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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 1,693 records · Page 94
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine how specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs

Wootan, David W.↗
DOE OSTI2024

Demonstrating autonomous controls on hardware test beds is a necessity for successful missions to Mars and beyond

NASA and the Department of Defense are planning for a mission to Mars in the 2030s–2040s using nuclear thermal propulsion (NTP). NTP uses a nuclear reactor to heat flowing hydrogen and create thrust. A serious concern for crewed and uncrewed missions to Mars is the loss of reactor control. The reactor startup and initial rocket impulse are initiated in cislunar or near-earth orbital regions; therefore, radio communications between ground control and the NTP engine should occur in real time. However, radio communications can take more than 20 min, depending on planet positions, to reach Mars orbiters from ground control. To address this delay, local autonomous controls are implemented onboard the NTP engine to ensure acceptable operation. However, autonomous controls have not been demonstrated or implemented in research or power reactor contexts because of safety and reliability concerns. To enable autonomous controls development, demonstration, and validation, Oak Ridge National Laboratory has created a nonnuclear hardware-in-the-loop test bed. Sensors throughout the test bed relay system status and hardware response to the user control algorithm, including measurements of temperature, flow, pressure of a loop, control drum position, and drum speed. This paper discusses the development of this facility and user accessibility.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2022

Analysis of Deformation Mechanisms, Strain Localization, and Fracture in Highly Irradiated Austenitic Steels – Light-Water Reactor Core Materials – via Advanced inSEM Techniques

The present report describes the establishing of an advanced experimental approach – high-resolution digital image correlation or µDIC – for testing miniature irradiated specimens in the ORNL’s Low Activation Materials Development and Analysis (LAMDA) facility. As expected, the µDIC approach will allow for studying strain localization inside grains, for instance, during dislocation channel-grain boundary interaction processes. This information is of high importance for understanding degradation processes in irradiated austenitic steels – in-core materials of Light Water Reactors. The Level 4 report provides a brief introduction, lists materials designated for µDIC testing and delivered in LAMDA this FY, discusses methodical concerns and preliminary observations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Sample Testing Basket: Design and Capabilities Report

SFR designs all require integral components that can withstand prolonged submersion in the liquid sodium primary coolant as well as in the gas space of the reactor vessel. Materials are often desired to withstand these conditions for the lifetime of the reactor. The Sample Testing Basket (STB) was designed and fabricated to meet the anticipated need for material exposure testing in prototypic reactor coolant environments; in the liquid sodium primary coolant and the sodium-vapor-laden gas space. Developed at Argonne National Laboratory (ANL), the STB is a piece of equipment designed for use in the Mechanisms Engineering Test Loop (METL) to allow for sodium exposure testing of materials within the METL facility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY17 FInal Project Report

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for this three year project.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Logistics for the 2026 International Nuclear Accident Dosimetry Intercomparison Exercise

The purpose of this exercise is to test participants’ nuclear accident dosimeters (NADs) and personnel to the requirements established by American National Standards Institute (ANSI), Health Physics Society (HPS) standard ANSI/HPS-N13.3-2013 (R2025) Dosimetry for Criticality Accidents. IER-634 is the second nuclear accident dosimetry exercise taking place at the Armed Forces Radiobiology Research Institute (AFRRI) in Bethesda, MD, a Department of Defense (DOD) facility. This effort continues use of the high-dose neutron exposure test bed maintained by the DOD and represents the continued collaboration between the DOD and Department of Energy (DOE). Previous work at AFRRI included a reactor characterization (IER 484) in 2023 and a nuclear accident intercomparison exercise (IER-602) in 2024.

61 RADIATION PROTECTION AND DOSIMETRY↗
NASA NTRS2011

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Bradley, D. E.↗
DOE OSTI2020

Development and Implementation of a CTF Code Verification Suite.

CTF is a thermal hydraulic subchannel code developed to predict light water reactor (LWR) core behavior. It is a version of Coolant Boiling in Rod Arrays (COBRA) developed by Oak Ridge National Laboratory (ORNL) and North Carolina State University (NCSU) and used in the Consortium for the Advanced Simulation of LWRs (CASL). Here, the existing CTF code verification matrix is expanded, which ensures that the code is a faithful representation of the underlying mathematical model. The suite of code verification tests are mapped to the underlying conservation equations of CTF and significant gaps are addressed. As such, five new problems are incorporated: isokinetic advection, conduction, pressure drop, convection, and pipe boiling. Convergence behavior and numerical errors are quantified for each of the tests and all tests converge at the correct rate to their corresponding analytic solution. A new verification utility that generalizes the code verification process is used to incorporate these problems into the CTF automated test suite.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

In-Canal Assay of High Specific Activity 60 Co at the Advanced Test Reactor

Production of high specific activity (HSA) 60 Co has recently resumed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The technical steps of performing in-canal assay of HSA 60 Co targets at the ATR are described herein. The HSA targets are assayed on a regular basis, between cycles at the ATR, in order to assess the progress of activation. The targets are also assayed at the conclusion of the irradiation in order to provide activity estimates for the distributor and to be used for safety and shipping evaluations. These target assay activities must take place in the ATR canal to provide sufficient radiological shielding. Furthermore, a specialized assay fixture is used in conjunction with custom 60 Co radiation standards to assess the irradiated target. The specific activity of each irradiated target was determined by first measuring the 60 Co standards to determine a sensitivity factor. Then, each irradiated target was measured, and the integrated measurement was multiplied by the sensitivity factor to determine the total activity of the irradiated target. Finally, a correction factor was determined to adjust the reported activity, accounting for differences in the physical geometries of the standards and targets. These methods were used to accurately assay the gross activity of nine irradiated 60 Co targets, two of which were delivered to the distributor where confirmatory hot-cell assays were performed verifying the accuracy of the in-canal assay method.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Influence of temperature, oxygen partial pressure, and microstructure on the high-temperature oxidation behavior of the SiC Layer of TRISO particles

Tristructural isotropic (TRISO)-coated fuel particles are designed for use in high-temperature gas-cooled nuclear reactors, featuring a structural SiC layer that may be exposed to oxygen-rich environments over 1000 °C. Surrogate TRISO particles were tested in 0.2–20 kPa O 2 atmospheres to observe the differences in oxidation behavior. Oxide growth mechanisms remained consistent from 1200–1600 °C for each P O$_2$ , with activation energies of 228 ± 7 kJ/mol for 20 kPa O 2 and 188 ± 8 kJ/mol for 0.2 kPa O 2 . At 1600 °C, kinetic analysis revealed a change in oxide growth mechanisms between 0.2 and 6 kPa O2. In 0.2 kPa O 2 , oxidation produced raised oxide nodules on pockets with nanocrystalline SiC. Oxidation mechanisms were determined using Atom probe tomography. Active SiC oxidation occurred in C-rich grain boundaries with low P O$_2$ , leading to SiO 2 buildup in porous nodules. Here, this phenomenon was not observed at any temperature in 20 kPa O 2 environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Assessment of EPRI’s Tan Delta Approach to Manage Cables in Submerged Environments: Statistical Review of EPRI Data

Research conducted by the Electric Power Research Institute (EPRI) and other research institutions have concluded that water-trees are one of the leading degradation mechanisms that contribute to the loss of dielectric insulation strength in medium-voltage cable insulating materials in wet or submerged environments. The electrochemical reactions are caused by the combined effect of water presence and relatively high electrical stress. Records of cable failures provided by the licensees in response to Generic Letter (GL) 2007-01 (NRC 2007) have called into question the reliability of medium voltage cables in wetted or submerged environments. EPRI’s dissipation factor or Tan Delta testing guidelines and acceptance criteria have been adopted by most nuclear power plant operators as the primary tool for condition monitoring of medium voltage cables in wet or submerged environments. EPRI has been collecting member data since late 2009 to analyze and provide feedback to members, validate the EPRI-developed acceptance criteria guidelines, support analysis of test results, recommend appropriate actions for the "action required" category, and gather candidate cables for EPRI-sponsored forensic research on causes for insulation degradation. EPRI has collected data from 37 nuclear sites, which represent 44 operating units. The test results have been organized by insulation type, such as cross-linked polyethylene (XLPE); butyl rubber; black, pink, and brown ethylene-propylene rubber (EPR); and compact insulation (black and pink EPR)1. The data have been analyzed, and follow-up information was obtained from members for “action required” test results. EPRI has also performed correlations between Tan Delta tests and the information gathered under the EPRI forensic research on medium-voltage cables. In addition, EPRI has developed guidance by cable insulation type on how to systematically analyze Tan Delta test results. The analysis described here reviewed the two primary EPRI reports (EPRI 3002000557 and EPRI 3002005321) as well as two precedent EPRI reports (EPRI 1028262 and EPRI 1021070) that were cited in the primary reports. The principal results and conclusions from the project analyses are provided.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1978

Lithium and potassium heat pipes for thermionic converters

A prototypic heat pipe system for an out-of-core thermionic reactor was built and tested. The emitter of the concentric thermionic converter consists of the condenser of a tungsten heat pipe utilizing a lithium working fluid. The evaporator section of the emitter heat pipe is radiation heated to simulate the thermal input from the nuclear reactor. The emitter heat pipe thermal transport is matched to the thermionic converter input requirement. The collector heat pipe of niobium, 1% zirconium alloy uses potassium as the working fluid. The thermionic collector is coupled to the heat pipe by a tapered conical joint designed to minimize the temperature drop. The collector heat flux matches the design requirements of the thermionic converter.

Miskolczy, G.↗
DOE OSTI2016

Experimental study on the guillotine break accident of the main steam line with Atlas

A main steam line break (MSLB) accident is defined as a pipe break in the main steam system of a nuclear power plant. The MSLB accident is characterized as cool-down event due to the increased steam flow rate, which causes excessive energy removal from the steam generators and eventually reactor coolant system. This results in a decrease in reactor coolant temperature. The cool-down causes an increase in core reactivity due to the negative moderator and Doppler reactivity coefficients. The thermalhydraulic integral effect test was performed to simulate a double-ended guillotine break accident of the main steam line of a pressurized water reactor. With an aim of simulating the MSLB accident as realistically as possible, a pertinent scaling approach was taken, especially focusing on a break flow. The main objectives of this test were not only to provide physical insight into the system response of a nuclear power plant during the MSLB accident but also to produce an integral effect test data to validate the safety analysis codes. In the present test, asymmetric thermal-hydraulic behavior was clearly observed between the loops. In the cold leg-1A and -1B where the affected steam generator is connected, the flow rates were higher and the fluid temperatures were lower than those in the cold leg-2A and -2B. This could be attributed to the excessive heat removal in the cold leg-1A and -1B through the affected steam generator. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes. Furthermore, this data can be utilized to identify any code deficiency for a MSLB simulation, especially for DVI-adapted plants. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Advances in Westinghouse fuel rod design technology (PAD5)

The Westinghouse Performance Analysis and Design (PAD) code for light-water reactors has been recently updated to version PAD5 and submitted to NRC for review. PAD5 incorporates extensive updates in fuel performance and material property models, including fuel thermal conductivity, pellet fragment relocation, fission gas release, fission gas swelling, and cladding creep and growth. Advances in thermal models are mainly in pellet fragment relocation, radial power distribution within the pellet, and the changes in fuel thermal conductivity due to changes in temperature, burnup and fuel porosity. The PAD5 thermal fission gas release model is a phenomenological model based on the irradiation dependence of the gas release from mechanistic processes for the diffusion of the fission gas generated in the fuel grains to the grain boundaries, formation of fission gas bubbles on the grain boundaries and release of the fission gas when these grain boundary bubbles inter-link. The PAD5 cladding creep model is predominantly a combination of the following mechanisms: thermal creep, irradiation induced creep, and irradiation enhanced creep. The fuel performance database used for the development, calibration and testing of the PAD5 code covers a broad range of both commercial irradiation data and test rod data and requires continuously additional data focusing on the latest development of the PAD5 code. Test programs are also included in the database. These data are checked and stored in a hierarchical database structure that supports the automation of model development and calibration. New data, once incorporated into the database, can be quickly evaluated with respect to key fuel and cladding performance models to confirm expected trends or to evaluate potential model improvements. This automated calibration system has been integrated with state-of-the-art optimization packages in DAKOTA Analysis Kit to provide for advanced statistical assessments and model coefficient optimization. Calibration results are automatically tabulated and plotted to provide easy assessment of trends or new behaviors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MSR Salt Spill Accident Testing Using Eutectic NaCl-UCl 3

Assessing the potential consequences of identified accident scenarios is an essential part of the licensing process for a new nuclear reactor and is achieved by using accident progression models. A likely accident scenario that will be evaluated by developers of molten salt reactors (MSRs) is a spill of radionuclide-bearing fuel salt onto the reactor containment floor (i.e., a salt spill accident). Models to determine the consequences of a molten salt spill accident need to be parameterized and validated using experimental data, but little experimental effort has been dedicated to fill these data gaps to date. Specifically, data is needed to enable model development for individual processes (e.g., spreading, heat transfer, corrosion, radionuclide vaporization, and aerosol generation) that quantifies the sensitivities towards the initial conditions of the spill, the ambient environment, and the features of the containment. In addition, integrated experiments that simulate salt spill accidents will need to be conducted to provide insight into coupled processes and data for model validation, but these experiments will require the use of proven methods to quantify the processes under evaluation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Challenges in practical button cell testing for hydrogen production from high temperature electrolysis of water

High temperature electrolysis of water using solid oxide electrochemical cells (SOEC) is a promising technology for hydrogen production with high energy efficiency and may promote decarbonization when coupled with renewable energy sources and excess heat from nuclear reactors. Over the past several decades there have been extensive scientific and engineering studies on cell materials and degradation behaviors that have greatly improved current density, decreased total resistance, and lowered degradation rates. Although the technology is now at a near-commercial level, maintaining consistency in cell testing and minimizing variance in practical testing environments is an often overlooked but crucial topic. To promote high quality data collection, testing procedures and balance of plant component details are extremely important to consider. This work discusses some key factors affecting the reproducibility of practical SOEC testing on the button cell level, namely, current collection layers, cell sealing procedures, the reliability of steam and hydrogen delivery systems, cell testing fixture design, and reduction procedures. To provide a baseline and a level of standardization for the SOEC community, this work also discloses details of the standard operating procedure and techniques adopted for o-SOEC testing at Idaho National Laboratory (INL).

08 HYDROGEN↗
DOE OSTI2017

Counterpart Test on IBLOCA of 13% Cold Leg Break in ATLAS Facility

In the analysis for Loss-of-Coolant Accident (LOCA) as a design-basis accident (DBA) of a pressurized water reactor (PWR), redefinition of the break size based on the risk-informed regulation (RIR) is being extensively investigated due to the potential safety benefits. The pipe integrity and break frequency has been found to depend on the pipe size, so that a complete rupture of a smaller pipe such as a pressurizer surge-line or an emergency core cooling system (ECCS) nozzle is more probable than a double-ended guillotine break (DEGB) of primary coolant circuit piping. The scenario with the break of these pipes can be regarded as intermediate break (IB) LOCA. With an aim of expanding code validation to address safety issues of the IBLOCA and developing an evaluation model (EM) for full spectrum LOCA, the database for the IBLOCA test in an integral effect test (IET) facility is essential. In particular, an experimental data from the counterpart test can contribute to enhance our understanding on the phenomena and resolve the scaling issues in the scaled-down test facilities. The objective of this study is to investigate thermal hydraulic phenomena during the IBLOCA by the counterpart test for the 13% cold leg break in ATLAS facility, which was the A.2 test in the OECDATLAS project. This integral effect test would contribute to evaluate the scaling effect of the ATLAS design and test condition.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗