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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 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Silicon Carbide Composite Cladding Qualification

This report examines the process to qualify a new fuel-cladding type focusing on the data requirements and types of irradiation testing necessary to meet the regulatory burden and related guidance applicable to fuel qualification of accident-tolerant fuel. The report considers the use of accelerated fuel qualification techniques and lead-test specimen programs that may shorten the timeline for qualifying fuel for use in a nuclear reactor at the desired burnup. The assessment framework particularly emphasizes the identification of key fuel manufacturing parameters, the specification of a fuel performance envelope to inform testing requirements, the use of evaluation models in the fuel qualification process, and the assessment of the experimental data used to develop and validate the empirical models and empirical safety criteria.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Overview of Virtual Test Bed FY23 Activities

The National Reactor Innovation Center (NRIC) was created for accelerating the deployment of novel reactor concepts. This will be achieved by providing both physical and virtual spaces for building and testing various components, systems, and complete pilot plants. The Virtual Test Bed (VTB) represents the virtual arm. It is being developed in collaboration with Department of Energy’s (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Evaluating alternate test techniques to characterize mechanical properties in nuclear-grade graphites - 18421

The deployment of graphite as a core structural material in the very high temperature reactor (VHTR) design presents a number of challenges with regard to accurate property and performance assessments. In the United States Advanced Reactor Technologies (ART) program, candidate VHTR graphite grades are being evaluated not only with respect to irradiated property values, but also with a focus on comprehensive data sets that explicate the property variability inherent to nuclear-grade graphites. The large data sets being compiled on baseline physical and mechanical properties provide a means not only to quantify variability within and between major graphite grades, but also to evaluate test techniques utilized on irradiated specimens that are severely limited in size and geometry. Detailed appraisals of emerging mechanical test techniques that either (a) fall outside of the prescribed test size and configuration constraints in order to be assembled into irradiation capsule such as the three-point bend test on sub-sized cylinders or (b) are not yet fully evaluated and standardized by ASTM International such as the Brazilian disc splitting tensile strength test can be more accurately generated when a broad comparison is possible that leverages multiple mechanical tests on full-size specimens using approved standards. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Evaluation of Machine Learning Models for Automated Data Analysis in In-Service Nuclear Power Plant Inspections

The commercial nuclear power industry is facing a potential shortage of certified nondestructive evaluation (NDE) analysts to meet future in-service inspection demands. Automated data analysis (ADA) currently supports human inspectors in tasks such as eddy current evaluations for steam generator examinations. Machine learning (ML) systems are nearing the capability to pass performance demonstration tests for ultrasonic testing (UT) inspections of reactor pressure vessel upper head penetrations in nuclear power plants (NPPs). Current research and development is focused on assisted analysis (AA) of ADA versus fully automated examinations. This presentation will cover assessment of ML flaw detection on dissimilar metal weld (DMW) piping joints.

36 MATERIALS SCIENCE↗
NASA NTRS1988

SP-100 flight qualification testing assessment

The SP-100 is a compact space power system driven by a nuclear reactor that provides 100 kWe to the user at 200 VDC. The thermal energy generated by the nuclear reactor is converted into electrical energy by passive thermoelectric devices. Various options for tailoring the MIL-STD-1540B guidelines to the SP-100 nuclear power system are discussed. This study aids in selecting the appropriate qualification test program based on the cost, schedule, and test effectiveness of the various options.

Jeanmougin, Nanette M.↗
DOE OSTI2023

Irradiation Behavior of Piezoelectric Materials for Nuclear Reactor Sensors

In-situ measurement of surface acoustic wave (SAW) resonators were used to characterize piezoelectric materials behavior in a nuclear reactor environment. Lithium niobate and aluminum nitride devices were tested up to 500°C temperature and 450 kW reactor power (1.9x10 12 n/cm 2 s). The resonant frequency responds to step changes in temperature and power. Materials’ response are inferred from the observed resonant frequency changes, in particular the kinetic behavior is used to determine mechanisms responsible for observed changes. The observed response is attributed to two mechanisms: temperature increase from gamma heating and accumulation of radiation-induced defects. Both of these mechanisms alter the physical properties of the piezoelectric materials, particularly the elastic constants. The demonstrated repeatability and reliability of SAW devices are attractive for sensor applications in extreme environments.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Optimization and characterization of a silicon photomultiplier-based ZnS(Ag) proton recoil fast neutron detector for nuclear fuel performance monitoring at TREAT

The restart of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory and consequent refurbishment of the Fuel Motion Monitoring System (FMMS), or Hodoscope, offers the opportunity to upgrade the detector system used for neutron imaging. Silicon photomultipliers (SiPMs) are a viable option for updating the Hodoscope to yield improved fuel monitoring capability. The Hodoscope uses ZnS(Ag) proton recoil scintillators (PRS) that provide good gamma-ray suppression and discrimination. Previous work showed that the Hamamatsu S13360-6075CS SiPM offers the best neutron detection and gamma-ray discrimination capability with the ZnS PRS. This work optimizes a SiPM-based detector and develops a PRS prototype for testing. Specifically, possible overvoltages for use are determined by confirming steady operation over extended measurement times. In addition, various SiPM-circuit implementations are tested to optimize the detector according to desired properties, and ultimately a PRS prototype is developed with modifiable components for versatile testing. Measurements of the neutron detection efficiency and gamma-ray rejection efficiency of SiPM-based PRS detectors and a reference PMT-based detector are also carried out. Neutron detection efficiency ranges between 1-2%, and detected gamma-ray rejection efficiency is on the order of 10 -7 . In conclusion, use of a low-pass filter only or a low-pass filter and 50-ω shunt resistor is recommended for the SiPM-based detector, and both configurations demonstrate improved performance over the PMT-based detector.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2017

Experimental Capability for Preparation and Purification of Molten Salts to Support Advanced Reactor Concepts, Nuclear Hybrid Energy Systems and Dynamic Grid Energy Storage Concepts

High-temperature multi-fluid, multi-loop test facility to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications is being developed at Idaho National Laboratory (INL). In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. The main objectives to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers at prototypical operating conditions, flow, and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for light water reactors at prototypical conditions. Fluoride salt coolants are eutectic binary or tertiary mixtures of fluoride salts with melting points in the range of 320 to 500 deg. C. Fluoride high temperature reactors have reactor outlet temperatures of 600 deg. C or higher for high-efficiency power generation or process heat applications. To address research needs related to salt coolants experimental capability has been added to study thermodynamic properties such as melting point temperature, heat capacity, and thermal conductivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Chemistry and Reactivity of Oxide Films on Aluminum Clad Test Reactor Fuels

This paper describes recent work performed to assess the degradation of aluminum spent nuclear fuel (ASNF) as part of an ongoing Department of Energy project. As part of a multi-task effort, this paper specifically describes efforts initiated in 2018 to better understand chemical phenomena and degradation in ASNF). In this paper experimental and modeling activities are described. The modeling effort presents initial results in two areas potentially influencing stability. Using an established electrolyte model based on experimental data, the solubility of aluminum as a function of temperature and pH was examined. This work describes interesting results for the minimum pH providing relevant data that was not available from the literature. It was found that solubility increases with temperature and that the minimum pH shifts to low pH with increase in temperature. The second area involves understanding the metallurgy of aluminum alloys during fuel drying. In this study, the temperature range from 100°C to 300°C was analyzed. Thermodynamic models predict that AA1100 will not experience significant phase changes while the more highly alloyed AA6061 and AA5052 may experience significant phase changes. These thermodynamic measurements establish a potential for phase changes during drying, however they do not predict the kinetics for such processes. Phase 1 of the experimental program has initiated and results are forthcoming. The test program will examine aluminum oxidation under anticipated dry storage conditions (temperature and humidity). Examination of localized corrosion due to tight crevices and galvanic contact (with stainless steel) will be included. Finally, water deliquescence will be assessed using thermogravimetric analysis with dynamic humidity profiles.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Thermal-hydraulic and Fuel Performance Scoping Studies of a Flowing Water Capsule in TREAT

The restart of the Transient Reactor Test (TREAT) facility has provided a much needed capability for integral transient testing of nuclear fuel. This testing is necessary to qualify new fuel concepts such as those developed under the Accident Tolerant Fuel (ATF) program, increasing the burnup limit of light water reactor (LWR) fuels, or a variety of other programs under the Advanced Fuels Campaign (AFC). The ATF campaign has been the main driver behind the development and implementation of a variety of capsules for TREAT experiments. The Separate-Effect Test Holder (SETH) was a dry capsule that enabled testing of ATF concepts under reactivity-initiated accident (RIA) heating conditions. Following SETH, the Static Environment Rodlet Transient Test Apparatus (SERTTA) was developed to enable RIA testing in a static water environment. In efforts to support the need for future Loss-of-Coolant Accident (LOCA) tests, the Transient Water Irradiation System for TREAT (TWIST) capsule has been developed that allows for water to drain from around the fuel rod and lower the pressure to simulate LOCA conditions. All these capsules that have been developed for LWR fuel testing all lack the capability for forced convective cooling which in some applications may limit their ability to test under prototypic conditions. A design and modeling effort has been started to modify the TWIST capsule by adding a flow tube and impeller that can create flowing coolant conditions for the fuel rod. RELAP5-3D and BISON models are being used to study the differences between RIA, LOCA, Anticipated Operational Occurrences (AOO), LWR power cycling, and other transient scenarios under flowing conditions capable in the flowing capsule and the current stagnant water capsules (SERTTA and TWIST). The scoping study will provide guidance on the needed capabilities for the flowing capsule and the limitations of the currently developed capsule for LWR testing in TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report

The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach has not been to create testing conditions that are fully prototypic of a fast reactor, but to create a thermal profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of limited fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The available evidence supports the assertion that use of cadmium-filtering creates irradiation conditions in the ATR that are both relevant to, and suitable for, fast reactor fuel performance testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Neutronic Safety Analysis of Pu-238 Production at Idaho National Laboratory

This analysis was completed to support the irradiation of plutonium fuel services (PFS) targets in the NEFT in the Advanced Test Reactor (ATR) as a part of the campaign to restart domestic production of plutonium-238 used in radioisotope power systems (RPS) by the National Aeronautical and Space Administration (NASA) and Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Nuclear Infrastructure Program (NE-3). Referred to as the PFS-ATR-GEN1-NEFT experiment, the assembly was designed to hold 46 PFS targets in the NEFT. The scope of this paper is to outline the MOPY method used to calculate the heat generation rates (HGRs), flux, fission density, and quantify the viability of the target design for Pu-238 production in ATR.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Development and experimental evaluation of an acoustic flow meter prototype for measuring air flow

Here, this study outlines the development and testing of a flow meter prototype specifically designed for use in nuclear reactor environments. The meter uses flow-induced vibration to measure flow rates and is designed for remote monitoring in harsh environments where radiation can damage electronics. The device generates a vibration signal produced by a feedback system when fluid flows out of an orifice and interacts with a downstream wedge. The frequency of the vibration signal corresponds directly to flow rate and velocity. The study investigates geometric parameters intrinsic to the prototype, such as orifice height, width, edge distance, and chamber length. Different configurations were also tested, including a bypass loop with variable cross-sections around the device. The investigation results reveal the impact of these geometric parameters on the vibration signal output, providing valuable insights for future design improvements and enhancing the effectiveness of flow metering systems used in nuclear reactor applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Dissolution of Material Test Reactor Fuel in an H-Canyon Dissolver

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO 3 flowsheet. Since the development of the existing flow sheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Developing Separate Effects Transient Test Experiments Using an Out-of-Pile Flowing Water Loop

In 2017 the Transient Reactor Test Facility was restarted after being placed in a standby state since 1994. This Test Reactor’s restart has since enabled the progressive development of new nuclear technologies within the U.S. that were previously required to outsource to other countries. While the reactor’s restart was a large feat worthy of recognition, the experimental use of its characteristics has required the further development of in-pile experimental infrastructure sufficient to support programmatic needs. This hardware have taken the form of capsule designs (compact and elongated) as well as loop concepts representing the phenomena of interest for a subset of separate effects tests desire for each respective testing campaign. The Transient Testing Program has a large-integrated effort that aligns with the U.S. Department of Energy’s current needs. This study compliments those programmatic elements by developing, fabricating and demonstrating a full-scale flowing water loop in an out-of-pile environment. The goal of this effort is to develop a pragmatic understanding of the engineering capabilities and limitations associated with geometric form-factors, metering technology, and controls logic under representative thermal hydraulic conditions that would be experienced within the TREAT reactor during an in-pile reactivity initiated accident test. The outcomes of this study result in an evaluation of the conceptual design of a comprehensive flowing water loop including objective figures of merit (FoMs) for comparing unique instrumentation and the basis for their selection during operations. These efforts directly contribute to and are required for the further advancement of transient testing capabilities within the U.S.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development and Validation of a Two-Phase Thermal-Hydraulic CFD Code NEK-2P

A project is underway to develop, verify and validate an advanced two-phase flow modeling capability for the highly-scalable, high-performance Computational Fluid Dynamics (CFD) code NEK5000. The goal of this work is to verify and validate the two-phase version of the NEK5000 code, named NEK-2P, to simulate the two-phase flow and heat transfer phenomena that occur in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. The NEK-2P two-phase flow models follow the approach used for the Extended Boiling Framework (EBF) previously developed at Argonne but include more fundamental physical models of boiling phenomena and advanced numerical algorithms for improved computational accuracy, robustness, and computational speed. The development of the NEK-2P two-phase solver and the implementation of the Extended Boiling Framework two-phase models were initially supported by Argonne National Laboratory (Argonne) through a Laboratory Directed Research and Development (LDRD) project during FY2014-2016. The development and validation of the two-phase models through analyses of selected two-phase boiling flow experiments was supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in FY2017-2020. This report focuses on verification and validation of the water-steam boiling model NEK-2P Two-Phase, CFD code. The NEK-2P was validated with Nuclear Power Engineering Corporation (NUPEC) Pressurized Water Reactor (PWR) Sub-channel and Bundle Test (PSBT) void distribution benchmark. Three different simulations were performed and analyzed for various operating conditions such as wall-heat flux and sub-cooled inlet temperatures. Reasonably good agreement with measured data was obtained in predicting the measured void distributions. Simulations were performed for Virginia Tech. (VT) 3x3 rod bundle geometry with and without spacers. The preliminary results were presented for Simplified Spacer Grid (SSG). In addition, the implementation of interface reconstruction model was tested with one of the Becker benchmark Critical Heat Flux (CHF) experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗