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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 289 records · Page 16

NMSBA Leveraged Project Interim Status Report

We have investigated the quality of copper particles available on the market today and determined their complete unsuitability to be used for copper inks, since all of them were significantly oxidized: there was only 50% of metallic copper in each batch and the rest was Cu 2 O and CuO. To date, we have fully identified the challenges and developed the synthesis for large amounts of the copper ink precursor, namely, copper(I) mesityl. Currently, the amounts of tens of grams of the precursor have been obtained. From this precursor, four small batches of copper nanoparticles (50 mg each) have been synthesized to investigate the possibility of decreasing particle sensitivity to oxygen. These particles have been treated with different surface-stabilizing agents (namely, octylamine, oleylamine, pyridine, benzotriazole, and dodecanethiol) in order to investigate their influence on particle oxygen sensitivity. A batch of copper nanoparticles in the amount of 2 grams has also been synthesized in order to start preparation of test ink batches with different solvents, surfactants, and stabilizers.

36 MATERIALS SCIENCE↗

FY17 Status Report on the Computing Systems for the Yucca Mountain Project TSPA-LA Models.

Sandia National Laboratories (SNL) continued evaluation of total system performance assessment (TSPA) computing systems for the previously considered Yucca Mountain Project (YMP). This was done to maintain the operational readiness of the computing infrastructure (computer hardware and software) and knowledge capability for total system performance assessment (TSPA) type analysis, as directed by the National Nuclear Security Administration (NNSA), DOE 2010. This work is a continuation of the ongoing readiness evaluation reported in Lee and Hadgu (2014), Hadgu et al. (2015) and Hadgu and Appel (2016). The TSPA computing hardware (CL2014) and storage system described in Hadgu et al. (2015) were used for the current analysis. One floating license of GoldSim with Versions 9.60.300, 10.5, 11.1 and 12.0 was installed on the cluster head node, and its distributed processing capability was mapped on the cluster processors. Other supporting software were tested and installed to support the TSPA- type analysis on the server cluster. The current tasks included preliminary upgrade of the TSPA-LA from Version 9.60.300 to the latest version 12.0 and address DLL-related issues observed in the FY16 work. The model upgrade task successfully converted the Nominal Modeling case to GoldSim Versions 11.1/12. Conversions of the rest of the TSPA models were also attempted but program and operational difficulties precluded this. Upgrade of the remaining of the modeling cases and distributed processing tasks is expected to continue. The 2014 server cluster and supporting software systems are fully operational to support TSPA-LA type analysis.

97 MATHEMATICS AND COMPUTING↗

Analysis of simultaneous thermal/gamma radiation aging of cross-linked polyethylene (XLPE) insulation—interim status report

Cross-linked polyethylene (XLPE) is the most common cable insulation found in nuclear containment, and is therefore a priority material for investigation of long term aging effect from elevated temperature combined with gamma radiation exposure. Prior work has identified the possibility of anomalous aging behavior in XLPE such as the inverse temperature effect in which radiation exposure is more damaging at lower temperatures than at higher temperatures. We explored simultaneous aging of XLPE insulation from modern Firewall® III nuclear cables at 60, 90, and 115 °C, at gamma dose rates from 116 to 540 Gy/h, for exposure periods up to 25 d. XLPE samples exposed in this way were characterized using the percent gel and uptake factor method. For the conditions and material examine, degradation behavior was seen to track proportionally with increasing temperature, rather than to exhibit greater degradation at lower temperatures. Ongoing work including similar aging at 25 °C and characterization of the XLPE samples using other methods will further elucidate these initial results

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

FY17 Status Report on the Micromechanical Finite Element Modeling of Creep Fracture of Grade 91 Steel

Advanced reactors designed to operate at higher temperatures than current light water reactors require structural materials with high creep strength and creep-fatigue resistance to achieve long design lives. Grade 91 is a ferritic/martensitic steel designed for long creep life at elevated temperatures. It has been selected as a candidate material for sodium fast reactor intermediate heat exchangers and other advanced reactor structural components. This report focuses on the creep deformation and rupture life of Grade 91 steel. The time required to complete an experiment limits the availability of long-life creep data for Grade 91 and other structural materials. Design methods often extrapolate the available shorter-term experimental data to longer design lives. However, extrapolation methods tacitly assume the underlying material mechanisms causing creep for long-life/low-stress conditions are the same as the mechanisms controlling creep in the short-life/high-stress experiments. A change in mechanism for long-term creep could cause design methods based on extrapolation to be non-conservative. The goal for physically-based microstructural models is to accurately predict material response in experimentally-inaccessible regions of design space. An accurate physically-based model for creep represents all the material mechanisms that contribute to creep deformation and damage and predicts the relative influence of each mechanism, which changes with loading conditions. Ideally, the individual mechanism models adhere to the material physics and not an empirical calibration to experimental data and so the model remains predictive for a wider range of loading conditions. This report describes such a physically-based microstructural model for Grade 91 at 600° C. The model explicitly represents competing dislocation and diffusional mechanisms in both the grain bulk and grain boundaries. The model accurately recovers the available experimental creep curves at higher stresses and the limited experimental data at lower stresses, predominately primary creep rates. The current model considers only one temperature. However, because the model parameters are, for the most part, directly related to the physics of fundamental material processes, the temperature dependence of the properties are known. Therefore, temperature dependence can be included in the model with limited additional effort. The model predicts a mechanism shift for 600° C at approximately 100 MPa from a dislocation- dominated regime at higher stress to a diffusion-dominated regime at lower stress. This mechanism shift impacts the creep life, notch-sensitivity, and, likely, creep ductility of Grade 91. In particular, the model predicts existing extrapolation methods for creep life may be non-conservative when attempting to extrapolate data for higher stress creep tests to low stress, long-life conditions. Furthermore, the model predicts a transition from notchstrengthening behavior at high stress to notch-weakening behavior at lower stresses. Both behaviors may affect the conservatism of existing design methods.

36 MATERIALS SCIENCE↗

Status Report and Research Plan for Cables Harvested from Crystal River Unit 3 Nuclear Generating Plant

Harvested cables from operating or decommissioned nuclear power plants present an important opportunity to validate models, understanding material aging behavior, and validate characterization techniques. Crystal River Unit 3 Nuclear Generating Plant is a pressurized water reactor that was licensed to operate from 1976 to 2013. Cable segments were harvested and made available to the Light Water Reactor Sustainability research program through the Electric Power Research Institute. Information on the locations and circuits within the reactor from whence the cable segments came, cable construction, sourcing and installation information, and photographs of the cable locations prior to harvesting were provided. The cable variations provided represent six of the ten most common cable insulations in the nuclear industry and experienced service usage for periods from 15 to 42 years. Subsequently, these cables constitute a valuable asset for research to understand aging behavior and measurement of nuclear cables. Received cables harvested from Crystal River Unit 3 Nuclear Generating Plant consist of low voltage, insulated conductor surrounded by jackets in lengths from 24 to 100 feet each. Cable materials will primarily be used to investigate aging under simultaneous thermal and gamma radiation exposure. Each cable insulation and jacket material will be characterized in its as-received condition, including determination of the temperatures associated with endothermic transitions in the material using differential scanning calorimetry and dynamic mechanical analysis. Temperatures for additional thermal exposure aging will be selected following the thermal analysis to avoid transitions in accelerated laboratory aging that do not occur in field conditions. Aging temperatures above thermal transitions may also be targeted to investigate the potential for artifacts in lifetime prediction from rapid accelerated aging. Total gamma doses and dose rates targeted for each material will be determined based on filling gaps in prior work, known limits of material classes and resource constraints. Experimental plans will be developed in the context of existing data for the insulation and jacket materials available in published Department of Energy and Electric Power Research Institute reports toward addressing identified knowledge gaps.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Complete Status Report Documenting Development of Friction Stir Welding for Joining Thin Wall Tubing of ODS Alloys

The development of friction stir welding (FSW) for joining thin sections of the advanced oxide dispersion strengthened (ODS) 14YWT ferritic alloy was initiated in Fuel Cycle Research and Development (FCRD), now the Nuclear Technology Research and Development (NTRD), in 2015. The first FSW experiment was conducted in late FY15 and successfully produced a bead-on-plate stir zone (SZ) on a 1 mm thick plate of 14YWT (SM13 heat). The goal of this research task is to ultimately demonstrate that FSW is a feasible method for joining thin wall (0.5 mm thick) tubing of 14YWT.

36 MATERIALS SCIENCE↗

Status Report on the High-Temperature Steam Electrolysis Plant Model Developed in the Modelica Framework (FY17)

This report has been prepared as part of an effort to design and build a modeling and simulation (M&S) framework to assess the economic viability of a nuclear-renewable hybrid energy system (N-R HES). In order to facilitate dynamic M&S of such an integrated system, research groups in multiple national laboratories have been developing various subsystems as dynamic physics-based components using the Modelica programming language. In fiscal year 2015 (FY15), Idaho National Laboratory (INL) performed a dynamic analysis of two region-specific N-R HES configurations, including the gas-to-liquid (natural gas to Fischer-Tropsch synthetic fuel) and brackish water reverse osmosis desalination plants as industrial processes. In FY16, INL developed two additional subsystems in the Modelica framework: (1) a high-temperature steam electrolysis (HTSE) plant as a high priority industrial plant to be integrated with a light water reactor (LWR) within an N-R HES and (2) a gas turbine power plant as a secondary energy supply. In FY17, five new components (i.e., a feedwater pump, a multi-stage compression system, a sweep-gas turbine, flow control valves, and pressure control valves) have been incorporated into the HTSE system proposed in FY16, aiming to better realistically characterize all key components of concern. Special attention has been given to the controller settings based on process models (i.e., direct synthesis method), aiming to improve process dynamics and controllability. A dynamic performance analysis of the improved LWR/HTSE integration case was carried out to evaluate the technical feasibility (load-following capability) and safety of such a system operating under highly variable conditions requiring flexible output. The analysis (evaluated in terms of the step response) clearly shows that the FY17 model resulted in superior output responses with much smaller settling times and less oscillatory behavior in response to disturbances in the electric load than those observed with the FY16 model. Simulation results involving several case studies show that the suggested control scheme could maintain the controlled variables (including the steam utilization factor, cathode stream inlet composition, and temperatures and pressures of the process streams at various locations) within desired limits under various plant operating conditions. The results also indicate that the proposed HTSE plant could provide operational flexibility to participate in energy management at the utility scale by dynamically optimizing the use of excess plant capacity within an N-R HES.

08 HYDROGEN↗

FY17 Status Report on NEAMS Neutronics Activities

Under the U.S. DOE NEAMS program, the high-fidelity neutronics code system has been developed to support the multiphysics modeling and simulation capability named SHARP. The neutronics code system includes the high-fidelity neutronics code PROTEUS, the cross section library and preprocessing tools, the multigroup cross section generation code MC2-3, the in-house meshing generation tool, the perturbation and sensitivity analysis code PERSENT, and post-processing tools. The main objectives of the NEAMS neutronics activities in FY17 are to continue development of an advanced nodal solver in PROTEUS for use in nuclear reactor design and analysis projects, implement a simplified sub-channel based thermal-hydraulic (T/H) capability into PROTEUS to efficiently compute the thermal feedback, improve the performance of PROTEUS-MOCEX using numerical acceleration and code optimization, improve the cross section generation tools including MC2-3, and continue to perform verification and validation tests for PROTEUS.

47 OTHER INSTRUMENTATION↗

Evaluation of Iodine Remediation Technologies in Subsurface Sediments: Interim Status Report

Isotopes of iodine were generated during plutonium production from nine production reactors at the U.S. Department of Energy Hanford Site. The long half-life 129I generated at the Hanford Site during reactor operations was 1) stored in single-shell and double-shell tanks, 2) discharged to liquid disposal sites (e.g., cribs and trenches), 3) released to the atmosphere during fuel reprocessing operations, or 4) captured by off-gas absorbent devices (silver reactors) at chemical separations plants (PUREX, B-Plant, T-Plant, and REDOX). Releases of 129I to the subsurface have resulted in several large, though dilute, plumes in the groundwater, including the plume in the 200-UP-1 operable unit. There is also 129I remaining in the vadose zone beneath disposal or leak locations. Because 129I is an uncommon contaminant, relevant remediation experience and scientific literature are limited.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Status Report - Fiberboard Properties and Degradation Rates for Storage of the 9975 Shipping Package in KAC

Thermal, mechanical and physical properties have been measured on fiberboard samples following accelerated aging for up to approximately 12 years. The aging environments include elevated temperature up to 250 ºF (the maximum allowed service temperature for fiberboard in 9975 packages) and elevated humidity. The results from this testing have been analyzed, and aging models fit to the data. Correlations relating several properties (thermal conductivity, energy absorption, weight, dimensions and density) to their rate of change in potential storage environments have been developed. Combined with an estimate of the actual conditions the fiberboard experiences in KAC, these models allow development of service life predictions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Discharge Permit-1845 Quarterly Status Report April 2018

Discharge Permit (DP)-1845 was issued by the New Mexico Environment (NMED) Ground Water Quality Bureau (GWQB) for discharges via up to three injection wells in a phased Treatability Study of in-situ bioremediation of groundwater at Sandia National Laboratories, New Mexico, Technical Area-V Groundwater Area of Concern. This report fulfills the quarterly reporting requirements set forth in DP-1845, Section IV.B, Monitoring and Reporting. This reporting period is October 1st through December 31st, 2017. The report is due to NMED GWQB by May 1st, 2018.

54 ENVIRONMENTAL SCIENCES↗

Annual Status Report (FY2017): Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds

This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-EP-06451). These estimates are compared with performance objectives defined in U.S. Department of Energy (DOE) requirements (DOE O 435.1 Chg 1 2 and its companion documents DOE M 435.1-1 Chg 1 3 and DOE G 435.1-1 4 ). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063 5 ) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations, which indicate potential impacts well below performance objectives. Therefore, future compliance with DOE O 435.1 Chg 1 is expected.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Annual Status Report (FY2017): Performance Assessment for the Disposal of Low-Level Waste in the 200 East Area Burial Grounds

This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. The estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-SD-WM-TI-7301). The estimates are compared with performance objectives defined in U.S. Department of Energy (DOE) requirements (DOE O 435.1 Chg 1, 2 and companion documents DOE M 435.1-1 Chg 1 3 and DOE G 435.1-1 4 ). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-00635) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations, which indicate potential impacts well below performance objectives. Therefore, future compliance with DOE O 435.1 Chg 1 is expected.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Annual Status Report (FY2017): Performance Assessment for the Environmental Restoration Disposal Facility

In accordance with U.S. Department of Energy (DOE) requirements in DOE O 435.1 Chg 11 and DOE M 435.1-1 Chg 1,2 a determination of continued adequacy of the performance assessment (PA), composite analysis (CA), and disposal authorization statement (DAS) is required on an annual basis. This determination must consider the results of data collection and analysis from research, field studies, and monitoring.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Discharge Permit-1845 Quarterly Status Report July 2018

Trichloroethene (TCE) and nitrate have been identified as constituents of concern in groundwater at the Sandia National Laboratories, New Mexico (SNL/NM) Technical Area (TA)-V Groundwater (TAVG) Area of Concern (AOC) based on detections above the U.S. Environmental Protection Agency (EPA) maximum contaminant level (MCL) in samples collected from monitoring wells. The EPA MCLs and the State of New Mexico drinking water standards for TCE and nitrate are 5 micrograms per liter and 10 milligrams per liter (as nitrogen), respectively.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

FY2018 Status Report: Model 9975 O-Ring Fixture Long-Term Leak Performance

A series of experiments to monitor the aging performance of Viton® GLT and GLT-S O-rings used in the Model 9975 shipping package has been ongoing since 2004 at the Savannah River National Laboratory. One approach has been to periodically evaluate the leak performance of O-rings being aged in mock-up 9975 Primary Containment Vessels (PCVs) at elevated temperature. Other methods such as compression stress-relaxation (CSR) tests and field surveillance are also on going to evaluate O-ring behavior. Seventy tests using PCV mock-ups with GLT O-rings were assembled and heated to temperatures ranging from 200 to 450 ºF. They were leak-tested initially and have been tested periodically to determine if they continue to meet the leak-tightness criterion defined in ANSI standard N14.5-97. Due to material substitution, a smaller test matrix with fourteen additional tests was initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 ºF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗