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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 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

ACCELERATED THERMAL AGING OF HARVESTED HYPALON JACKET FOR REMAINING USEFUL LIFE DETERMINATION AND DIAGNOSIS

For nuclear power plants (NPPs) that are considering second license renewal for their operation beyond 60 years, knowledge of long-term operation, condition monitoring, and viability for the reactor components including reactor pressure vessel, concrete structures, and cable systems is essential. Such knowledge will provide NPPs with a game plan to predict performance and to estimate costs that are associated with monitoring or replacement programs for the affected systems. For cable systems that encompass a wide variety of materials, manufacturers, and in-plant location, accelerated aging of harvested cable jacket and insulation can provide insight as to remaining useful life and methods for monitoring. Accelerated thermal aging in air at temperatures between 80°C and 120°C was carried out on a multi-conductor control rod cable that had been in-service for over 30 years and was made by Boston Insulated Wire with Hypalon cable jacket and ethylene-propylene rubber insulation. From elongation at break (EAB) measurements and supporting Arrhenius analysis of the jacket material, an activation energy of 97.84 kJ/mol was estimated and the time to degradation, as represented by 50% EAB at the expected maximum operating temperature of 45°C, was estimated as consistent with 80 years. These values were slightly below previous measurements and could either be attributed to in-service degradation or variations in material properties from a given manufacturer. In addition, results from indenter modulus measurements and Fourier transform infrared spectroscopy suggest possible markers that could be beneficial to monitor cable conditions.

Glass III, Samuel W.↗
DOE OSTI2018

INTER-LABORATORY RESULTS AND ANALYSES OF MINI-C(T) SPECIMEN TESTING OF AN IRRADIATED LINDE 80 WELD METAL

An irradiated low-upper-shelf Linde 80 weld metal has been tested by four laboratories as part of an inter-laboratory assessment of use of the miniature compact tension [mini-C(T)] test specimen for Master Curve fracture toughness evaluation following ASTM E1921. The preliminary results from each of the laboratories have been compiled and evaluated together to assess the validity and use of the mini-C(T) specimen for an irradiated reactor pressure vessel material which can exhibit ductile crack growth at low temperatures relative to cleavage initiation fracture toughness. The preliminary results from this mini-C(T) testing can also be compared to extensive specimen test results from larger C(T) specimens of the same irradiated material. Comparisons of the results from each of the laboratories and some inter-laboratory differences in the fracture testing are assessed. The evaluations indicate reasonable agreement between the mini-C(T) and larger specimen results, but the selection of test temperature and the number of test specimens needed to obtain reliable results are more difficult when testing a low-upper-shelf toughness material.

Server, W. L.↗
DOE OSTI2019

Self-powered Wireless Through-wall Data Communication for Nuclear Environments

In the nuclear industry, many vital components, such as spent fuel storage canisters and nuclear reactor pressure vessels (RPV), are entirely enclosed by metal and surrounded by thick concrete walls that manage the potentially harmful radiation and prevent release to the environment. Due to the casks’ long storage, monitoring temperature, pressure, radiation, humidity, structural health, etc., within these enclosed vessels is crucial to ensure the fuel containment safety and security. In this paper, a self-powered wireless through-wall data communication system for nuclear environments was designed. The whole package includes: a radiation energy harvester with power management; ultrasound wireless communication using high-temperature piezoelectric transducers; electronics modules for harvesting, sensing, and data transmission; and radiation shielding for electronics and sensors. The package is able to harvest tens to hundreds of mW power from a nuclear canister environment directly and provides a path forward for continuous monitoring of the dry cask for 50 years.

Wu, Yongjia↗
DOE OSTI2022

LWR Aging Management for Life Beyond 80

While several decades of materials degradation research enabled the current US fleet of Light Water Reactors (LWR) to plan to produce electrical power through extended operations up to 80 years of plant life, it now seems appropriate to evaluate what issues, methods, and timelines need to be considered to meet the expected electric power demands for life beyond eighty (LBE) years. To address these questions, it will be necessary to review the electrical capacity and the projected LWR fleet capacity including new builds and advanced reactors. This effort should begin with a materials degradation assessment focused on known and possible unknown issues using a reduced Expanded Materials Degradation Assessment [1], which was employed to identify knowledge gaps for a second license renewal (60-80 yrs.). The focus of the LBE assessment should include: Establishing a timeline to initiate assessments of possible materials aging issues and to ensure sufficient time to evaluate materials degradation research gaps; assessing components and materials to be evaluated, e.g., metals (reactor pressure vessels and alloys within and outside the pressure boundary), concrete, cables, as well as mitigation methods, advanced monitoring, and validation with ex-service materials; initiating a LBE research plan focused on developing an expanded mechanistic understanding of materials degradation and refined models through Codes and Standards evaluation for use by the nuclear industry; and maintaining and strengthening the nuclear materials human knowledge base to address new and as yet unknown degradation modes.

Chen, Xiang↗
DOE OSTI2023

ADDITIVE MANUFACTURING AND REGULATORY TESTING OF CANISTERS FOR SPENT NUCLEAR FUEL MANAGEMENT

The back end of the fuel cycle focuses on the interim storage, transportation, and final disposition of the spent nuclear fuel from nuclear reactors. Commercial light water nuclear power stations across the United States operate with fuel assemblies being irradiated for up to 6 years (planned) in the reactor pressure vessel. After their planned irradiation, the fuel assemblies are moved to a spent fuel pool within the facility complex. After the spent nuclear fuel is removed from the fuel pool, it is typically inserted into a welded metal canister that can be transferred between overpacks for storage, transportation and possibly disposal. Most dry storage systems being used by industry today use dual purpose canisters (DPCs), designed for use in storage and transportation overpacks, but not specifically designed for disposal. Triple purpose-canisters, designed for disposal in addition to storage and transportation, have also been researched. Traditional manufacturing methods for spent fuel canisters involve fusion welding along the length or circumference of the canister which results in high tensile residual stresses in the joint weld zone (WZ) and heat affected zone (HAZ). In this paper, spent fuel canister designs were printed by wire arc additive manufacturing (AM) using the 316L SS welding wire to demonstrate: 1. Feasibility of spent nuclear fuel canister fabrication using this advanced manufacturing method, 2. Dynamic response of the additive manufacture canister design when subjected to the federally mandated Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) physical tests for Type B packages. This paper will focus on the canister printing design and structural tests. The AM 3D printed design, regulatory testing, and post-test evaluation of the canister tested to the 10 CFR 71.71 and 10 CFR 71.73 requirements will be presented. One AM canister design was subjected to the penetration, free drop, and puncture test. Before and after the dynamic structural tests, the AM canister design was scanned with a handheld scanner to capture a 3D CAD geometry to compare to the 3D printed canister design in the deformed shape. The scanned geometry was sectioned in areas with deformation and the cross-section profile was measured to determine accurate and repeated results of the deformed shape of the AM canister design.

Martinez, Oscar↗
DOE OSTI2015

Reactor pond walls deposit investigation during dismantling MR reactor - 15028

Specialists of the NRC 'Kurchatov Institute' now carry out the dismantling of the multiloop research MR reactor. During this works the water level was decreased in the MR reactor pond on 4 meter. Dependence of exposure dose rate above the pond from deposits on the pond walls was confirmed. Detailed research of deposits on MR reactor pond walls was made, using α-, β-, γ- spectrometric methods. Correlation coefficients between concentration Pu and U radionuclides and concentration {sup 137}Cs and {sup 241}Am was calculated. The surface activity of the main dose-forming radionuclide {sup 137}Cs on the walls of the pond was measured. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Remediation of Irradiated Graphite Using Microorganisms Isolated from Oil Contaminated Soil - 19643

The deployment of nuclear energy globally is hindered by the fear cleared by nuclear reactor disasters such as the Chernobyl and Fukushima reactor meltdowns. Additionally, accumulation of radioactive waste globally will, in the near future, present a significant threat to humans and the environment. Notably, the majority of nuclear waste from the energy generation industry is generated as Low Level Waste (LLW) in which the small radioactive component renders the bulk of the waste unsuitable for conventional treatment and disposal. In the case of radioactive graphite, the small radioactive component comprising radiocarbon-14 (C-14) and metallic pollutants can be removed to allow safe disposal or reuse as of the graphite. In the current study, mixed-cultures of organisms isolated from oil contaminated sites showed potential of selective bioaccumulation of C-14 from a powdered C-12/C-14 graphite matrix. The study was conducted in a continuously mixed reactor (growth vessel) constructed from Pyrex panels (30 cm x 20 cm x 15 cm) with finely ground graphite as the principal carbon source (1500-2000 mg.L{sup -1}) and spikes of glucose (500 mg.L{sup -1} every 48 hours). The growth vessel was connected to a biofilter unit by submersible pump where particulate matter was isolated for further analysis. Results showed that approximately 68% more C-14 was released from ignited dry cells than from respiration product. C-14 accumulated in cells corrected using a trap filter system, with a marked decrease of C-14 in carbon dioxide than vegetative suspended cells in the main tank. The only mechanism that could account for the differential C-14 distribution between the liquid medium and cells was possible fractionation during uptake of carbon compounds. The discriminating factor suggested in this study is the kinetic isotope effect (KIE) which allows heavier atoms to have different affinities for reaction sites on catalysis and biocatalysis. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Tensile Behavior of Inconel X750 Sheets: Effect of Heat Treatment

X-750 is a nickel-chromium based super alloy of usefulness in a wide variety of applications. Its excellent relaxation resistance is useful for high-temperature springs and bolts; it is utilized in gas turbines, rocket engines, nuclear reactors, pressure vessels, tooling, and aircraft structures. Its performance is similar to that of Inconel alloy 600 but it is made precipitation-hardenable by additions of aluminum and titanium. It has good resistance to corrosion and oxidation along with high tensile and creep-rupture properties at temperatures up to 700 deg. C. In this study, the tensile flow behavior of the alloy was studied by performing uniaxial tensile tests under a range of temperatures and at different strain rates. To assess the microstructure-mechanical property relationships in the alloy, microstructure analysis (through the use of electron microscopy and XRD analysis) was utilized post-experiment. This paper will however focus on reporting the tensile properties and temperature and strain rate effects on the as-received alloy vs. the heat-treated alloy. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2023

Adaptive, Active Learning, and Multifidelity Monte Carlo Methods in the MOOSE Stochastic Tools Module

MOOSE is an open-source computational platform for constructing multi-physics models and executing them in a massively parallel fashion. It has a stochastic tools module (STM) for forward/inverse uncertainty quantification (UQ) and surrogate modeling. This presentation details some recent developments to the STM with respect to the implementation of adaptive, active learning, and multifidelity Monte Carlo methods for forward UQ of computational models. Specifically, the adaptive Monte Carlo methods include Markov Chain Monte Carlo (MCMC)-driven algorithms like adaptive importance sampling and parallelized subset simulation for statistical QoI estimation, rare events analysis, and stochastic gradient-free optimization. The active learning methods include Gaussian Process (GP) surrogates and their training via Adam optimization, design of acquisition functions, and integration with samplers like Monte Carlo, adaptive importance, and parallelized subset simulation. These active learning methods are also designed to work in a batch mode, wherein, the required calls to the full computational model are executed in parallel whenever a user-specified batch size is met. The multifidelity methods in STM are broadly divided into two categories: hierarchical, where a defined hierarchy exists among the low-fidelity models, and peer, where all the low-fidelity models are treated equally. A GP surrogate is used to learn the differences between the low- and high-fidelity models in both multifidelity categories, and acquisition functions from the active learning classes are used to decide whether to rely on a low-fidelity model or call the expensive high-fidelity model. Alongside the software description and usage, applications are also presented to nuclear engineering computational models including a TRISO nuclear fuel particle, a reactor pressure vessel, and a heat-pipe microreactor.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Preliminary Analysis of a Nuclear Grade Sandwich Composite for Transportable Micro-Reactor Shielding

Transportable microreactors pose a unique shielding challenge compared with conventional reactors. Many vendors are seeking to transport a microreactor to a site, operate it for a number of years, and then transport it to a secondary site. This involves ensuring the reactor has sufficient shielding during operations and post-irradiation transportation to ensure the safety of those around it, all while ensuring it can be transported via traditional means. This work examines the use of a nuclear grade sandwich composite (NGSC) which seeks to combine the reactor pressure vessel and biological shielding into a single unit. Through this work, it was determined that the NGSC was sufficient to adequately reduce neutron and gamma current exiting the structure to warrant further investigation as a potential shielding option for microreactors.

42 ENGINEERING↗
DOE OSTI2017

Accelerated Thermal Aging of Harvested Hypalon Jacket for Remaining Useful Life Determination and Diagnosis

For nuclear power plants (NPPs) considering second license renewal for their operation beyond 60 years, knowledge of long-term operation, condition monitoring, and viability for the reactor components including reactor pressure vessel, concrete structures, and cable systems is essential. Such knowledge will provide NPPs with a game plan to predict performance and to estimate costs that are associated with monitoring or replacement programs for the affected systems. For cable systems that encompass a wide variety of materials, manufacturers, and in-plant location, accelerated aging of harvested cable jacket and insulation can provide insight as to remaining useful life and methods for monitoring. Accelerated thermal aging in air at temperatures between 80°C and 120°C was carried out on a multi-conductor control rod cable that had been in-service for over 30 years and was made by Boston Insulated Wire with Hypalon™ cable jacket and ethylene-propylene rubber insulation. From elongation at break (EAB) measurements and supporting Arrhenius analysis of the jacket material, an activation energy of 97.84 kJ/mol was estimated and the time to degradation, as represented by 50% EAB at the expected maximum operating temperature of 45°C, was estimated to be 80 years. These values were slightly below previous measurements on similar BIW Hypalon cable jacket and could either be attributed to in-service degradation or variations in material properties from process variations. In addition, results from indenter modulus measurements and Fourier transform infrared spectroscopy suggest possible markers that could be beneficial to monitor cable conditions.

cable aging, cable insulation, Hypalon, second lic↗
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↗
DOE OSTI2025

Material Selection and Heat Transfer Model for PTUHS Device

The Path to Ultimate Heat Sink Device (PTUHS) is a passive safety system that utilizes a radiative heat transfer valve to control the rate at which heat is removed from a nuclear reactor pressure vessel (RPV) and dispersed into surrounding soil. The material selection for the PTUHS device is investigated, where the thermophysical properties are either maximized or minimized bases on what section of the PTUHS that material is being used for. Final recommendations of material choice are then presented. The heat transfer model of the PTUHS device is solved in ABAQUS, where a normal operating conditions and SCRAM conditions are solved. These temperature maps and heat flux of the system show the system's ability, where minimal heat is lost during normal operating conditions and system failure in a SCRAM scenario due to temperatures that were reached.

42 - ENGINEERING↗
DOE OSTI2026

STANDARD REFERENCE MATERIAL CHARPY TESTING FROM HIGH FLUENCE SURVEILLANCE CAPSULE

The Palisades Nuclear Generating Station included in its original surveillance program a surveillance capsule, designated A-60. The capsule was removed from its surveillance position in early 1995 and has been resident in the spent fuel pool since that time. It was harvested to perform characterization of surveillance specimens in this capsule in 2023. This capsule was irradiated to a fluence of 1.96x1020 n/cm2 (E> 1MeV) that is equivalent for more than 150 effective full power years for the current US reactor pressure vessel (RPV) fleet. This capsule contained several materials, including Charpy specimens of standard reference material (SRM) from Heavy-Section Steel Technology (HSST) A533-B Plate 01. To perform Charpy testing of this highly irradiated material, the new Charpy specimen transfer system was designed and implemented on Charpy impact testing machine in the hot cell to accommodate remote testing of these specimens. This new transfer system includes integrated environmental chamber to cool or heat Charpy specimens to the desired temperature. Testing of this highly irradiated material revealed very large shift of Charpy transition temperature, 188oC.

Sokolov, Mikhail [ORNL] (ORCID:0000000256373346)↗
NASA NTRS1981

Wireless, in-vessel neutron monitor for initial core-loading of advanced breeder reactors

An experimental wireless, in-vessel neutron monitor was developed to measure the reactivity of an advanced breeder reactor as the core is loaded for the first time to preclude an accidental critically incident. The environment is liquid sodium at a temperature of approx. 220 C, with negligible gamma or neutron radiation. With ultrasonic transmission of neutron data, no fundamental limitation was observed after tests at 230 C for 2000 h. The neutron sensitivity was approx. 1 count/s-nv, and the potential data transmission rate was approx. 10,000 counts/s.

Delorenzo, J. T.↗
DOE OSTI2018

Remove Welding Residual Stress for CFETR Vacuum Vessel by Trailing Ultrasonic Impact Treatment

The China fusion engineering test reactor (CFETR) vacuum vessel is welded by narrow gap TIG (NG-TIG) welding, and the welding residual stress of the CFETR vacuum vessel can be redistributed by trailing welding ultrasonic impact treatment. In order to investigate the feasibility of the residual stress removing scheme, and to obtain the optimal trailing ultrasonic impact treatment technological parameters in the process of removing welding residual stress, a welding model that similar to vacuum vessel welding seam is established by using ABAQUS software, a NG-TIG welding heat source subroutine which is written in FORTRAN language used to simulate NG-TIG welding. According to the welding simulation results, a trailing welding ultrasonic impact treatment model is established, and the effects of the impact pin number, the impact method, the impact pin diameter and the impact frequency on welding residual stress are studied. The results show that the longitudinal residual stress in welding seam and its adjacent area and the lateral residual stress in the whole region have been obviously decreased by different trailing welding ultrasonic impact processes, and have made the tensile stress in the welding seam and its adjacent area has been changed into compressive stress, which can provide theoretical guidance and reference for actual production.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Systems and methods for removing components of a gas mixture

A system for removing components of a gaseous mixture is provided comprising: a reactor fluid containing vessel having conduits extending therefrom, aqueous fluid within the reactor, the fluid containing a ligand and a metal, and at least one reactive surface within the vessel coupled to a power source. A method for removing a component from a gaseous mixture is provided comprising exposing the gaseous mixture to a fluid containing a ligand and a reactive metal, the exposing chemically binding the component of the gaseous mixture to the ligand. A method of capturing a component of a gaseous mixture is provided comprising: exposing the gaseous mixture to a fluid containing a ligand and a reactive metal, the exposing chemically binding the component of the gaseous mixture to the ligand, altering the oxidation state of the metal, the altering unbinding the component from the ligand, and capturing the component.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Gravity-based, non-invasive reactor system and method for coolant inventory monitoring

A gravity-based, non-invasive method of measuring a level of fluid in a container comprises use of at least one gravity meter located as proximate a center of mass of the fluid as possible. In a nuclear reactor system a method for monitoring the level of fluid in a nuclear reactor module, a report of a loss or gain of fluid within a cylindrical module may be generated from capturing a time series of gravity data from a first gravity meter mounted as an upper gravity meter and a second gravity meter mounted as a lower gravity meter, for example, proximate a cylindrical nuclear reactor module so as not to require any invasive conduit through, for example, a containment pressure vessel (CPV) or a reactor pressure vessel (RPV). In one embodiment, the upper and lower gravity meters are mounted on stable mounts as close to the fluid in the module as possible within a coolant pool or a structure containing cooled air. If a coolant pool of water surrounds a nuclear reactor module, the meters may be housed within a dry housing in the coolant pool such that the meters may be accessed from above the coolant pool and are located as close as possible to the reactor module and its contained mass of fluid.

Bloom, Scott Harris↗