hot Cells design and concept in order to integrate new operability modes and functions. The overall idea is to bring modularity, flexibility and adaptability to the Hot Lab nuclear industry. Taking the key principles of Model Based System Engineering, an interoperable data model, the most advanced digital technologies (Building Information Modelling - BIM, Virtual Reality / Augmented Reality, Artificial Intelligence), and looking at all the life-cycle of such a facility, a concept design of a new type of Hot Lab has been developed. Four mains aspects are implemented during this study: Implement the principles of the Industry 4.0; use modular and flexible solutions; explore the concept of in service maintenance and predictive maintenance; optimise the decommissioning and dismantling process by taking into account the entire life-cycle of the facility. The design was developed as a case study taking several assumptions to adapt to any project. The design developed demonstrates that modularity and flexibility can be brought in without compromising the core safety functions of such a facility. This Hot Lab concept brings the current Hot Lab to the next level as the design has been adapted to meet the current operational needs, implement novel ideas and consider its entire life-cycle. This paper will demonstrate how decommissioning and dismantling operations can become more efficient and safer with such a concept. In addition, the Model Based System Engineering approach used from the beginning of the design phase allows to have an optimum architecture during its entire lifetime. It also allowed to flag new topics to be further developed for the benefit of the nuclear industry as a whole. A first example would be the development of the Control System using digital tools and artificial intelligence to run remotely controlled equipment inside the Hot Lab. Assystem is an international engineering group. As a key participant in the nuclear industry for over 50 years, the Group supports its clients in managing their capital expenditure throughout their asset life-cycles. Assystem can offer support in project management, design, digital engineering, industrial control systems, security and safety and industrial digitalisation. Assystem's Imagine program is an internal program intending to serve as an incubator for the development of innovations. The Innovative Hot Lab project integrates one of the four innovation projects Assystem choose to develop and invest in. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2018

Radiation Transport and Shielding Challenges at NASA Marshall Space Flight Center

Several unique radiation transport and shielding challenges for various spacecraft and ground test facilities have recently been explored at NASA Marshall Space Flight Center (MSFC). Many radiation transport methods and code packages have been utilized during this effort, and some additional analytical tools have been developed to support this work. This includes the development of a nuclear thermal propulsion stage, in which radiation damage to hardware, radiation heating, and crew dose must be assessed for a large spacecraft. Radiation shielding for such a stage must be optimized to minimize weight, and additional dose consequences are expected from cosmic sources. Development of a nuclear engine requires ground test facilities, which themselves require significant development effort that includes deep-penetration shielding analysis for facility design and site selection. Additional radiation transport analysis has been performed in support of a notional lander mission to Europa, the icy moon of Jupiter. The radiation burden to spacecraft has been analyzed, and charged particle radiation transport methods have been utilized to apply correct doses to material samples in ground test facilities.

Caffrey, Jarvis A.↗
DOE OSTI2025

Study of hydride reorientation and ductility of SRA ZIRLO and PRXA ZIRLO using ring compression testing (RCT)

Before dry storage, spent nuclear fuel is dried at high temperatures, which induces hydride dissolution. Radial hydride particles precipitate when cooled under a stress above a threshold stress. This process can severely degrade the ductility of cladding tubes. Zirconium alloys with different microstructures, such as grain boundary density, were quantified and studied to examine their effect on hydride reorientation. The ring compression loading mode creates a range of stress states depending on the location within the ring with respect to the applied load, producing various hydride morphologies that can be used to study the effects of microstructures on hydride reorientation. Finite element modeling was performed to calculate the hoop stress, which is the dominant stress, during hydride reorientation. These results were coupled with hydride morphology to determine the threshold stress for hydride reorientation, depending on the microstructure. In conclusion, the samples were then mechanically tested using ring compression to evaluate the ductility of zirconium alloys as a function of different hydride morphologies.

Hydride embrittlement↗
NASA NTRS1993

Nuclear thermal propulsion technology: Results of an interagency panel in FY 1991

NASA LeRC was selected to lead nuclear propulsion technology development for NASA. Also participating in the project are NASA MSFC and JPL. The U.S. Department of Energy will develop nuclear technology and will conduct nuclear component, subsystem, and system testing at appropriate DOE test facilities. NASA program management is the responsibility of NASA/RP. The project includes both nuclear electric propulsion (NEP) and nuclear thermal propulsion (NTP) technology development. This report summarizes the efforts of an interagency panel that evaluated NTP technology in 1991. Other panels were also at work in 1991 on other aspects of nuclear propulsion, and the six panels worked closely together. The charters for the other panels and some of their results are also discussed. Important collaborative efforts with other panels are highlighted. The interagency (NASA/DOE/DOD) NTP Technology Panel worked in 1991 to evaluate nuclear thermal propulsion concepts on a consistent basis. Additionally, the panel worked to continue technology development project planning for a joint project in nuclear propulsion for the Space Exploration Initiative (SEI). Five meetings of the panel were held in 1991 to continue the planning for technology development of nuclear thermal propulsion systems. The state-of-the-art of the NTP technologies was reviewed in some detail. The major technologies identified were as follows: fuels, coatings, and other reactor technologies; materials; instrumentation, controls, health monitoring and management, and associated technologies; nozzles; and feed system technology, including turbopump assemblies.

Clark, John S.↗
DOE OSTI2020

Selected Uses of TSUNAMI in Critical Experiment Design and Analysis

Validation in criticality safety is performed by comparing the results of critical experiments with the calculated results from models of the experiments using the computational method to be validated. Laboratory critical experiments are controlled systems that achieve a k eff of approximately 1 and enable investigation of the parameters at which such a critical condition is achieved. For the critical experiments used in a validation to capture the biases of the materials and neutron energy spectra of interest, those materials must be included in the experiment such that they influence k eff or another observable parameter with statistical significance. This paper discusses the use of sensitivity uncertainty (S/U) methods to develop critical experiments for various purposes. S/U techniques are useful for understanding the underlying components of nuclear data which affect the k eff or another parameter of a given configuration. S/U calculations are most commonly used to compare existing experiments to applications of interest; however, S/U techniques can also be used to identify, optimize, or assess features of proposed experiments so that they can better test specific portions of nuclear data or match an application of interest. The S/U techniques discussed here are from the TSUNAMI code system. The two primary codes discussed in this work are TSUNAMI-3D, which implements the KENO criticality code to calculate the sensitivity of k eff to nuclear data, and TSAR, which calculates the sensitivity of a reactivity difference between two configurations based on their TSUNAMI-3D generated sensitivity profiles. The methods used in these tools are discussed in more detail in the SCALE manual. This paper is one of a series on the development and use of TSUNAMI tools. The other papers address development of TSUNAMI methods and a review of TSUNAMI applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗