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At least 1,639 records · Page 91

Computational Nondestructive Evaluation Case Studies at NASA

The state of the art of numerical simulation of nondestructive evaluations (NDE) has begun to transition from research to application. The simulation software, both commercial and custom, has reached a level of maturity where it can be readily deployed to solve real world problems. The next area of research that is beginning to emerge is determining when and how to NDE simulation should be applied. At NASA Langley Research Center, NDE simulations have already been utilized for several aerospace projects to facilitate or enhance understanding of the inspection optimization and interpretation of results. Researchers at NASA have identified several different scenarios where it is appropriate to utilized NDE simulations. In this presentation, we will describe these scenarios, give examples of each instance, and demonstrate how NDE simulations were applied to solve problems with an emphasis on the mechanics of integrating with other workgroups. These examples will include inspection planning for multi-layer pressure vessels as well as on-orbit inspections.

Peter D. Juarez↗

A Combined Computational, Experimental, and Technology Development Approach to In-Space Laser Manufacturing Maturation at NASA Marshall Space Flight Center

In-space manufacturing (ISM) is emerging as a field vital to continued access and capabilities in the space environment. NASA Marshall Space Flight Center (MSFC) is advancing the frontier of in-space laser manufacturing (ISLM) techniques through work initially focused on maturing laser beam welding (LBW) and laser forming (LF) for use in space. Such techniques proffer the ability to assemble and join structures in space from sheet metal or other stock – extant satellites, in situ resource utilization of Lunar regolith, etc. – by forming to desired shapes and then joining via in-space welding (ISW). ISLM processes are useful for assembly, joining, modification, and repair of structures in free space and on the Lunar surface such as large observatories, antennas, trusses, blast/thermal/radiation shields, pressure vessels, and more. However, these techniques are not yet qualified & certified (Q&C) for regular application in space. It would be prohibitively expensive, laborious, and time-consuming to perform Q&C via traditional experimental approaches as data collection & experimentation in space is resource-intensive. As such, benchmark experiments and focused, properly instrumented technology demonstration efforts in space can collect sufficient data that – when combined with verified computational models in an integrated computational materials engineering (ICME) approach – can validate ICME tools capable of translating more readily obtained ground data to in-space, in situ, computationally informed Q&C of ISLM techniques. Several ISLM projects at MSFC are obtaining the data required to validate ICME tools through both ground and flight experiments. A parabolic flight experiment of LBW under vacuum is manifested for August 2024, including both microgravity and Lunar gravity profiles. This collaboration with the Ohio State University is investigating common aerospace alloys such as 316L stainless steel, 2219 aluminum alloy, and Ti64 titanium alloy. In situ data collection includes videography, thermography, and reference thermocouples to build a thermal model of the welds. This will elucidate the relevant physics when combined with post-flight microstructural examination and mechanical testing. MSFC is also progressing towards a suborbital flight experiment of LBW under vacuum, which could provide reams of data on ISW during sustained, high-quality reduced gravity. The effect of combined thermal (cryogenic and high-temperature) and vacuum exposure on both LBW (NASA-funded) and LF (DARPA-funded) is being investigated through ground experiments. In addition to the copious data collected during these ground experiments, ruggedization of LBW hardware will also be pursued. The datasets from these experiments will be used to validate computational models which will inform future ISLM efforts in an ICME framework. A variety of techniques across lengths scales, from CALPHAD-driven thermodynamics & kinetics to phase field modeling of solidification to kinetic Monte Carlo simulations of grain evolution at the mesoscale, will be employed to accelerate the infusion and eventual Q&C of LBW and LF for use in space. The development of data-driven surrogate models to bridge ground to flight experiments and thereby reduce the need for resource-intensive experiments in space will also be investigated. These ICME techniques, surrogate models, and datasets from ground testing can also be employed to advance manufacturing in terrestrial environments.

in-space welding↗

Developing a Pyrolysis Gas Thermal Blocking Model for Reentry Demise

In NASA’s Object Reentry Survival Analysis Tool (ORSAT), aerodynamic drag and aerothermal heating coefficients are computed for each of the free-molecular, continuum, and transitional flow regimes using analytical and semi-analytical methods. These heating coefficients were derived for typical metallic materials that melt and do not have a strong gas-phase contribution to the flow in the boundary layer. Modern satellites typically feature fiber-reinforced polymer (FRP) components, such as solar array booms, facesheets of sandwich panels, or overwraps for composite-overwrapped pressure vessels (COPV). These FRP materials do not behave the same as metals in the reentry environment, but instead will pyrolyze and develop significant volumes of gas into the boundary layer. Accurately predicting the reentry demise of FRP components is critical to assessing the reentry casualty risk for modern spacecraft. Research in recent years has shown that this demisability can depend heavily on how the expulsion of gaseous pyrolysis products through the outer surface of the material affects the heat flux at the surface. The ODPO has been developing a reduced-order model of the effect of pyrolysis gas blowing on the heat flux based on correlations between a blowing factor and a non-dimensional heat flux to be incorporated in the upcoming version 7.3 of the Object Reentry Survivability Analysis Tool (ORSAT). This presentation discusses the progress of this development project and the challenges remaining for generalizing the model across families of FRP materials.

Benton Greene↗

Submerged electricity generation plane with marine current-driven motors

An underwater apparatus for generating electric power from ocean currents and deep water tides. A submersible platform including two or more power pods, each having a rotor with fixed-pitch blades, with drivetrains housed in pressure vessels that are connected by a transverse structure providing buoyancy, which can be a wing depressor, hydrofoil, truss, or faired tube. The platform is connected to anchors on the seafloor by forward mooring lines and a vertical mooring line that restricts the depth of the device in the water column. The platform operates using passive, rather than active, depth control. The wing depressor, along with rotor drag loads, ensures the platform seeks the desired operational current velocity. The rotors are directly coupled to a hydraulic pump that drives at least one constant-speed hydraulic-motor generator set and enables hydraulic braking. A fluidic bearing decouples non-torque rotor loads to the main shaft driving the hydraulic pumps.

13 HYDRO ENERGY↗

STRESS CORROSION CRACK GROWTH RESPONSE FOR ALLOY 152/52 DISSIMILAR METAL WELDS IN PWR PRIMARY WATER

As part of ongoing research into primary water stress corrosion cracking (PWSCC) susceptibility of alloy 690 and its welds, SCC tests have been conducted on alloy 152/52 dissimilar metal (DM) welds with cracks positioned with the goal to assess weld dilution and fusion line effects on SCC susceptibility. No increased crack growth rate was found when evaluating a 20% Cr dilution zone in alloy 152M joined to carbon steel (CS) that had not undergone a post-weld heat treatment (PWHT). However, high SCC crack growth rates were observed when the crack reached the fusion line of that material where it propagated both on the fusion line and in the heat affected zone (HAZ) of the carbon steel. Crack surface and crack profile examinations of the specimen revealed that cracking in the weld region was transgranular (TG) with weld grain boundaries not aligned with the geometric crack growth plane of the specimen. The application of a typical pressure vessel PWHT on a second set of alloy 152/52 – carbon steel DM weld specimens was found to eliminate the high SCC susceptibility in the fusion line and carbon steel HAZ regions. PWSCC tests were also performed on alloy 152-304SS DM weld specimens. Constant K crack growth rates did not exceed 5x10-9 mm/s in this material with post-test examinations revealing cracking primarily on the fusion line and slightly into the 304SS HAZ.

alloy 152↗

Transient behavior of a scaled RCCS test facility under postulated fault and accident scenarios

Tests were performed on the Natural convection Shutdown heat removal Test Facility (NSTF) to simulate design basis accident and postulated fault scenarios. Residing at Argonne National Laboratory, the NSTF stands nearly 26-m in total height and reflects a ½ scale reactor cavity cooling system (RCCS) for high temperature gas cooled reactors. The following manuscript details three test conditions performed on the experimental test facility. The first simulated the reactor pressure vessel (RPV) boundary condition during depressurized conduction cool down accident with small primary leak, and was repeated during both winter and summer seasons. The second examined a short-circuit break between the inlet and outlet flow paths, and was performed in three incremental stages of nominal flow area break size. The third and final test case studied system behavior with varying amounts of cooling channel blockages, up to and including 50% flow areas. Nominal component temperatures, heat removal performance, and system stability will be presented to characterize the behavior at these conditions.

decay heat removal↗

A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoid overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety system components such as the safety relief valve (SRV), the RCIC system, the wet well, and the dry well. The results show reasonable system behaviors while exhibiting rich dynamics such as variable flow rates through RCIC turbine and pump during the SBO transient. The model has the potential to resolve the Fukushima RCIC mystery after adding the off-design two-phase turbine operation model and other additional improvements.

RCIC↗

USE OF SMALL SPECIMENS FOR FRACTURE TOUGHNESS EVALUATION OF RPV STEELS

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of standard Charpy specimen or its broken half may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs and most likely to be used in advanced reactors as per ASME code.

Sokolov, Mikhail A↗

Passive filtration of air egressing from nuclear containment

A nuclear reactor includes a reactor core comprising fissile material disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor. A containment compartment contains the radiological containment. A heat sink includes a chimney configured to develop an upward-flowing draft in response to heated fluid flowing into a lower portion of the chimney. A fluid conduit is arranged to receive fluid from the containment compartment and to discharge into the chimney. A filter may be provided, with the fluid conduit including a first fluid conduit arranged to receive fluid from the containment compartment and to discharge into an inlet of the filter, and a second fluid conduit arranged to receive fluid from an outlet of the filter and to discharge into the chimney. As the draft is developed passively, there is no need for a blower or pump configured to move fluid through the fluid conduit.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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.↗

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.↗

Bandwidth Study on Energy Use and Potential Energy Savings Opportunities in U.S. Titanium Manufacturing

Energy bandwidth studies can serve as foundational references in framing the range (or bandwidth) of potential energy savings opportunities in U.S. manufacturing. This bandwidth study is one of a series of six energy bandwidth studies on lightweight structural materials manufacturing. The study examines energy consumption and potential energy savings opportunities in U.S. titanium manufacturing. The study relies on multiple sources to estimate the energy used in the manufacturing of the lightweight material for four application areas (automotive, wind energy, aerospace, and pressure vessels), accounting for approximately 75% of overall U.S. titanium production in 2010. Energy savings opportunities for the three manufacturing subareas studied are based on technologies currently in use or under development.

energy bandwidth study, titanium, advanced manufac↗

Bandwidth Study on Energy Use and Potential Energy Saving Opportunities in U.S. Carbon Fiber Reinforced Polymer Manufacturing

Energy bandwidth studies can serve as foundational references in framing the range (or bandwidth) of potential energy savings opportunities in U.S. manufacturing. This bandwidth study is one of a series of six energy bandwidth studies on lightweight structural materials manufacturing. The study examines energy consumption and potential energy savings opportunities in U.S. carbon fiber reinforced polymer composites manufacturing. The study relies on multiple sources to estimate the energy used in the manufacturing of the lightweight material for four application areas (automotive, wind energy, aerospace, and pressure vessels), accounting for approximately 51% of overall U.S. carbon fiber production in 2010. Energy savings opportunities for the six manufacturing sub-processes studied are based on technologies currently in use or under development.

energy energy bandwidth study, carbon fiber, reinf↗

Bandwidth Study on Energy Use and Potential Energy Saving Opportunities in U.S. Magnesium Manufacturing

Energy bandwidth studies can serve as foundational references in framing the range (or bandwidth) of potential energy savings opportunities in U.S. manufacturing. This bandwidth study is one of a series of six energy bandwidth studies on lightweight structural materials manufacturing. The study examines energy consumption and potential energy savings opportunities in U.S. magnesium manufacturing. The study relies on multiple sources to estimate the energy used in the manufacturing of the lightweight material for four application areas (automotive, wind energy, aerospace, and pressure vessels), accounting for approximately 32% of overall U.S. magnesium production in 2010. Energy savings opportunities for the four manufacturing subareas studied are based on technologies currently in use or under development.

energy energy bandwidth study, magnesium, advanced↗

Bandwidth Study on Energy Use and Potential Energy Saving Opportunities in U.S. Glass Fiber Reinforced Polymer Manufacturing

Energy bandwidth studies can serve as foundational references in framing the range (or bandwidth) of potential energy savings opportunities in U.S. manufacturing. This bandwidth study is one of a series of six energy bandwidth studies on lightweight structural materials manufacturing. The study examines energy consumption and potential energy savings opportunities in U.S. glass fiber reinforced polymer composites manufacturing. The study relies on multiple sources to estimate the energy used in the manufacturing of the lightweight material for four application areas (automotive, wind energy, aerospace, and pressure vessels), accounting for approximately 47% of overall U.S. glass fiber production in 2010. Energy savings opportunities for the six manufacturing sub-processes studied are based on technologies currently in use or under development.

energy energy bandwidth study, glass fiber, reinfo↗

Natural gas intestine packed storage tank

A high-pressure pressure vessel for storing natural gas comprises a plurality of first vessel regions of first diameters, a plurality of couplers, and a fiber layer. A three dimensional volume is filled using at least in part the plurality of first vessel regions. Each coupler of the plurality of couplers couples each pair of first vessel regions of the plurality of first vessel regions. Each coupler of the plurality of couplers comprises a second vessel region of a second diameter and two third vessel regions that transition diameters between the first diameter and the second diameter. The three dimensional volume is filled using at least in part the plurality of couplers. The first vessel regions and the couplers comprise a material with low permeability to natural gas. The fiber layer surrounds the plurality of first vessel regions and the plurality of couplers.

Griffith, Saul↗

A study of fission product migration paths in irradiated TRISO particle SiC

Tristructural isotropic (TRISO) coated fuel particles used in high-temperature gas-cooled nuclear reactors (HTGRs) provide key advantages in terms of passive safety and reactor energy efficiencies. TRISO fuel consists of a number of special coating layers surrounding the fissile material in the kernel as shown in Fig. 1. The coating layers consist of a carbon buffer layer, inner pyrolytic carbon (PyC) layer, a silicon carbide (SiC) layer as well as an outer pyrolytic carbon layer. Together the layers act as a pressure vessel and barrier to the release of radioactive fission products into the reactor environment. The cubic 3C-SiC layer of thickness approximately 35 µm, acts as the primary barrier to the migration of metallic fission products out of the particle. The finding, more than three decades years ago, that silver can be released by intact TRISO particles has led to significant research to determine possible transport mechanisms of certain fission products in SiC. Out-of-reactor experiments did not reveal any significant lattice diffusion of silver in SiC1. Many earlier findings alternatively proposed a dominating role of fast diffusion paths such as grain boundaries and dislocations in enhancing fission product migration in SiC. Recently the role of the fission product palladium in enhancing the transport rate of silver in SiC2 was confirmed by transmission electron microscopy (TEM) of neutron-irradiated SiC from a TRISO coated particle from Compact 4-1-1, which was irradiated to an average burnup of 19.38% fissions per initial metal atom (FIMA); a time-averaged, volume-averaged temperature of 1072 °C; a time-averaged peak temperature of 1182 °C; and an average fast fluence of 4.13 x 1021 n/cm2 as part of the US Department of Energy's Advanced Gas Reactor (AGR)-1 experiment3. Although other studies found that Ag was found separately (without Pd) in TRISO coated particles irradiated at 11.4 %, 16.7 and 18.6% FIMA, Pd could have been present at lower than measureable concentrations4. In this study, SiC focussed ion beam (FIB) sections obtained from Compact 4-1-1 from the AGR-1 experiment were analysed in a double Cs-corrected JEOL ARM 200F using atomically resolved scanning transmission electron microscopy (STEM). By using atom resolved STEM in combination with electron energy loss spectroscopy (EELS) and large angle energy dispersive spectroscopy (EDS) a much better understanding of the migration routes and mechanisms of fission products in irradiated polycrystalline 3C-SiC was achieved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Graphite Licensing (ASME) and Testing (ASTM) Technical Assistance

DOE ART Graphite R&D program is currently involved in developing the structural code for licensing the new High Temperature Reactor (HTR) designs through the ASME Boiler Pressure Vessel Code (BPVC). The basic structural design code for graphite core components has been written, approved, and within the BPVC since 2012. However, several areas within the graphite code remain to be improved, specifically the addition of material property changes resulting from environmental degradation (neutron flux and oxidation). The incorporation of this data, which can significantly affect the graphite components and alter the structural integrity of the core during operation, is proving to be an important activity. How this data will be incorporated within the code and how it will be used for license applications will be critical to a number of HTR applicants in the future. These modifications to the code must take into account the unique irradiation and oxidation response of the wide spectrum of available nuclear graphite grades currently under consideration for HTR applications. Additionally, the ASME BPVC requires material property testing within the expected operational conditions of the HTR core including neutron radiation flux, elevated temperatures, and after chronic/acute oxidation of the graphite components. Currently, no standardized graphite testing methods are approved for these environmental conditions. New American Society for Testing and Materials (ASTM) standardized tests for elevated temperature testing, small sample testing, and testing after oxidation must be developed to support the requirements within the ASME code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗