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At least 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
NASA NTRS2008

Production of Methane and Water from Crew Plastic Waste

Recycling is a technology that will be key to creating a self sustaining lunar outpost. The plastics used for food packaging provide a source of material that could be recycled to produce water and methane. The recycling of these plastics will require some additional resources that will affect the initial estimate of starting materials that will have to be transported from earth, mainly oxygen, energy and mass. These requirements will vary depending on the recycling conditions. The degredation products of these plastics will vary under different atmospheric conditions. An estimate of the the production rate of methane and water using typical ISRU processes along with the plastic recycling will be presented.

Captain, Janine↗
DOE OSTI2011

CH-TRU Waste Content Codes (CH-TRUCON) (Rev. 41)

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) contact-handled transuranic (CH-TRU) waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

CH-TRU Waste Content Codes (CH-TRUCON) (Rev. 42)

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) contact-handled transuranic (CH-TRU) waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC.

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DOE OSTI2012

CH-TRU Wasate Content Codes (CH-TRUCON) (Rev. 44)

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) contact-handled transuranic (CH-TRU) waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC.

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DOE OSTI2019

Characterization techniques for TRU waste from the Analytical Laboratory Hot Cells at Idaho National Laboratory

Characterization techniques for TRU waste from the Analytical Laboratory Hot Cells at Idaho National Laboratory Since the first reactor went hot in 1942, there has been a never ending desire to improve the nuclear reactor and the fuel that it burns. The Idaho National Laboratory has been a key player in reactor and fuel development. As more research projects begin, the spectrum of waste parameters grows. Over the years, the Analytical Laboratory (AL) at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL) has supported a large range of various research projects. Each project brings its own challenge and analytical technique that, in turn, yields its own unique waste parameters. The AL Hot Cells at INL consist of 6 small cells that can support around 12 different fuel research projects at a time along with infrastructure and environmental samples. Each different project yields different levels of transuranic (TRU) material, RCRA constituents, dose, and contamination levels. The waste parameters from each project dictate a path forward for the waste Low-Level Waste (LLW), Contact Handled TRU (CH-TRU), Mixed Contact Handled TRU (M-CH-TRU), Remote Handled TRU (RH-TRU), or Mixed Remote Handled TRU (M-RH-TRU). With visual examination techniques and process knowledge as an aid, there have been waste handling methods developed that allow debris and sample material to be segregated. Unique characterization mechanisms have been applied for each debris and sample material that are generally more accurate than smear and surface contamination data. For debris found in the Hot Cells, analytical data is gathered and conservatively summarized for each fuel type and research method. This consists of data from Quadrupole Inductively Coupled Plasma Mass Spectrometry (Q-ICP-MS), Gamma Spectroscopy, and some data verification through Thermal Ionization Mass Spectrometry (TIMS) and Multi-Collector Inductively Coupled Plasma Mass Spectrometry (MC-ICP-MS). This data gives a range of TRU isotope concentrations. Fuel age is taken into account to determine the dominant TRU activity isotopes. Cesium concentrations are also gathered. Each parameter/fuel type then ties a measured Cesium number to TRU content based on historical analytical data. Debris in the hot cells is segregated by project and packaged in individual 2 gallon metal paint cans. Each can is counted by gamma spectroscopy in a calculated and measured geometry. Using process knowledge to determine the fuel or project type, newly counted cesium numbers can be applied to the historic analytical data and TRU content can be estimated. The TRU content allows waste debris to be segregated between CH-TRU and RH-TRU. Packaging techniques can allow for the minimization if RH-TRU waste. Sample material received into the AL must be dissolved into a liquid matrix in order to be analyzed in most cases. When analysis is complete, these liquids go through a consolidation, complexation, neutralization, and solidification process where real time analytical data can be used to calculate TRU content and RCRA constitutes. These calculations allow for no further analytical data to be gathered. The waste matrix is considered stable and acceptable by all AK [define AK] documents. The unique techniques used in characterizing and handling the Analytical Laboratories Hot Cell waste can be applied to various situations and various laboratories. They allow for more accurate data for waste constituents and also for the minimization of RH-TRU waste.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2019

A Monte Carlo Rectangle Packing Algorithm for Identifying Likely Spatial Distributions of Final Closure Cap Subsidence in the E-Area Low-Level Waste Facility

A Monte Carlo rectangle packing algorithm, summarized in Figure ES-1, has been developed to provide insights into the likely spatial distribution of subsided areas or holes in the final closure cap for the E-Area Low-Level Waste Facility (ELLWF) resulting from the placement of non-crushable waste containers within slit and engineered trench disposal units. Using historical records provided by Solid Waste Management as an indicator of future waste streams, a probability distribution of unique waste containers was constructed, and Monte Carlo sampling was used to sequentially select waste containers for placement in a rectangular region bounded by the disposal unit footprint. Within this Monte Carlo approach, the probability of any container type being selected is weighted by the fractional abundance relative to all other container types. For each Monte Carlo realization, the frequency of non-crushable waste containers appearing at each location on a spatially discretized mesh is updated. After many realizations, the location(s) with the highest frequencies of non-crushable waste containers are regarded as the most likely locations to experience subsidence (i.e., for a hole to form). Utilizing this approach has enabled the down-selection of a computationally intractable set of subsidence hole combinations to be explored in three-dimensional finite element groundwater flow and contaminant transport models. Such an approach can be applied to a number of different packaging problems.

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DOE OSTI2019

Strategic Implications of Active In-Container Vitrification Trials on Thermal Treatment of UK LLW and ILW - 19240

Thermal treatment is being considered in the UK for the treatment of a wide range of low and intermediate level wastes (LLW and ILW). To support this objective, a range of in-container vitrification trials were conducted to test a wide range of simulated waste streams. One of the major aims of these trials was to assess the disposability of the thermally treated products and the potential for using this process, either for management of wastes that currently have no clear disposal route, or for reducing the cost of disposal (through volume and hazard reduction). To this end, this paper considers the disposability of the disposal products produced during these treatment trials. Particular focus is given to the waste loading and volume reduction achieved by the process, based on the explicitly calculated values for each of the melts. A brief discussion of other disposability assessment criteria is also provided covering: Homogeneity (are radionuclides evenly distributed through the package): Porosity and permeability of the product; Water content; Potential for gas generation; Organic content; and chemical properties. It was observed that the highest waste loading and greatest volume reduction was achieved when treating wastes that could themselves act as glass formers, such as silicate minerals or soils, removing the need for addition of glass frit. Glass-forming wastes also produced homogenous products with low visible porosity and favourable chemical characteristics. Conversely treatment of wastes containing significant volumes of metal gave relatively poor volume reduction and a heterogenous product with visible porosity. (authors)

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DOE OSTI2018

Separation and processing of plastic films

A successful flexible plastic packaging (FPP) recycling process will require the best of technological development, conscientious industry strategy, an environmentally aware community and likely, governmental support and legislation. FPP is used in almost every industry and can have a wide range of physical and chemical characteristics that lead to the series of challenges encountered at current recycling facilities. FPP is typically a one-time use product, and makes up about 30% of municipal waste plastics (Davis, 2016). Due to its versatility and light weight nature, plastics are the second most commonly used packaging material next to corrugated board (WPO, 2008). Furthermore, FPP and glass are the most detrimental of recyclables to the success of recycling processes. While glass is known to jam conveyors and melt in some separations at recovery facilities, FPP is far more burdensome and challenges the separation processes at recycling facilities. In the U.S. over 250 M tons of solid trash is generated each year, about 32 M tons consisting of plastic (Sandford, 2016; EPA, 2015b). 8.7 M tons of FPP was produced in 2013 in the U.S. (Sandford, 2016) with only 7.5% of FPP being recycled currently (EPA, 2015a). Of the material that is being recycled, 83% originated from commercial sources with 51% of it being clear polyethylene (PE) stretch wrap and poly bags, 20% consisting of mixed color PE stretch wrap films and 12% PE agricultural films (Moore Recycling Associates, 2017). Of the remaining recycled films, 16% originated from consumer sources as PE retail bags, sacks and wraps collected at store and consumer drop off centers. Only 1% of recycled PE films were obtained from curbside recycling programs and processed through material recovery facilities (MRFs). Other types of films such as polyvinyl chloride (PVC) and polypropylene (PP) are essentially not recycled. Only 43% of the recycled FPP is truly recycled back into films/sheets, 44% is down-gauged into plastic composite lumber and the remaining material goes to uses such as marine and agricultural products, crates, buckets and pallets (Moore Recycling associations, 2017).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Status Report on Laboratory Testing and International Collaborations in Salt.

This report is a summary of the international collaboration and laboratory work funded by the US Department of Energy Office of Nuclear Energy Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D work package. This report satisfies milestone levelfour milestone M4SF-17SN010303014. Several stand-alone sections make up this summary report, each completed by the participants. The first two sections discuss international collaborations on geomechanical benchmarking exercises (WEIMOS) and bedded salt investigations (KOSINA), while the last three sections discuss laboratory work conducted on brucite solubility in brine, dissolution of borosilicate glass into brine, and partitioning of fission products into salt phases.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Recycled PET Nanofibers for Water Filtration Applications

Water shortage is an immediate and serious threat to our world population. Inexpensive and scalable methods to clean freshwater and wastewater are in high demand. Nanofiber filtration membranes represent a next generation nonwoven filter media due to their unique properties. Polyethlyene terephthalate (PET) is often used in the packaging of water and other commonly used materials, leading to a large amount of plastic waste often with limited incentive for recycling (few value-added uses). Here, we present work in the generation of nanofiber liquid filtration membranes from PET plastic bottles and demonstrate their use in microfiltration. PET nanofiber membranes were formed via solution electrospinning with fiber diameters as low as ca. 100 nm. Filtration efficiency was tested with latex beads with sizes ranging from 30 to 2000 nm. Greater than 99% of the beads as small as 500 nm were removed using gravity filtration. To reduce biofouling, the mats were functionalized with quaternary ammonium and biguanide biocides. The biguanide functionalized mats achieved 6 log reduction for both gram negative and gram positive bacteria.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Damaged Fuel Rod Service with German Turnkey Technology NCS 45 Transport Casks and Under-Water-Brazing Technology - 18182

Transportation of damaged irradiated fuel rods with the package NCS 45 developed and operated by DAHER Nuclear Technologies GmbH (DNT) saves considerable technical and financial resources in the German phase-out scenario. The technology can be used world-wide to speed up decommissioning of NPPs. In each German reactor (and most likely in all reactors in the world) a small number of damaged fuel rods have accumulated over the operation phase. In general, transport and storage solutions for fuel are concentrated on the bulk of material, i. e. complete fuel assemblies containing undamaged fuel rods. It is hence a rather straightforward process to remove almost all the fuel from the reactor pools. However, damaged fuel rods - which have to be removed from the pool prior to dismantling - can cause considerable delay in dismantling and high costs for safe-guarding and specialized solutions, even though they only contain a comparatively very tiny amount of fuel. DNT has all required tools, equipment and experience to mitigate the described problem: - The package NCS 45 licensed for the transport of damaged Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel rods with a burn-up of up to 120 GWd/MgHM; - A patented underwater brazing (encapsulation) technology to render very high burn-up damaged fuel rods fit for transport; - Qualified personnel to plan, license and operate such encapsulation process and the NCS 45 in a NPP and to carry out the transports; - Proven record of successful removal of damaged fuel rods from several shut-down reactors in Germany. The package NCS 45 is currently licensed in several European countries. Due to its state-of-technique design - type B(U)F-96 approval - it could be licensed and used for world-wide applications. The underwater brazing technology was developed and patented by DNT to address the challenge of transporting high burn-up and/or damaged UOX and MOX fuel rods. Above a burn-up of 55 GWd/MgHM there are neither enough data to assess the behavior of the cladding under stress nor sufficient benchmarks available to validate the source terms of the fuel. Hence, high burn-up fuel rods are encapsulated in the fuel pool by using the underwater-brazing-technology before they are loaded into the package NCS 45. Furthermore, the capacity of the NCS 45 to transport damaged fuel rods can be considerably enhanced by encapsulating these rods. (authors)

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DOE OSTI2024

FY24 Advancements in PFLOTRAN Development for GDSA Framework

The Spent Fuel & Waste Science and Technology (SFWST) Campaign of the Office of Spent Fuel & Waste Disposition of U.S. Department of Energy Office of Nuclear Energy (DOE-NE) is conducting research and development on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). This report describes fiscal year 2024 accomplishments in the Geologic Disposal Safety Assessment (GDSA) PFLOTRAN Development work package, which is charged with developing subsurface simulation software for postclosure performance assessment of deep geologic disposal of SNF and HLW.

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DOE OSTI2018

Developments in the Regulatory Framework of Interim Storage of Spent Nuclear Fuel (SNF) in Germany - 18137

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office for Safety of Nuclear Waste Management (BfE). In 1995, a license for storage of spent nuclear fuel (SNF) and vitrified high level waste (HAW) was granted for the interim storage facility in Gorleben. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the nuclear power plant NPP Lingen. Since this time twelve NPPs erected their own storage facilities additionally to the three central storage facilities (Ahaus, Gorleben and Rubenow). In all of these facilities the concept of dry storage in metallic dual purpose casks (DPCs) is implemented. The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2034. In Germany the decision about a repository site will be taken in the year 2031. To enable the utilities to hand over the SNF and HAW to the repository regime while the current licenses are still valid would require a very fast process licensing and construction of the required facilities at this location. As a delay cannot be excluded, necessary preparations have to be taken on the side of the regulator. In the German Atomic Energy Act precautions have already been taken and consulting the German parliament (Bundestag) is mandatory if inevitable events would require an extension of the currently licensed interim storage period. The new structure with the Federal Office BfE, and two companies for final disposal 'Bundesgesellschaft fuer Endlagerung' (BGE mbH) and interim storage 'Bundesgesellschaft fuer Zwischenlagerung' (BGZmbH) was established in 2017. The Federal Office for Safety of Nuclear Waste Management (BfE) is responsible for regulation in the field of nuclear disposal since 30.07.2016. So the package design approval for DPCs and the storage licenses for SNF will be granted by this office. Furthermore, it is the regulator for the repository and will finally grant a license for the German deep geological repository. Site selection and construction and operation of the repository will be the responsibility of BGE. On January 1, 2018, the BGZ will take over all remaining interim storage facilities operating under a license due to article six of the Atomic Energy Act. (authors)

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DOE OSTI2012

Retrieval System for Calcined Waste for the Idaho Cleanup Project - 12104

This paper describes the conceptual approach to retrieve radioactive calcine waste, hereafter called calcine, from stainless steel storage bins contained within concrete vaults. The retrieval system will allow evacuation of the granular solids (calcine) from the storage bins through the use of stationary vacuum nozzles. The nozzles will use air jets for calcine fluidization and will be able to rotate and direct the fluidization or displacement of the calcine within the bin. Each bin will have a single retrieval system installed prior to operation to prevent worker exposure to the high radiation fields. The addition of an articulated camera arm will allow for operations monitoring and will be equipped with contingency tools to aid in calcine removal. Possible challenges (calcine bridging and rat-holing) associated with calcine retrieval and transport, including potential solutions for bin pressurization, calcine fluidization and waste confinement, are also addressed. The Calcine Disposition Project has the responsibility to retrieve, treat, and package HLW calcine. The calcine retrieval system has been designed to incorporate the functions and technical characteristics as established by the retrieval system functional analysis. By adequately implementing the highest ranking technical characteristics into the design of the retrieval system, the system will be able to satisfy the functional requirements. The retrieval system conceptual design provides the means for removing bulk calcine from the bins of the CSSF vaults. Top-down vacuum retrieval coupled with an articulating camera arm will allow for a robust, contained process capable of evacuating bulk calcine from bins and transporting it to the processing facility. The system is designed to fluidize, vacuum, transport and direct the calcine from its current location to the CSSF roof-top transport lines. An articulating camera arm, deployed through an adjacent access riser, will work in conjunction with the retrieval nozzle to aid in calcine fluidization, remote viewing, clumped calcine breaking and recovery from off-normal conditions. As the design of the retrieval system progresses from conceptual to preliminary, increasing attention will be directed toward detailed design and proof-of- concept testing. (authors)

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DOE OSTI2015

SORTIE DB - Source term integrated environmental management disposal facility database - 15249

The Environmental Management Disposal Facility (EMDF) in Oak Ridge, Tennessee, is a proposed U.S. Department of Energy low-level Comprehensive Environmental Response Compensation and Liability Act (CERCLA) waste disposal facility with an operational start in the early 2020's. As a DOE low-level waste disposal facility, it is subject to DOE Order 435.1, requiring a composite analysis and possibly a performance assessment. The Source Term Integrated EMDF Data Base - SORTIE DB - provides source term estimates used in these assessments. An Oak Ridge Office of Environmental Management (OREM) Federal and Contractor team developed this state-of-the-art descriptive analytics and visualization schema. DOE OREM senior management has recognized complementary beneficial uses of SORTIE DB in decisions for accelerating clean-up progress, dealing with conflicting sequencing and budget priorities, and enhancing stakeholder and regulatory interfaces. SORTIE DB is designated as an OREM planning tool. SORTIE DB contains estimated forecast waste volumes and masses for over 800 Decontamination and Decommissioning (D and D) and Remedial Action (RA) sequenced baseline packages over the next 25+ years. Available characterization data, radionuclide identification and concentrations, and data from safety basis and emergency management documents are incorporated into SORTIE DB. (authors)

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DOE OSTI2015

Field Test to Evaluate Deep Borehole Disposal.

The U.S. Department of Energy (DOE) has embarked on the Deep Borehole Field Test (DBFT), which will investigate whether conditions suitable for disposal of radioactive waste can be found at a depth of up to 5 km in the earth’s crust. As planned, the DBFT will demonstrate drilling and construction of two boreholes, one for initial scientific characterization, and the other at a larger diameter such as could be appropriate for waste disposal (the DBFT will not involve radioactive waste). A wide range of geoscience activities is planned for the Characterization Borehole, and an engineering demonstration of test package emplacement and retrieval is planned for the larger Field Test Borehole. Characterization activities will focus on measurements and samples that are important for evaluating the long-term isolation capability of the Deep Borehole Disposal (DBD) concept. Engineering demonstration activities will focus on providing data to evaluate the concept’s operational safety and practicality. Procurement of a scientifically acceptable DBFT site and a site management contractor is now underway. The concept of deep borehole disposal (DBD) for radioactive wastes is not new. It was considered by the National Academy of Science (NAS 1957) for liquid waste, studied in the 1980’s in the U.S. (Woodward–Clyde 1983), and has been evaluated by European waste disposal R&D programs in the past few decades (for example, Grundfelt and Crawford 2014; Grundfelt 2010). Deep injection of wastewater including hazardous wastes is ongoing in the U.S. and regulated by the Environmental Protection Agency (EPA 2001). The DBFT is being conducted with a view to use the DBD concept for future disposal of smaller-quantity, DOE-managed wastes from nuclear weapons production (i.e., Cs/Sr capsules and granular solid wastes). However, the concept may also have broader applicability for nations that have a need to dispose of limited amounts of spent fuel from nuclear power reactors. For such nations the cost for disposing of volumetrically limited waste streams could be lower than mined geologic repositories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

SRNL's Lessons Learned from the University of Nevada Reno{sup TM} Nuclear Packaging Internships - 19464

The U.S. Department of Energy's Packaging Certification Program (DOE PCP) and the University of Nevada, Reno (UNR) Mechanical Engineering Department have developed an accredited, graduate level nuclear packaging certificate. The mission is to ensure that the DOE staff and users of PCP services are knowledgeable and competent in packaging and transportation activities. Since 2008, the Savannah River National Laboratory (SRNL) has been teaching two classes for the DOE PCP Packaging University. This paper introduces a recently added third class, where SRNL provided Site Supervisors, targeted toward engineering interns. As part of ensuring a future workforce, the Packaging Technology and Transportation Engineering Group (PT and TE) at Savannah River National Laboratory (SRNL) normally employs two to three summer interns each year. The internships provide SRNL innovative s performing valuable research and support work related to package certification. It also provides SRNL the opportunity to bring in perspective engineers on 'trial-basis.' The interns benefit by learning valuable career skills and gaining exposure to a broad-based career field in the field of packaging. As part of the job specific training, the summer interns are provided an overview of package design, testing, and certification. This includes package operations, containment vessel leak testing, and SARP preparation. Two years ago, SRNL recognized the efficiency of using the SRNL-taught courses for the summer interns working in SRNL PT and TE. SRNL requested that UNR allow the PT and TE two summer interns, who were rising college juniors, to enroll in the SRNL-taught courses for graduate credit. UNR agreed and allowed the interns to enroll under a special admission status. Since that time, UNR developed a new class within the Graduate Certificate in Nuclear Packaging (GCNP), for the summer intern research work. This new class offers a formalized framework for the summer interns to conduct research, while ensuring that the research is performed in a rigor consistent with an accredited college level setting. Last summer, the SRNL PT and TE had three summer interns - all participating in the new course, UNR ME-699 Nuclear Packaging Summer Internship - with each earning three graduate credits from UNR in mechanical engineering. This paper provides a brief overview of the DOE PCP Packaging University/ UNR GCNP, a discussion of SRNL's use of summer interns as a tool to recruit and train the future workforce, and lessons learned from having last year's three SRNL PPT and TE summer interns enroll in the UNR Packaging Internship Class. (authors)

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