Many membrane distillation models have been created to simulate the heat and mass exchange process involved but most of the literature only validates models to a couple of cases with minor configuration changes. Tools are needed that allow tradeoffs between many configurations. The multiconfiguration membrane distillation model handles many configurations. This report introduces membrane distillation, provides theory, and presents the work to verify and validate the model against experimental data from Colorado School of Mines and a lower resolution model created at the National Renewable Energy Laboratory. Though more data analysis and testing are needed, an initial look at the model to experimental comparisons indicates that the model correlates to the data well but that design comparisons are likely to be incorrect across a broad range of configurations. More accurate quantification of heat and mass transfer through computational fluid mechanics is suggested. The model is open source software: https ://github.com/dlvilla/MCMD1
Membrane distillation is a water purification technology which uses a porous hydrophobic membrane. Liquid water cannot penetrate the membrane at operational pressures but vapor flows through the membrane if there is a vapor pressure difference across the membrane. Many configurations for membrane distillation have been investigated over the last several decades. In this modeling effort, two successful direct contact membrane model development using steady-state control volume balances on energy and mass are presented. Verification and validation of the models is applied to the extent necessary to use the models for comparative design purposes. Significant errors between modeling and experimental membrane distillation data are argued to be due to uncertainty in membrane material property measurements. A second effort to model a vacuum membrane distillation system designed by Memsys(r) is still progressing. Two efforts have not yet produced output mass flow comparable to the literature. Even so, much of the framework needed to model the Memsys(r) system is complete. Membrane Distillation Modeling Progress Report Fiscal Year 2016 February 7, 2017 4 REVISION HISTORY Document Number/Revision Date Description SAND2017-0200 November 2016 Official Use Only - Third Party Proprietary SAND2017-1448 February 6, 2017 Approved for Unlimited Release.
This report summarizes an investigation into the technical feasibility and economic viability of use grain wastes from the beer brewing process as fuel to generate the heat needed in subsequent brewing process. The study finds that while use of spent grain as a biofuel is technically feasible, the economics are not attractive. Economic viability is limited by the underuse of capital equipment. The investment in heating equipment requires a higher utilization that the client brewer currently anticipates. It may be possible in the future that changing factors may swing the decision to a more positive one.
Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine decreases the developed turbine torque; the RCIC speed then slows, and thus the pump flow rate to the RPV decreases. Subsequently, RPV water level decreases due to continued boiling and the liquid fraction flowing to the RCIC decreases, thereby accelerating the RCIC and refilling the RPV. The feedback cycle then repeats itself and/or reaches a quasi-steady equilibrium condition. In other words, the water carry-over is limited by cyclic RCIC performance degradation, and hence the system becomes self-regulating. The indications achieved to date with the system model are more qualitative than quantitative. The avenues being pursued to increase the fidelity of the model are expected to add quantitative realism. The end product will be generic in the sense that the RCIC model will be incorporable within the larger reactor coolant system model of any nuclear power plant or experimental configuration.
The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34D accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.
In the summer of 2020, the National Aeronautics and Space Administration (NASA) plans to launch a spacecraft as part of the Mars 2020 mission. One option for the rover on the proposed spacecraft uses a Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) to provide continuous electrical and thermal power for the mission. NASA has prepared an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act. The EIS includes information on the risks of mission accidents to the general public and on-site workers at the launch complex. The Nuclear Risk Assessment (NRA) addresses the responses of the MMRTG option to potential accident and abort conditions during the launch opportunity for the Mars 2020 mission and the associated consequences. This information provides the technical basis for the radiological risks of the MMRTG option for the EIS. This paper provides a summary of the methods and results used in the NRA.
In the summer of 2020, the National Aeronautics and Space Administration (NASA) plans to launch a spacecraft as part of the Mars 2020 mission. One option for the rover on the proposed spacecraft uses a Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) to provide continuous electrical and thermal power for the mission. An alternative option being considered is a set of solar panels for electrical power with up to 80 Light-Weight Radioisotope Heater Units (LWRHUs) for local component heating. Both the MMRTG and the LWRHUs use radioactive plutonium dioxide. NASA is preparing an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act. The EIS will include information on the risks of mission accidents to the general public and on-site workers at the launch complex. This Nuclear Risk Assessment (NRA) addresses the responses of the MMRTG or LWRHU options to potential accident and abort conditions during the launch opportunity for the Mars 2020 mission and the associated consequences. This information provides the technical basis for the radiological risks of both options for the EIS.
The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.
The potential to treat non-traditional water sources using power plant waste heat in conjunction with membrane distillation is assessed. Researchers and power plant designers continue to search for ways to use that waste heat from Rankine cycle power plants to recover water thereby reducing water net water consumption. Unfortunately, waste heat from a power plant is of poor quality. Membrane distillation (MD) systems may be a technology that can use the low temperature waste heat (<100 F) to treat water. By their nature, they operate at low temperature and usually low pressure. This study investigates the use of MD to recover water from typical power plants. It looks at recovery from three heat producing locations (boiler blow down, steam diverted from bleed streams, and the cooling water system) within a power plant, providing process sketches, heat and material balances and equipment sizing for recovery schemes using MD for each of these locations. It also provides insight into life cycle cost tradeoffs between power production and incremental capital costs.
Most of the regulatory agencies world-wide require that containers used for the transportation of natural UF6 and depleted UF6 must survive a fully-engulfing fire environment for 30 minutes as described in 10CFR71 and in TS-R-1. The primary objective of this project is to examine the thermo-mechanical performance of 48Y transportation cylinders when exposed to the regulatory hypothetical fire environment without the thermal protection that is currently used for shipments in those countries where required. Several studies have been performed in which UF6 cylinders have been analyzed to determine if the thermal protection currently used on UF6 cylinders of type 48Y is necessary for transport. However, none of them could clearly confirm neither the survival nor the failure of the 48Y cylinder when exposed to the regulatory fire environment without the additional thermal protection. A consortium of five companies that move UF6 is interested in determining if 48Y cylinders can be shipped without the thermal protection that is currently used. Sandia National Laboratories has outlined a comprehensive testing and analysis project to determine if these shipping cylinders are capable of withstanding the regulatory thermal environment without additional thermal protection. Sandia-developed coupled physics codes will be used for the analyses that are planned. A series of destructive and non-destructive tests will be performed to acquire the necessary material and behavior information to benchmark the models and to answer the question about the ability of these containers to survive the fire environment. Both the testing and the analysis phases of this project will consider the state of UF6 under thermal and pressure loads as well as the weakening of the steel container due to the thermal load. Experiments with UF6 are also planned to collect temperature- and pressure-dependent thermophysical properties of this material.
The Plutonium Air Transportable Package, Model PAT-1, is certified under Title 10, Code of Federal Regulations Part 71 by the U.S. Nuclear Regulatory Commission (NRC) per Certificate of Compliance (CoC) USA/0361B(U)F-96 (currently Revision 9). The purpose of this SAR Addendum is to incorporate plutonium (Pu) metal as a new payload for the PAT-1 package. The Pu metal is packed in an inner container (designated the T-Ampoule) that replaces the PC-1 inner container. The documentation and results from analysis contained in this addendum demonstrate that the replacement of the PC-1 and associated packaging material with the T-Ampoule and associated packaging with the addition of the plutonium metal content are not significant with respect to the design, operating characteristics, or safe performance of the containment system and prevention of criticality when the package is subjected to the tests specified in 10 CFR 71.71, 71.73 and 71.74.
The Plutonium Air Transportable Package, Model PAT-1, is certified under Title 10, Code of Federal Regulations Part 71 by the U.S. Nuclear Regulatory Commission (NRC) per Certificate of Compliance (CoC) USA/0361B(U)F-96 (currently Revision 9). The National Nuclear Security Administration (NNSA) submitted SAND Report SAND2009-5822 to NRC that documented the incorporation of plutonium (Pu) metal as a new payload for the PAT-1 package. NRC responded with a Request for Additional Information (RAI), identifying information needed in connection with its review of the application. The purpose of this SAND report is to provide the authors responses to each RAI. SAND Report SAND2010-6106 containing the proposed changes to the Addendum is provided separately.
This document provides general guidance for the design and analysis of bolted joint connections. An overview of the current methods used to analyze bolted joint connections is given. Several methods for the design and analysis of bolted joint connections are presented. Guidance is provided for general bolted joint design, computation of preload uncertainty and preload loss, and the calculation of the bolted joint factor of safety. Axial loads, shear loads, thermal loads, and thread tear out are used in factor of safety calculations. Additionally, limited guidance is provided for fatigue considerations. An overview of an associated Mathcad{copyright} Worksheet containing all bolted joint design formulae presented is also provided.
Due to increasing concerns over the buildup of long-lived transuranic isotopes in spent nuclear fuel waste, attention has been given in recent years to technologies that can burn up these species. The separation and transmutation of transuranics is part of a solution to decreasing the volume and heat load of nuclear waste significantly to increase the repository capacity. A fusion neutron source can be used for transmutation as an alternative to fast reactor systems. Sandia National Laboratories is investigating the use of a Z-Pinch fusion driver for this application. This report summarizes the initial design and engineering issues of this ''In-Zinerator'' concept. Relatively modest fusion requirements on the order of 20 MW can be used to drive a sub-critical, actinide-bearing, fluid blanket. The fluid fuel eliminates the need for expensive fuel fabrication and allows for continuous refueling and removal of fission products. This reactor has the capability of burning up 1,280 kg of actinides per year while at the same time producing 3,000 MWth. The report discusses the baseline design, engineering issues, modeling results, safety issues, and fuel cycle impact.
This report summarizes the work conducted for the Z-inertial fusion energy (Z-IFE) late start Laboratory Directed Research Project. A major area of focus was on creating a roadmap to a z-pinch driven fusion power plant. The roadmap ties ZIFE into the Global Nuclear Energy Partnership (GNEP) initiative through the use of high energy fusion neutrons to burn the actinides of spent fuel waste. Transmutation presents a near term use for Z-IFE technology and will aid in paving the path to fusion energy. The work this year continued to develop the science and engineering needed to support the Z-IFE roadmap. This included plant system and driver cost estimates, recyclable transmission line studies, flibe characterization, reaction chamber design, and shock mitigation techniques.