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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Parasitic erbium photoluminescence in commercial telecom fiber optical components

Noiseless optical components are critical for applications ranging from metrology to quantum communication. Here, we characterize several commercial telecom C-band fiber components for parasitic noise using a tunable laser. We observe the spectral signature of trace concentrations of erbium in all devices from the underlying optical crystals including YVO 4 , LiNbO 3 , TeO 2 , and amorphous material transmitting infrared radiation glass. Furthermore, due to the long erbium lifetime, these signals are challenging to mitigate at the single photon level in the telecom range and suggests the need for higher purity optical crystals.

36 MATERIALS SCIENCE↗
DOE OSTI2018

RitR is an archetype for a novel family of redox sensors in the streptococci that has evolved from two-component response regulators and is required for pneumococcal colonization

To survive diverse host environments, the human pathogen Streptococcus pneumoniae must prevent its self-produced, extremely high levels of peroxide from reacting with intracellular iron. However, the regulatory mechanism(s) by which the pneumococcus accomplishes this balance remains largely enigmatic, as this pathogen and other related streptococci lack all known redox-sensing transcription factors. Here we describe a two-component-derived response regulator, RitR, as the archetype for a novel family of redox sensors in a subset of streptococcal species. We show that RitR works to both repress iron transport and enable nasopharyngeal colonization through a mechanism that exploits a single cysteine (Cys128) redox switch located within its linker domain. Biochemical experiments and phylogenetics reveal that RitR has diverged from the canonical two-component virulence regulator CovR to instead dimerize and bind DNA only upon Cys128 oxidation in air-rich environments. Atomic structures show that Cys128 oxidation initiates a “helical unravelling” of the RitR linker region, suggesting a mechanism by which the DNA-binding domain is then released to interact with its cognate regulatory DNA. Expanded computational studies indicate this mechanism could be shared by many microbial species outside the streptococcus genus.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2017

The component formula of Salvia miltiorrhiza and Panax ginseng induces apoptosis and inhibits cell invasion and migration through targeting PTEN in lung cancer cells

Lung cancer still remains the leading cause of cancer-related death worldwide. It is an urgent need for development of novel therapeutic agents to improve current treatment of this disease. Here we investigate whether the effective component formula of traditional Chinese Medicine could serve as new potential therapeutic drugs to treat lung cancer. We optimize the most effective component formula of Salvia miltiorrhiza and Panax Ginseng (FMG), which is composed of Salvianolic acid A, 20(S)-Ginsenoside and Ginseng polysaccharide. We discovered that FMG selectively inhibited lung cancer cell proliferation and induced apoptosis but had no any cytotoxic effects on normal lung epithelial BEAS-2B cells. Moreover, FMG inhibited lung cancer cell migration and invasion. Mechanistically, we found that FMG significantly promoted p-PTEN expression and subsequently inhibited PI3K/AKT signaling pathway. The phosphatase activity of PTEN protein was increased after FMG bound to PTEN protein, indicating that PTEN is one of the FMG targeted proteins. In addition, FMG regulated expression of some marker proteins relevant to cell apoptosis, migration and invasion. Collectively, these results provide mechanistic insight into the anti-NSCLC of FMG by enhancing the phosphatase activity of PTEN, and suggest that FMG could be as a potential option for lung cancer treatment.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2015

Rapid development of an ice sheet climate application using the components-based approach

In the third year of a five-year jointly funded project, Sandia’s Albany/FELIX simulation code for ice sheet dynamics is being developed to be run in two modes: (1) as a stand-alone model for scientific investigations, and (2) as part of the land-ice component of coupled climate simulations in the Department of Energy’s (DOE’s) earth system model. The land-ice component simulates changes to the Greenland and Antarctic ice sheets, including their contributions to global sea-level rise. Using high performance computing (HPC), Sandia has recently completed a controlled mesh convergence study to compute the rate at which the Greenland ice sheet is flowing. The study demonstrated that the HPC code is capable of working accurately, efficiently, and reliably on larger scales (over 1 billion unknowns). The study also identified, for the first time, the minimum number of vertical levels needed in the mesh to maintain accuracy

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Thermal and Structural Analysis of Beamline Components in the Mu2e Experiment

Fermi National Accelerator Laboratory will be conducting the high energy particle physics experiment Muons to Electrons (Mu2e). In this experiment, physicists will attempt to witness and understand an ultra-rare process which is the conversion of a muon into the lighter mass electron, without creating additional neutrinos. The experiment is conducted by first generating a proton beam which will be collided into a target within the production solenoid (PS). This creates a high-intensity muon beam which passes through a transport solenoid (TS) and into the detector solenoid (DS). In the detector solenoid the muons will be stopped in an aluminum target and a series of detectors will measure the electrons produced. These components have been named the DS train since they are coupled and travel on a rail system when being inserted or extracted from the DS. To facilitate the installation and removal of the DS train, a set of external stands and a support stand for the instrumentation feed-through bulkhead (IFB) have been designed. Full analysis of safety factors and performance of these two designs has been completed. The detector solenoid itself will need to be maintained to a temperature of 22°C ± 10°C. This will minimize thermal strain and ensure the accurate position of the components is maintained to the tolerance of 2 mm. To reduce the thermal gradient, a passive heating system has been developed and reported.

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Enhanced Component Performance Study: Air-Operated Valves 1998-2014

This report presents a performance evaluation of air-operated valves (AOVs) at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The AOV failure modes considered are failure-to-open/close, failure to operate or control, and spurious operation. The component reliability estimates and the reliability data are trended for the most recent 10-year period, while yearly estimates for reliability are provided for the entire active period. One statistically significant trend was observed in the AOV data: The frequency of demands per reactor year for valves recording the fail-to-open or fail-to-close failure modes, for high-demand valves (those with greater than twenty demands per year), was found to be decreasing. The decrease was about three percent over the ten year period trended.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Enhanced Component Performance Study: Emergency Diesel Generators 1998–2014

This report presents an enhanced performance evaluation of emergency diesel generators (EDGs) at U.S. commercial nuclear power plants. This report evaluates component performance over time using (1) Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES) data from 1998 through 2014 and (2) maintenance unavailability (UA) performance data from Mitigating Systems Performance Index (MSPI) Basis Document data from 2002 through 2014. The objective is to show estimates of current failure probabilities and rates related to EDGs, trend these data on an annual basis, determine if the current data are consistent with the probability distributions currently recommended for use in NRC probabilistic risk assessments, show how the reliability data differ for different EDG manufacturers and for EDGs with different ratings; and summarize the subcomponents, causes, detection methods, and recovery associated with each EDG failure mode. Engineering analyses were performed with respect to time period and failure mode without regard to the actual number of EDGs at each plant. The factors analyzed are: sub-component, failure cause, detection method, recovery, manufacturer, and EDG rating. Six trends with varying degrees of statistical significance were identified in the data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Enhanced Component Performance Study: Motor-Driven Pumps 1998–2014

This report presents an enhanced performance evaluation of motor-driven pumps at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The motor-driven pump failure modes considered for standby systems are failure to start, failure to run less than or equal to one hour, and failure to run more than one hour; for normally running systems, the failure modes considered are failure to start and failure to run. An eight hour unreliability estimate is also calculated and trended. The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. Statistically significant increasing trends were identified in pump run hours per reactor year. Statistically significant decreasing trends were identified for standby systems industry-wide frequency of start demands, and run hours per reactor year for runs of less than or equal to one hour.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Enhanced Component Performance Study: Turbine-Driven Pumps 1998–2014

This report presents an enhanced performance evaluation of turbine-driven pumps (TDPs) at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The TDP failure modes considered are failure to start (FTS), failure to run less than or equal to one hour (FTR=1H), failure to run more than one hour (FTR>1H), and normally running systems FTS and failure to run (FTR). The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. Statistically significant increasing trends were identified for TDP unavailability, for frequency of start demands for standby TDPs, and for run hours in the first hour after start. Statistically significant decreasing trends were identified for start demands for normally running TDPs, and for run hours per reactor critical year for normally running TDPs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x10 25 n/m 2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x10 25 n/m 2 .

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Air Dispersion Modeling for the INL Application for a Synthetic Minor Sitewide Air Quality Permit to Construct with a Facility Emission Cap Component

The Department of Energy Idaho Operations Office (DOE-ID) is applying for a synthetic minor, Sitewide, air quality permit to construct (PTC) with a facility emission cap (FEC) component from the Idaho Department of Environmental Quality (DEQ) for Idaho National Laboratory (INL) to limit its potential to emit to less than major facility limits for criteria air pollutants (CAPs) and hazardous air pollutants (HAPs) regulated under the Clean Air Act. This document is supplied as an appendix to the application, Idaho National Laboratory Application for a Synthetic Minor Sitewide Air Quality Permit to Construct with a Facility Emissions Cap Component, hereafter referred to as “permit application” (DOE-ID 2015). Air dispersion modeling was performed as part of the permit application process to demonstrate pollutant emissions from the INL will not cause a violation of any ambient air quality standards. This report documents the modeling methodology and results for the air dispersion impact analysis. All CAPs regulated under Section 109 of the Clean Air Act were modeled with the exception of lead (Pb) and ozone, which are not required to be modeled by DEQ. Modeling was not performed for toxic air pollutants (TAPs) as uncontrolled emissions did not exceed screening emission levels for carcinogenic and non-carcinogenic TAPs. Modeling for CAPs was performed with the EPA approved AERMOD dispersion modeling system (Version 14134) (EPA 2004a) and five years (2000-2004) of meteorological data. The meteorological data set was produced with the companion AERMET model (Version 14134) (EPA 2004b) using surface data from the Idaho Falls airport, and upper-air data from Boise International Airport supplied by DEQ. Onsite meteorological data from the Grid 3 Mesonet tower located near the center of the INL (north of INTEC) and supplied by the local National Oceanic and Atmospheric Administration (NOAA) office was used for surface wind directions and wind speeds. Surface data (i.e., land use data that defines roughness, albedo, Bowen ratio, and other parameters) were processed using the AERSURFACE utility (Version 13016) (EPA 2013). Emission sources were modeled as point sources using actual stack locations and dimensions. Emissions, flow rates and exit temperatures were based on the design operating capacity of each source. All structures close enough to produce an area of wake effect were included for all sources. For multi-tiered structures, the heights of the tiers were included or the entire building height was assumed to be equal to the height of the tallest tier. Concentrations were calculated at 1,352 receptor locations provided by DEQ. All receptors were considered for each pollutant and averaging period. Maximum modeled CAP concentrations summed with average background concentration values were presented and compared to National Ambient Air Quality Standards (NAAQS). The background concentration values used were obtained using the Washington State University’s Laboratory for Atmospheric Research North West Airquest web-based retrieval tool (http://lar.wsu.edu/nw airquest/lookup.html). The air dispersion modeling results show the maximum impacts for CAPs are less than applicable standards and demonstrate the INL will not cause a violation of any ambient air quality standards.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x1025 n/m2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x1025 n/m2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

9975 Shipping package component long-term degradation rates

Special nuclear materials are being stored in the K-Area Complex using 3013 containers that are held within Model 9975 shipping packages. The service life for these packages in storage was recently increased from 15 to 20 years, since some of these packages have been stored for nearly 15 years. A strategy is also being developed whereby such storage might be extended beyond 20 years. This strategy is based on recent calculations that support acceptable 9975 package performance for 20 years with internal heat loads up to 19 watts, and identifies a lower heat load limit for which the package components should degrade at half the bounding rate or less, thus doubling the effective storage life for these lower wattage packages. The components of the 9975 package that are sensitive to aging under storage conditions are the fiberboard overpack and the O-ring seals, although some degradation of the lead shield and outer drum are also possible. This report summarizes degradation rates applicable to lower heat load storage conditions. In particular, the O-ring seals should provide leak-tight performance for more than 40 years in packages for which their maximum temperature is ≤135 °F. Similarly, the fiberboard should remain acceptable in performance of its required safety functions for up to 40 years in packages with a maximum fiberboard temperature ≤125 °F.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Additively Manufactured IN718 Components with Wirelessly Powered and Interrogated Embedded Sensing

A methodology is described for embedding commercial-off-the-shelf sensors together with wireless communication and power circuit elements using direct laser metal sintered additively manufactured components. Physics based models of the additive manufacturing processes and sensor/wireless level performance models guided the design and embedment processes. A combination of cold spray deposition and laser engineered net shaping was used to fashion the transmitter/receiving elements and embed the sensors, thereby providing environmental protection and component robustness/survivability for harsh conditions. By design, this complement of analog and digital sensors were wirelessly powered and interrogated using a health and utilization monitoring system; enabling real-time, in situ prognostics and diagnostics.

embedded sensing↗
DOE OSTI2018

Thermal Analysis of Fermilab Mu2e Beamstop and Structural Analysis of Beamline Components

The Mu2e project at Fermilab National Accelerator Laboratory aims to observe the unique conversion of muons to electrons. The success or failure of the experiment to observe this conversion will further the understanding of the standard model of physics. Using the particle accelerator, protons will be accelerated and sent to the Mu2e experiment, which will separate the muons from the beam. The muons will then be observed to determine their momentum and the particle interactions occur. At the end of the Detector Solenoid, the internal components will need to absorb the remaining particles of the experiment using polymer absorbers. Because the internal structure of the beamline is in a vacuum, the heat transfer mechanisms that can disperse the energy generated by the particle absorption is limited to conduction and radiation. To determine the extent that the absorbers will heat up over one year of operation, a transient thermal finite element analysis has been performed on the Muon Beam Stop. The levels of energy absorption were adjusted to determine the thermal limit for the current design. Structural finite element analysis has also been performed to determine the safety factors of the Axial Coupler, which connect and move segments of the beamline. The safety factor of the trunnion of the Instrument Feed Through Bulk Head has also been determined for when it is supporting the Muon Beam Stop. The results of the analysis further refine the design of the beamline components prior to testing, fabrication, and installation.

Narug, Colin S.↗
DOE OSTI2018

Embedded Sensors and Controls to Improve Component Performance and Reliability - Final Report

Oak Ridge National Laboratory has researched embedded instrumentation and controls (I&C) for extreme environments under the Department of Energy, Office of Nuclear Energy's Nuclear Energy Enabling Technology program. Advanced sensors and controls have the potential to dramatically increase the performance and efficiency of nuclear reactor components, however, the extreme environments of exiting and advanced reactor designs limit the options for sensors, actuators, and control designs. To address two of the extreme environment challenges to embedded I&C, the focus of this research is to fill technology gaps needed for electromagnetic systems that can operate at temperatures greater than 700 C in corrosive environments. A magnetically suspended, canned-rotor pump was chosen for the embedded I&C development platform because it is a system with a practical need in industry and because the sensors, actuators, and embedded I&C technologies needed to successfully build this platform will be widely applicable to any high-temperature components that utilize electromagnetic interactions. This design also eliminates the need for pump seals and mechanical bearings which are the main sources of pump failures and maintenance costs. During the project a conceptual design for the high-temperature magnetically suspended pump was developed and analyzed for manufacturability and reliability. Specifications for the materials, control system, sensors, mechanical performance, high-temperature wiring, and electromagnetic properties were developed for the conceptual design. The canned rotor design of the high-temperature pump presents unique fluid force characteristics arising from the thin fluid channel between the rotor and stator. A model of the fluid forces was developed and the impact of these forces on the stability and performance of magnetic bearing controllers was studied. Then a room temperature bench-scale magnetic testbed was designed and fabricated for testing the performance of sensors and electromagnetic actuators, modern control design techniques, closed-loop system identification, state estimation, and model uncertainty. Finally, a loop-scale testbed with submerged magnetic bearings was designed and fabricated. This testbed will be used to analyze the stability and performance of control system designs in response to the nonlinear and cross-coupling fluid effects between the shaft axes of motion, rotordynamics, gyroscopic effects, and impeller disturbances.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Report on the completion of long-lead components procurement of new sodium materials test loop

This report provides an update on the completion of long-lead components procurement of new sodium materials test loop. The loop, designated as SMT-3, is a forced-convection sodium loop to be used for the evaluation of sodium compatibility of advanced Alloy 709 with emphasis on long term exposures of tensile, creep, fatigue, creep fatigue, and fracture toughness ASTM-size specimens in support of ASME Code qualification and NRC licensing. The report is a deliverable (Level 4) in FY18 (M4NT-18AN050503013), under the Work Package NT-18AN05050301, “A709 Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of the Advanced Materials Program for the Advanced Reactor Technology. This work package enables the development of advanced structural materials by providing corrosion, microstructure, and mechanical property data from the standpoint of sodium compatibility of advanced structural alloys. The first sodium loop (SMT-1) with a single tank was constructed in 2011 at ANL and has been in operation for exposure of subsize sheet specimens of advanced alloys at a single temperature. The second sodium loop with dual tanks (SMT-2) was constructed in 2013 and has been in operation for the exposure of subsize sheet specimens of advanced alloys at two different temperatures. The current loop (SMT-3) has been designed to incorporate sufficient chamber capacity to expose a large number of ASTM-size specimens to evaluate the sodium effects on tensile, creep, fatigue, creep-fatigue, and fracture toughness properties, in support of ASME Code Qualification and USNRC Licensing. Three major long lead items were identified: 1) Selection and approval of location for the new loop and installation of the necessary infrastructure, such as transformer, power supply panels, and wiring, that is adequate for the long term operation of the loop, 2) Perform a detailed piping analysis that meets the ANL QA and Safety guidelines as well as complying with ASME Code requirements, and 3) Design and fabrication of the specimen exposure vessels that are capable of at least 42 tensile, creep, and fatigue specimens in each vessel and is compliant and compatible with ASME Code and the piping design analysis. The SMT-3 loop will be located in Building 206 adjacent to the currently operating SMT-2 loop. In addition, we have removed the aged power supply system in Building 206 and installed a new transformer, wiring, and power panels for the new loop. The preliminary loop design was sent to an outside vendor for a detailed piping analysis, to ensure compliance with the ASME Code guidelines. The piping design was completed in FY2017. The construction of components such as cold trap, economizers, piping arrangement etc. will be performed in the central shops at ANL. About 150 liters of sodium for the loop will be procured in early FY2018. The loop system is designed to circulate sodium through the sample tanks and he associated loop without an operator for an extended period of time. With the three sodium loops (i.e., single-tank, dual-tank and four–tank), materials can be tested at different sodium temperatures, and large tensile, creep, fatigue, creep-fatigue, and fracture toughness specimens can be exposed to sodium for extended periods of time and generate data on mechanical properties in support of ASME Code Qualification and USNRC Licensing of advanced Alloy 709 for use as a structural material in SFRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗