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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

GSI 191 Closure Issue for EDF PWR Fleet: The Fuel Assembly Test Results

EDF has re started recirculation studies within the international GSI 191 framework by 2002, when EDF decided to join the NEI sump strainers task force. All the EDF PWR fleet have been upgraded, with major strainer modifications, according to the NEI 04-07 guideline. The so-called chemical effects have been carefully studied by EDF and demonstration have been released that there are no chemical effects on EDF fleet, owing to the nature of our debris. The study of strainer downstream effects started by 2009, with decoupling Downstream Sump Debris Source Term (DST) and by the end of 2016, all the equipment have been successfully qualified. At a previous ANS meeting,it was said that core fuel tests were scheduled. They have started at the beginning of 2014 and they have been closed by mid 2016 for AREVA Fuel type at ERLANGEN facility (Germany). The present presentation is aimed at describing the test facility, the test specifications and the main results. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Radiation Testing for Nuclear Inspection Systems (Final CRADA Report)

Higher-radiation hardened video cameras are needed in the operation and remote handling of equipment in nuclear reactor inspection and refueling applications. Vega Wave Systems has developed a radiation-hardened vision system for Nuclear Energy applications. The high-radiation hardness testing of this vision system was conducted using the Argonne Low-Energy Accelerator Facility (LEAF).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

AFC FY 2023 HFIR Irradiation Test Matrix – Supported by the Design of a Ring Specimen Irradiation Vehicle

The development and deployment of accident-tolerant cladding materials in light water nuclear reactors requires the collection of neutron-irradiated properties of these materials with specific consideration to the material properties orientation (i.e., tube geometry). To that end, a test plan to irradiate several cladding tube and ring specimens in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was proposed, along with the development of a new capsule design that features ring specimens with a preexisting gauge section. This ring specimen geometry removes the need to machine a gauge section on the irradiated specimen and thus allows for direct application of a ring-pull test post-irradiation. The test matrix includes uncoated and chromium-coated Zircaloy-4 and advanced zirconium alloy cladding specimens, pre-characterized at ORNL. Plans are to irradiate the specimens at about 4 and 10 dpa—corresponding to two and five cycles at the HFIR centerline—with a target specimen gauge section temperature of 330°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

In Situ TEM Microcompression Pillar Size Effects in Fe-9Cr ODS

The objective of this paper is to ascertain the existence of a specimen size threshold, below which the material no longer exhibits bulk mechanical behavior when tested using in situ transmission electron microscopic (TEM) micro-compression pillars. In this work, sub-micron sized pillars of a model Fe-9Cr oxide dispersion strengthened (ODS) alloy were compressed in TEM in situ at room temperature, and attempts have been made to quantify the resulting data. Results are compared to mechanical property values in the archival literature, and a discussion on the validity of quantitative in situ TEM micro-compression data for ODS steel is included. In situ TEM mechanical testing is a relatively new technique enabling concurrent TEM imaging and mechanical testing of sub-micron-sized electron-transparent specimens. Due to limited sample volume of many nuclear materials, in situ TEM mechanical testing holds great promise for advancing our understanding of deformation phenomena. But it is well known that the dramatically reduced specimen size required for in situ testing can affect quantitative and qualitative deformation behaviors. However, Kiener, et al. used in situ TEM micro-compression pillars on irradiated and unirradiated single-crystal copper. They observed size independence in the irradiated material, while the unnirradiated material displayed a size threshold below which the mechanical properties deviated from those of bulk Cu. They assert that the more densely-spaced obstacles in irradiated Cu allow it to display bulk properties even in specimens having ∼ 100 nm dimensions. Similar to Kiener's findings, we hypothesize that unirradiated ODS will demonstrate a size threshold, but the in situ TEM micro-compression technique can be extended from single crystal materials to engineering alloys. An Fe-9Cr model ODS alloy was chosen for this study because of the broad interest in ODS for advanced nuclear reactors. ODS alloys exhibit high irradiation tolerance and high temperature creep resistance. New insight into deformation processes gained through in situ TEM mechanical testing could be valuable for the validation of ODS alloys for next generation nuclear reactors. The fact that the compressive yield stress and elastic modulus do not follow distinct trends as a function of specimen size is likely due to the high density of obstacles to dislocation motion. The ODS microstructure contains such obstacles as oxide nanoparticles, a fine grain structure, a high density of dislocation lines, and carbide precipitates. Since the deformation behavior is governed by the minimum separation distance between obstacles [1], it is plausible that the high obstacle density enables ODS alloy micropillars to exhibit bulk-like mechanical properties even prior to irradiation. Future work will focus on: (1) developing finite element models to help explain the difference in elastic moduli between bulk and micropillar tests, (2) understanding the spread in yield stress measurements at fixed micropillar volumes, and (3) extending this study to irradiated ODS. (authors)

36 MATERIALS SCIENCE↗
NASA NTRS2009

Test Results From a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

The Brayton Power Conversion Unit (BPCU) located at NASA Glenn Research Center (GRC) in Cleveland, OH is a closed cycle system incorporating a turboaltemator, recuperator, and gas cooler connected by gas ducts to an external gas heater. For this series of tests, the BPCU was modified by replacing the gas heater with the Direct Drive Gas heater or DOG. The DOG uses electric resistance heaters to simulate a fast spectrum nuclear reactor similar to those proposed for space power applications. The combined system thermal transient behavior was the focus of these tests. The BPCU was operated at various steady state points. At each point it was subjected to transient changes involving shaft rotational speed or DOG electrical input. This paper outlines the changes made to the test unit and describes the testing that took place along with the test results.

Hervol, David S.↗
DOE OSTI2021

Testing to Evaluate Processes Expected to Occur during MSR Salt Spill Accidents

Obtaining a license for a new nuclear reactor requires the identification and assessment of the potential consequences of specified accident scenarios, which are achieved using accident progression models. Those models need to be parameterized and validated using experimental data, but existing experimental data addressing processes relevant to molten salt reactor accidents are sparse. Specifically, experimental data that quantify the sensitives of important processes to the initial conditions of the spill, the ambient environment, and the containment features are needed to parameterize individual process models. Integrated experiments that simulate accident scenarios are also needed to provide data for model validation, but these experiments will require the use of proven methods to quantify the processes under evaluation. The overarching objectives of this work are to develop the methods for simulating the targeted processes, to determine the effectiveness of the methods in producing the data required for model development, to generate data that can be used to parameterize individual process models, and to provide key insights into the behavior of spilled molten salt that should be considered in models. Experimental methods were designed to quantify aspects of individual processes expected to occur during or after a molten salt spill accident that will affect the fate of spilled molten salt and the radionuclides within. These processes include 1) molten salt spreading and heat transfer, 2) molten salt flowing and freezing in tubing, 3) stainless steel corrosion kinetics in molten salt, and 4) molten salt splashing and aerosol generation. The initial tests described in this report were conducted using eutectic FLiNaK to demonstrate the test methods, the data that are generated, and the analyses of the data to derive values needed for modeling. The primary variables that were tested include initial salt temperature and the presence of volatile surrogate fission products (e.g., cesium and iodine). The developed methods are shown to be effective in quantifying the desired processes and can be applied to study more complex salt compositions of interest to molten salt reactor developers, a wide range of environmental conditions of interest to modelers, and additional variables relevant to salt spill accidents. The developed methods and insights gained from laboratory tests can also be incorporated in future large-scale integrated tests used to simulate molten salt spill accidents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Radiation tolerance of piezoelectric bulk single-crystal aluminum nitride

For practical use in harsh radiation environments, we pose selection criteria for piezoelectric materials for nondestructive evaluation (NDE) and material characterization. Using these criteria, piezoelectric aluminum nitride is shown to be an excellent candidate. The results of tests on an aluminumnitride-based transducer operating in a nuclear reactor are also presented. We demonstrate the tolerance of single-crystal piezoelectric aluminum nitride after fast and thermal neutron fluences of 1.85 × 1018 neutron/cm2 and 5.8 × 1018 neutron/cm2, respectively, and a gamma dose of 26.8 MGy. The radiation hardness of AlN is most evident from the unaltered piezoelectric coefficient d33, which measured 5.5 pC/N after a fast and thermal neutron exposure in a nuclear reactor core for over 120 MWh, in agreement with the published literature value. The results offer potential for improving reactor safety and furthering the understanding of radiation effects on materials by enabling structural health monitoring and NDE in spite of the high levels of radiation and high temperatures, which are known to destroy typical commercial ultrasonic transducers.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Control And Optimization Modular Modeling Application For Nuclear Deployment

The purpose of the COMMAND code is to provide a flexible, scalable tool for use in developing, integrating, and testing the technologies necessary for achieving autonomous operations of advanced nuclear reactors. The code enables users to efficiently implement custom simulations and experiments by combining key methods from different software modules. These modules are focused on: modeling and simulation tools, such as nuclear simulation tools used for high-fidelity modeling (e.g., Reactor Excursion and Leak Analysis Program [RELAP5-3D] and Monte Carlo N-Particle [MCNP]); machine learning and optimization tools (e.g., anomaly detection and data-driven modeling techniques); advanced control in its digital, high-performance, and supervisory control forms (e.g., proportional integral derivative (PID) control and model predictive control (MPC); and integration with hardware through industrial communication protocols. To ensure flexibility and scalability, COMMAND was designed to be both modular—the software “pieces” all inherit from generic building blocks and can be combined and connected to create complicated simulations—and high performing—designed for parallel processing, enabling simulations and experiments to take advantage of multi-core computers, servers, and nodes. The code is written in the Python programming language due to the language's popularity, active community, and open-source and cross-platform nature. Maintaining consistency with other simulation tools used within the nuclear energy community, users implement simulations and experiments through text input files, which define components, parameters, connections, etc., through lines of text. Given that COMMAND is written in Python, these input files are native Python scripts, and so use the standard Python structure and formatting. This also enables users to take advantage of Python's extensive package library to develop custom capabilities for their specific use cases.

Faber, Jacob [Idaho National Laboratory (INL), Ida↗
DOE OSTI2016

Measured Thermal and Fast Neutron Fluence Rates for ATF-1 Holders During ATR Cycle 157D

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 157D which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains measurements of the fluence rates corresponding to the particular elevations relative to the 80-ft. core elevation. The data in this report consist of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, and (3) plots of both the thermal and fast neutron fluence rates. The fluence rates reported are for the average power levels given in the table of power history and distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

AGC 2 Irradiation Creep Strain Data Analysis

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

MCNP AND CFD MODELING OF A SODIUM FAST REACTOR SUB-ASSEMBLY CHANNEL TO CAPTURE LOCALIZED TEMPERATURE PEAKING

Typical evaluation tools for sodium-cooled fast reactors (SFR) use simplified neutronic/thermal models to efficiently compute the power and temperature distributions within the core. However, their modeling capability can be limited in extreme cases such as when sharp inter-assembly power gradients occur, or for non-standard SFR assembly geometries. Higher fidelity tools that rely on Monte Carlo methods for neutron transport, and Computational Fluid Dynamics (CFD) for thermal analysis, are better suited in these instances. The higher accuracy comes at the cost of higher computational expense. MCNP and STAR CCM+ were selected for this analysis. While MCNP is capable of generating pin-level power distributions for a whole-core SFR model, it is prohibitively expensive to then conduct CFD analysis for every single assembly. A semi-empirical model was therefore developed to rapidly estimate pin temperature distributions throughout the whole core. It relies on a geometric correction factor for turbulent flow-mixing. The correction factor is obtained from a CFD simulation of a single fuel assembly with a uniform power distribution and an adiabatic boundary condition. Without using CFD, each sub-channel outlet temperature is calculated separately using 1-D heat-balance and Nusselt number correlations and the MCNP calculated pin-powers. The rudimentary 1-D model is then corrected for the flow-mixing effect based on the CFD data. The temperature calculation is performed by a code written in the Python language called CFD Monte-Carlo Developed Reconstruction of Sub-channel at a Quasi-Representative Level (CMDR-SQRL). CMDR-SQRL was shown to adequately predict CFD evaluated temperature distribution for a wide range of SFR configurations, including an extreme case with fuel assemblies nearby containing moderating material. The agreement demonstrates how geometric and power-driven effects can be decoupled within an SFR assembly. CMDR-SQRL is intended to be mainly used for scoping studies to provide core-wide evaluations efficiently.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Development of safety requirements for HTGRs design - 18467

The safety requirements for the design of High Temperature Gas-cooled Reactors (HTGRs), which is analogous to the IAEA specific safety requirements (SSR) for Light Water Reactor (LWR) design (IAEA SSR-2/1), has been developed by the research committee established in the Atomic Energy Society of Japan so as to incorporate the HTGR safety features demonstrated by High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the coupling of the hydrogen production plants with nuclear plant. The safety design approach was determined to establish a high level of safety design standards by utilizing inherent safety features of HTGRs. The safety requirements is then developed using the objectives provisions tree methodology based on the safety design approach. The requirements for a structural integrity of fuel during accident conditions, for a confinement structure instead of a leak-tight pressure resistance containment vessel, for a passive residual heat removal from outside of reactor pressure vessel, etc., which are significantly different from those of LWR, are stated in the developed safety requirements. This paper describes overview of the proposed HTGR specific safety requirements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Probabilistic Analysis of Uncertainty in ATRC Flux Profiles

ATRC is a replica of the larger ATR design and is used to conduct research and obtain data such as flux measurements, excess reactivity, and loading requirements before being loaded into ATR. One method for determining the impact an experiment will have at ATR is by looking at the axial flux profile along the fuel rod in the corresponding ATRC experiment; however, flux wand measurements includes large amounts of variation which makes drawing conclusions from the data difficult. This poster describes a definitive method to propagate the uncertainty from ATRC measurements using Python code.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Overview of the US Nuclear Science User Facilities

Specialized and often unique and expensive instrumentation, adapted for use on radioactive materials, is essential to conduct research in the area of nuclear energy. Much of this equipment is located in facilities dedicated for such work that have been built and maintained over many decades. The Nuclear Science User Facilities (NSUF) was established in 2007 under the US Department of Energy Office of Nuclear Energy (DOE-NE) in order to bring these capabilities together with those innovative ideas from researchers at universities, national laboratories, small business, and industry that might not have regular access to reactors, hot cells, transportation casks, radiologically qualified instrumentation, and associated expertise. Early descriptions of the originally named Advanced Test Reactor National Scientific User Facility have been given. Through a competitively reviewed proposal process, the NSUF offers access to these capabilities and expertise at no cost to the user to promote research that will address the mission of DOE-NE. As a national organization comprised of a number of partner facilities across the US and located at National Laboratory, university, and industry sites, the NSUF supports materials science studies directed at irradiation effects in nuclear fuels and materials as well as in-pile and post irradiation examination (PIE) instrumentation development. The NSUF has created a Nuclear Fuels and Materials Library (NFML) of previously irradiated materials that users may access for their studies. Expanding the library for future applications of advanced technologies and novel concepts is an important objective of the NSUF. The NSUF has established and maintains the Nuclear Energy Infrastructure Database (NEID) as a tool for both DOE-NE and the nuclear energy research community. The extensive catalogue of capabilities is intended to aid DOE-NE, NSUF, and researchers more effectively and efficiently utilize the available capabilities. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Complete Non-Radioactive Operability Tests for Cladding Hull Chlorination

Non-radioactive operability tests were made to test the metal chlorination reactor and condenser and their accessories using batch chlorinations of non-radioactive cladding samples and to identify optimum operating practices and components that need further modifications prior to installation of the equipment into the hot cell for tests on actual used nuclear fuel (UNF) cladding. The operability tests included (1) modifications to provide the desired heating and reactor temperature profile; and (2) three batch chlorination tests using, respectively, 100, 250, and 500 g of cladding. During the batch chlorinations, metal corrosion of the equipment was assessed, pressurization of the gas inlet was examined and the best method for maintaining solid salt product transfer through the condenser was determined. Also, additional accessing equipment for collection of residual ash and positioning of the unit within the hot cell were identified, designed, and are being fabricated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗