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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 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
NASA NTRS1992

Water Sensors

Mike Morris, former Associate Director of STAC, formed pHish Doctor, Inc. to develop and sell a pH monitor for home aquariums. The monitor, or pHish Doctor, consists of a sensor strip and color chart that continually measures pH levels in an aquarium. This is important because when the level gets too high, ammonia excreted by fish is highly toxic; at low pH, bacteria that normally break down waste products stop functioning. Sales have run into the tens of thousands of dollars. A NASA Tech Brief Technical Support Package later led to a salt water version of the system and a DoE Small Business Innovation Research (SBIR) grant for development of a sensor for sea buoys. The company, now known as Ocean Optics, Inc., is currently studying the effects of carbon dioxide buildup as well as exploring other commercial applications for the fiber optic sensor.

Source record↗
DOE OSTI2018

Packing Optimization for Dismantled Parts of Nuclear Facilities - 18625

When dismantling nuclear facilities, the removed large components must be segmented into smaller pieces and then be stored in the containers. With good cutting and packing plans, the required quantity of the containers can be decreased, and the cost spent in preparing, handling, and disposing these containers can also be reduced significantly. To accomplish this, this article developed a methodology which integrates voxelization and metaheuristic to discretize the irregular shape of various components and put them into the containers with greatest efficiency. The proposed methodology is not limited to certain component types or shapes and can be used for all types of containers. This makes the proposed methodology useful and practical in the nuclear facility dismantling applications. (authors)

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DOE OSTI2014

Full-Scale Accident Testing in Support of Used Nuclear Fuel Transportation.

The safe transport of spent nuclear fuel and high-level radioactive waste is an important aspect of the waste management system of the United States. The Nuclear Regulatory Commission (NRC) currently certifies spent nuclear fuel rail cask designs based primarily on numerical modeling of hypothetical accident conditions augmented with some small scale testing. However, NRC initiated a Package Performance Study (PPS) in 2001 to examine the response of full-scale rail casks in extreme transportation accidents. The objectives of PPS were to demonstrate the safety of transportation casks and to provide high-fidelity data for validating the modeling. Although work on the PPS eventually stopped, the Blue Ribbon Commission on America’s Nuclear Future recommended in 2012 that the test plans be re-examined. This recommendation was in recognition of substantial public feedback calling for a full-scale severe accident test of a rail cask to verify evaluations by NRC, which find that risk from the transport of spent fuel in certified casks is extremely low. This report, which serves as the re-assessment, provides a summary of the history of the PPS planning, identifies the objectives and technical issues that drove the scope of the PPS, and presents a possible path for moving forward in planning to conduct a full-scale cask test. Because full-scale testing is expensive, the value of such testing on public perceptions and public acceptance is important. Consequently, the path forward starts with a public perception component followed by two additional components: accident simulation and first responder training. The proposed path forward presents a series of study options with several points where the package performance study could be redirected if warranted.

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DOE OSTI2015

Impact of different standard type A7A drum closure-ring practices on gasket contraction and bolt closure distance - 15621

The Department of Energy, the Savannah River National Laboratory, several manufacturers of specification drums, and the United States Department of Transportation (DOT) are collaborating in the development of a guidance document for DOE contractors and vendors who wish to qualify containers to DOT 7A Type A requirements. Currently, the effort is focused on DOT 7A Type A 208-liter (55-gallons) drums with a standard 12-gauge bolted closure ring. The U.S. requirements, contained in Title 49, Part 178.350 'Specification 7A; general packaging, Type A specifies a competent authority review of the packaging is not required for the transport of (Class 7) radioactive material containing less than Type A quantities of radioactive material. For Type AF drums, a 4 ft. regulatory free drop must be performed, such that the drum 'suffers maximum damage.' Although the actual orientation is not defined by the specification, recent studies suggest that maximum damage would result from a shallow angle top impact, where kinetic energy is transferred to the lid, ultimately causing heavy damage to the lid, or even worse, causing the lid to come off. Since each vendor develops closure recommendations/procedures for the drums they manufacture, key parameters applied to drums during closing vary based on vendor. As part of the initial phase of the collaboration, the impact of the closure variants on the ability of the drum to suffer maximum damage is investigated. Specifically, closure testing is performed varying: 1) the amount of torque applied to the closure ring bolt; and, 2) stress relief protocol, including: a) weight of hammer; and, b) orientation that the hammer hits the closure ring. After closure, the amount of drum lid gasket contraction and the distance that the closure bolt moves through the closure ring is measured. (authors)

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DOE OSTI2024

M3SF-24LL010301062-Surface Complexation/Ion Exchange Data Integration for Radionuclide Sorption to Clay Minerals

This progress report (Level 3 Milestone Number M3SF-24LL010301062) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite International Collaborations Work Package SF-24LL01030106. The activity is focused on our long-term commitment to engaging our partners in international nuclear waste repository research. The focus of this milestone is the establishment of international collaborations for sorption modeling and the associated impacts of unlocking larger, community-based datasets. More specifically, we are developing a database framework for Spent Fuel and Waste and Science Technology (SFWST) that is aligned with the Helmholtz Zentrum Dresden Rossendorf (HZDR) and other international sorption database development groups in support of the database needs of the SFWST program.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Incredibility of a Postulated Criticality Event for Enriched Uranyl Nitrate Storage

At the H-Canyon outside facilities, highly enriched uranyl nitrate solutions collected from leaks or during deliberate maintenance activities is collected into 5-gallon polyethylene bottles. The current criticality safety evaluation for the facility postulates that a criticality event is possible if the solutions are allowed to concentrate through evaporation, freezing, storing multiple bottles in a geometrically unfavorable array, or any combination of these conditions. Using chemical data on uranyl nitrate freezing from Savannah River National Laboratory, protected limits on concentration and enrichment, and normal operational restrictions, criticality was shown to be incredible under all normal and credible abnormal conditions. A bounding normal solution is analyzed under normal conditions and shown to be substantially subcritical. A bounding credible abnormal solution is defined as 9.2 g U{sup 235}/L and 73 wt. % enriched and 0 M excess nitric acid. The concentration and enrichment are limitations already protected by the facility for other processes. Because acidity is not specifically restricted, the most reactive lower bound is selected. The credible abnormal worst placement is in a concrete basin reflected on three sides. There the bottle(s) would be subject in the worst cases to flooding (full water reflection) and postulated concentration through evaporation or freezing. Computational modeling was performed using KENOVI in the SCALE 6.1 code package. (authors)

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DOE OSTI2019

Maximum Inner Container Dose Rate That Will Ensure Meeting Regulatory Limit at Package Surface - 19206

Often, the shipper of radioactive material does not know the exact radioactive isotopic composition of the materials that need to be shipped. This is especially important for legacy materials that need to be removed from a site that is in the process of being decommissioned. A method of determining the acceptability of such material for shipment before it is placed in into the shipping package will allow more flexibility in planning for decommissioning activities. A series of Monte-Carlo calculations were performed to calculate the dose rate at the surface of a convenience container or a shielding pig separately and then at the surface of a new shipping package being designed by the Savannah River National Laboratory (SRNL). The purpose of these calculations is to determine the maximum dose rate at the surface of the convenience container or the shielded pig that will ensure the dose rate at the outer surface the shipping package is less that the regulatory limit of 2 mSv/h. The calculations also considered the effect of shape of the radioactive material and its possible movement inside the convenience container along with movement of the convenience container inside the confinement vessel of the shipping package. (authors)

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DOE OSTI2020

Dismantling of the Bohunice V1 Reactor Coolant System - 20011

About 2.5 kilometers from the village of Jaslovske Bohunice in the Trnava District in western Slovakia at the Bohunice V1 nuclear power plant, Westinghouse and VUJE are working on the dismantling of two VVER-440 units. The consortium was awarded this contract in October 2017 from Jadrova a vyradovacia spolocnost (JAVYS) and has begun onsite work to dismantle the reactor coolant systems of both units that operated at the plant. The Bohunice V1 plant is composed of two V-230 type VVER 440 units (440 MWe each). The first unit ceased operation on December 31, 2006, and the second on December 31, 2008, after 28 years of operation. The project strategy relies on the early release of some plant areas to allow the erection of new facilities to enable a fast and safe dismantling. These up-front activities play a crucial role within the whole project. Since schedule and ALARA are two key elements to be considered, it has been decided to separate the dry cutting activities from the ones occurring underwater. Two Wet Cutting Workshops will allow the segmentation and packaging of the high activity components. To take advantage of the large space available in the turbine building, a Dry Cutting Workshop has been built for segmenting the twelve steam generators of both units. Two more Dry Cutting Workshops (one per unit) have been installed in the reactor building for the fragmentation of the remaining large components and auxiliary systems components. This paper describes all current preparatory works as well as the on-going segmentation and packaging activities. Interesting lessons learned are also provided. (authors)

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DOE OSTI2013

Simulation of cooling and solidification of three-dimensional bulk borosilicate glass. Effect of structural relaxations

The modeling of the viscoelastic stress evolution and specific volume relaxation of a bulky glass cast is presented in this article and is applied to the experimental cooling process of an inactive nuclear waste vitrification process. The concerned borosilicate glass is solidified and cooled down to ambient temperature in a stainless steel canister, and the thermomechanical response of the package is simulated. There exists a deviant compression of the liquid core due to the large glass package compared to standard tempered glass plates. The stress load development of the glass cast is finally studied for different thermal load scenarios, where the cooling process parameters or the final cooldown rates were changed, and we found a great influence of the studied cooldown rates on the maximum stress buildup at ambient temperature.

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DOE OSTI2019

Preclosure Risk Assessment for Deep Borehole Disposal

This report presents a preclosure radiological safety assessment for deep borehole disposal (DBD) of nuclear wastes. The primary purpose of the safety assessment is to identify risk factors for disposal operations, to aid in design for an engineering demonstration of technology for DBD. The assessment is based on a conceptual design for disposal packages and borehole systems that was developed previously. It considers operational steps that could be used for actual DBD, with internal and external initiating off-normal events, to develop insights that can be applied to an engineering demonstration that would be performed without using any form of nuclear waste. This research was performed as part of the deep borehole field test (DBFT). Based on revised U.S. Department of Energy (DOE) priorities in mid-2017, the DBFT and other research related to a DBD option was discontinued; ongoing work and documentation were closed out by the end of fiscal year (FY) 2017. This report was initiated as part of the DBFT and documented as an incomplete draft at the end of FY 2017. The report was finalized by Sandia National Laboratories in FY2018 without DOE funding, subsequent to the termination of the DBFT, and published in FY2019. iii

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NASA NTRS2012

Hands-Free Transcranial Color Doppler Probe

Current transcranial color Doppler (TCD) transducer probes are bulky and difficult to move in tiny increments to search and optimize TCD signals. This invention provides miniature motions of a TCD transducer probe to optimize TCD signals. The mechanical probe uses a spherical bearing in guiding and locating the tilting crystal face. The lateral motion of the crystal face as it tilts across the full range of motion was achieved by minimizing the distance between the pivot location and the crystal face. The smallest commonly available metal spherical bearing was used with an outer diameter of 12 mm, a 3-mm tall retaining ring, and 5-mm overall height. Small geared motors were used that would provide sufficient power in a very compact package. After confirming the validity of the basic positioning concept, optimization design loops were completed to yield the final design. A parallel motor configuration was used to minimize the amount of space wasted inside the probe case while minimizing the overall case dimensions. The distance from the front edge of the crystal to the edge of the case was also minimized to allow positioning of the probe very close to the ear on the temporal lobe. The mechanical probe is able to achieve a +/-20deg tip and tilt with smooth repeatable action in a very compact package. The enclosed probe is about 7 cm long, 4 cm wide, and 1.8 cm tall. The device is compact, hands-free, and can be adjusted via an innovative touchscreen. Positioning of the probe to the head is performed via conventional transducer gels and pillows. This device is amendable to having advanced software, which could intelligently focus and optimize the TCD signal.

Chin, Robert↗
DOE OSTI2021

Uncertainty and Sensitivity Analysis Methods and Applications in the GDSA Framework (FY2021)

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (FCT) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). Two high priorities for SFWST disposal R&D are design concept development and disposal system modeling. These priorities are directly addressed in the SFWST Geologic Disposal Safety Assessment (GDSA) control account, which is charged with developing a geologic repository system modeling and analysis capability, and the associated software, GDSA Framework, for evaluating disposal system performance for nuclear waste in geologic media. GDSA Framework is supported by SFWST Campaign and its predecessor the Used Fuel Disposition (UFD) campaign. This report fulfills the GDSA Uncertainty and Sensitivity Analysis Methods work package (SF-21SN01030404) level 3 milestone, Uncertainty and Sensitivity Analysis Methods and Applications in GDSA Framework (FY2021) (M3SF-21SN010304042). It presents high level objectives and strategy for development of uncertainty and sensitivity analysis tools, demonstrates uncertainty quantification (UQ) and sensitivity analysis (SA) tools in GDSA Framework in FY21, and describes additional UQ/SA tools whose future implementation would enhance the UQ/SA capability of GDSA Framework. This work was closely coordinated with the other Sandia National Laboratory GDSA work packages: the GDSA Framework Development work package (SF-21SN01030405), the GDSA Repository Systems Analysis work package (SF-21SN01030406), and the GDSA PFLOTRAN Development work package (SF-21SN01030407). This report builds on developments reported in previous GDSA Framework milestones, particularly M3SF 20SN010304032.

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DOE OSTI2016

Contingency Base Energy Management System

CB-EMS is the latest implementation of DSOM (Decision Support for Operations and Maintenance), which was previously patented by PNNL. CB-EMS WAS specifically designed for contingency bases for the US Army. It is a software package that is designed to monitor energy consumption at an Army contingency base to alert the camp manager when the systems are wasting energy. It's main feature that separates it from DSOM is it's ability to add systems using a plug and play menu system.

Source record↗
DOE OSTI2018

Impact of Different Vegetative Cover Scenarios on Infiltration Rates for the E-Area PA Intact Case

This memorandum addresses the effects of vegetative cover type (bamboo versus Bahia grass) and the timing of pine tree intrusion on intact infiltration rate for the proposed E-Area Low-Level Waste Facility (LLWF) closure cap system. It builds upon earlier reports whose purpose is to lay the foundation for the infiltration data package that will be assembled during the next revision of the E-Area LLWF Performance Assessment (PA).

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS1972

Heat pipe applications for the space shuttle.

Discussion of six specific applications for heat pipe (HP) devices on the space shuttle. These applications were chosen from 27 concepts formulated as part of a study to evaluate the potential benefits associated with HP use. The formulation process is briefly described along with the applications which evolved. The bulk of the discussion deals with the 'top' six - namely, HP radiators for waste heat rejection, an HP augmented cold rail, an HP circuit for electronic equipment cooling, modular heat sink for control of remote packages, an HP temperature control for compartments, and air-cooled equipment racks. The philosophy, physical design details, and performance data are presented for each concept along with a comparison with the baseline design where applicable.

Tawil, M.↗
DOE OSTI2018

Application of an AREVA Mobile Evaporator Design Used Successfully at NRC Licensed U.S. Commercial Nuclear Sites to Support Hanford Site Operations - 18134

The U.S. Department of Energy (DOE), Office of River Protection (ORP), manages the River Protection Project at the Hanford site in the South Eastern corner of Washington State in the Pacific Northwest of the United States. The mission of the River Protection Project is to retrieve and treat Hanford's tank waste and close the tank farms to protect the Columbia River. The Site Tank Operations Contractor (Washington River Protection Solutions (WRPS)) via its site operations teams creates additional space in the Double Shell Tanks (DSTs) using the site's only remaining operational evaporator - 242-A. The current 242-A Evaporator has long been identified as a potential single point of failure that in the event of a major breakdown, could significantly impact retrieval operations at the Hanford Site. Areva has experience both on the Hanford site as an integrated subcontractor to the mission and in the design, procurement and construction of a mobile forced circulation evaporator unit designed to be used on Nuclear Regulatory Commission commercial nuclear utility sites. The mobile unit has been used at the following nuclear power plants: SONGS, Salem, Davis Besse, Waterford and Palo Verde. It was designed to take spent Water-EDTA based chemical cleaning solvent and evaporate water off to reduce the volume of waste required to go through downstream processes. It was further designed to fit inside ISO shipping containers. A multi-step approach will be used to: 1. review lessons learned from the previous DOE study for deploying a mobile evaporator on the Hanford sit, 2. evaluate the proposed waste streams and the required throughput and waste arising, sizing the evaporator appropriately, 3. develop a concept design requirements document for the use of a customized evaporator to meet the Hanford need 4. Design, and build a mobile evaporator and 5. Test, install and commission at a site to be agreed with WRPS/DOE. With an overall budget for the approach described of around $35 m, an operational evaporator could be delivered to site in 18-24 months from the approval of a detailed design package. Potential befits include the elimination of the 242-A evaporator as a single point failure, the creation of additional DST tank space and the elimination of the need to stage waste at the 242-A evaporator by moving the mobile unit to a location where is could be deployed at or close to the tank such as in 200W. (authors)

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DOE OSTI2011

TRUPACT-III Operations Manual (Rev. 0)

This procedure provides instructions for assembling the following contact-handled (CH) packaging payload: Standard Large Box 2 (SLB2).

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