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At least 1,549 records · Page 86
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Integrated Project Teams - An Essential Element of Project Management during Project Planning and Execution - 12155

Managing complex projects requires a capable, effective project manager to be in place, who is assisted by a team of competent assistants in various relevant disciplines. This team of assistants is known as the Integrated Project Team (IPT). he IPT is composed of a multidisciplinary group of people who are collectively responsible for delivering a defined project outcome and who plan, execute, and implement over the entire life-cycle of a project, which can be a facility being constructed or a system being acquired. An ideal IPT includes empowered representatives from all functional areas involved with a project-such as engineering design, technology, manufacturing, test and evaluation, contracts, legal, logistics, and especially, the customer. Effective IPTs are an essential element of scope, cost, and schedule control for any complex, large construction project, whether funded by DOE or another organization. By recently assessing a number of major, on-going DOE waste management projects, the characteristics of high performing IPTs have been defined as well as the reasons for potential IPT failure. Project managers should use IPTs to plan and execute projects, but the IPTs must be properly constituted and the members capable and empowered. For them to be effective, the project manager must select the right team, and provide them with the training and guidance for them to be effective. IPT members must treat their IPT assignment as a primary duty, not some ancillary function. All team members must have an understanding of the factors associated with successful IPTs, and the reasons that some IPTs fail. Integrated Project Teams should be used by both government and industry. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

A multiobjective modeling approach to locate multi-compartment containers for urban-sorted waste

The location of multi-compartment sorted waste containers for recycling purposes in cities is an important problem in the context of urban waste management. The costs associated with those facilities and the impacts placed on populations are important concerns. This paper introduces a mixed-integer, multiobjective programming approach to identify the locations and capacities of such facilities. The approach incorporates an optimization model in a Geographical Information System (GIS)-based interactive decision support system that includes four objectives. The first objective minimizes the total investment cost; the second one minimizes the average distance from dwellings to the respective multi-compartment container; the last two objectives address the 'pull' and 'push' characteristics of the decision problem, one by minimizing the number of individuals too close to any container, and the other by minimizing the number of dwellings too far from the respective multi-compartment container. The model determines the number of facilities to be opened, the respective container capacities, their locations, their respective shares of the total waste of each type to be collected, and the dwellings assigned to each facility. The approach proposed was tested with a case study for the historical center of Coimbra city, Portugal, where a large urban renovation project, addressing about 800 buildings, is being undertaken. This paper demonstrates that the models and techniques incorporated in the interactive decision support system (IDSS) can be used to assist a decision maker (DM) in analyzing this complex problem in a realistically sized urban application. Ten solutions consisting of different combinations of underground containers for the disposal of four types of sorted waste in 12 candidate sites, were generated. These solutions and tradeoffs among the objectives are presented to the DM via tables, graphs, color-coded maps and other graphics. The DM can then use this information to 'guide' the IDSS in identifying additional solutions of potential interest. Nevertheless, this research showed that a particular solution with a better objective balance can be identified. The actual sequence of additional solutions generated will depend upon the objectives and preferences of the DM in a specific application.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

High Angle Remediation Antimony, Mercury, and Pu-239/240 Contaminated Soil at LANL near Los Alamos Town Site - 18579

The Manhattan Project and early Cold War era operations at Los Alamos National Laboratory (LANL) left behind soil contaminated with antimony, mercury, and Pu-239/240. TerranearPMC (TPMC) successfully completed a phased three-year high-angle legacy cleanup project to excavate contaminated soil from a steep, difficult-to-access canyon wall in the Upper Los Alamos Canyon Aggregate Area bordering Los Alamos town site. The first and third phases of the project (years 2015 and 2017) utilized a Menzi Muck (spider excavator) in concert with a 275 US ton telescoping crane to remove ∼158 cubic yards (cy) of mercury-contaminated rock and soil (i.e. media) from one solid waste management unit (SWMU) in 2015; and ∼70 cy of antimony contaminated media, and ∼65 cy of collocated mercury and Pu-239/240 contaminated media from two SWMUs in 2017. All soil excavated from the site was containerized in soft-sided 1.0 cy, 1.77 cy, and 3.5 cy IP-1 containers and shipped to an approved, licensed, off-site disposal facility. The crane was also used to move the spider excavator, field personnel, tools, the packaged waste, and restoration materials into and out of the canyon. In 2016, conventional but miniaturized excavation equipment was used to remove about 133 cy of Pu-239/240 contaminated media from four SWMUs west of both the 2015 and 2017 excavations. A total of 398 soil screening, waste composite, characterization/bounding, and confirmation samples were collected during the project. Media that was left in place after the excavation was sampled to confirm that human health concerns and environmental risk issues had been mitigated. This proven technical approach was the preferred method for cleaning up ∼426 cy of contaminated media from steep, difficult-to-access legacy waste sites within the Upper Los Alamos Canyon Aggregate Area between 2015 and 2017. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Catalytic degradation of waste rubbers and plastics over zeolites to produce aromatic hydrocarbons

Catalytic conversion of waste rubbers and plastics into aromatic hydrocarbons is a promising approach to waste management and energy recovery. In the present study, acidic HY zeolites were supported by cobalt, iron, and zirconium, and the catalysts were characterized by powder X-ray diffraction, nitrogen adsorption-desorption, ammonia temperature programmed desorption, X-ray photoelectron spectroscopy, and pyridine-Fourier transform infrared spectroscopy. Here, the catalytic degradation of waste polybutadiene rubbers (BR) was conducted to investigate the degradation mechanism and evaluate the catalytic activity of supported zeolites. Experimental results indicated that HY loaded by zirconium and iron led to a higher content of Lewis acid sites as opposed to cobalt supported one. Compared with the non-catalytic pyrolysis of BR, the zirconium supported HY (Zr/HY) led to a 10-fold increase in aromatic hydrocarbons production with a distinctively high selectivity of 97.9%. A series of waste polymers including waste tires (WT), polyethylene (PE), polycarbonate (PC), and BR, were subjected to catalytic pyrolysis to explore the effects of polymer type on aromatic hydrocarbons generation, and BR was the most effective substrate, with yield enhancement reaching 2.4 over Zr/HY. Catalytic co-pyrolysis of waste rubbers and plastics was conducted to probe the effect of polymer structure on aromatic hydrocarbons formation, where a significant synergistic effect was observed in the PE co-fed with PC run.

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

Integrated Systems Analysis for a Regional Deep Borehole Repository for Commercial Spent Nuclear Fuel - 16291

Some nuclear waste stakeholders have suggested private interim storage facilities (ISF), but little work has been done to consider the cost or logistics for nuclear waste management or disposal facilities smaller than the national scale. This paper performs an integrated system analysis for a deep borehole repository system which would accept waste from a regional subset of the national commercial reactor fleet. The regional subset is comprised of a centrally-located host state with a large SNF inventory surrounded by states with small SNF inventories. The chosen region will contain between one-fifth and one-quarter of the national commercial SNF inventory. Using best estimates of past and future SNF discharges in the sample region, the SNF inventory in the chosen region is characterized by reactor of origin. Basic parameters are then chosen for the regional repository, such as the annual rate of waste emplacement and the time necessary to process borehole-specific waste canisters. Scenarios chosen for analysis test the sensitivity of a variety of system parameters, including the annual rate of waste acceptance, the division of annual waste acceptance among sites, the use of bare fuel transportation canisters, and the size of the regional subset of reactors. Estimated costs for the facility are $13 billion for around 27,000 MT of SNF and $16 billion for around 35,000 MT. This translates to $490,000 per MT and $466,000 per MT, respectively. These unit costs are comparable to the $370,000 per MT estimate for the full 140,000 MT national inventory. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Updated Groundwater Flow Simulations of the Savannah River Site General Separations Area

The groundwater flow model supporting Performance Assessments (PAs) and Composite Analyses (CAs) at the Savannah River Site was significantly revised in 2016 and 2017 using new hydrostratigraphic surfaces, updated well water level calibration targets, and semi-automated model calibration with the PEST optimization code. This model is referred to as “GSA_2016”. This report documents further refinement of the GSA_2016 model in 2018 to incorporate updates to model calibration targets, closure of the H-Area Ash Basin, construction of E-Area Slit Trench operational covers, and plume information from the Mixed Waste Management Facility and Low-Level Radioactive Waste Disposal Facility. Another objective was to lower hydraulic head residuals by adding another calibration zone. The resulting model is referred to as “GSA_2018”.

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS1986

Environmental control/life support system for Space Station

The functional, operational, and design load requirements for the Environmental Control/Life Support System (ECLSS) are described. The ECLSS is divided into two groups: (1) an atmosphere management group and (2) a water and waste management group. The interaction between the ECLSS and the Space Station Habitability System is examined. The cruciform baseline station design, the delta and big T module configuration, and the reference Space Station configuration are evaluated in terms of ECLSS requirements. The distribution of ECLSS equipment in a reference Space Station configuration is studied as a function of initial operating conditions and growth orbit capabilities. The benefits of water electrolysis as a Space Station utility are considered.

Miller, C. W.↗
NASA NTRS1991

Conceptual designs for lunar base life support systems

Three designs for lunar-base life support are described emphasizing the choices of individual processes for initial, intermediate, and advanced systems. Mass balances for the systems are employed to demonstrate the interactions of air, water, and waste loops, and several waste-treatment processes are considered for the initial life-support system. NASA space-station technologies are adopted for the start-up air, water, and waste treatment subsystems, and the intermediate subsystems provide enhanced capabilities. The intermediate waste-management subsystem permits the recovery of reusable waste, and the advanced system provides biological waste treatment. The reduction of resupply requirements and power use are identified as critical issues as is the ability to operate over extended periods.

Dall-Bauman, Liese↗
NASA NTRS2004

Compaction of Space Mission Wastes

The current solid waste management system employed on the International Space Station (ISS) consists of compaction, storage, and disposal. Wastes such plastic food packaging and trash are compacted manually and wrapped in duct tape footballs by the astronauts. Much of the waste is simply loaded either into the empty Russian Progress vehicle for destruction on reentry or into Shuttle for return to Earth. This manual method is wasteful of crew time and does not transition well to far term missions. Different wastes onboard spacecraft vary considerably in their characteristics and in the appropriate method of management. In advanced life support systems for far term missions, recovery of resources such as water from the wastes becomes important. However waste such as plastic food packaging, which constitutes a large fraction of solid waste (roughly 21% on ISS, more on long duration missions), contains minimal recoverable resource. The appropriate management of plastic waste is waste stabilization and volume minimization rather than resource recovery. This paper describes work that has begun at Ames Research Center on development of a heat melt compactor that can be used on near term and future missions, that can minimize crew interaction, and that can handle wastes with a significant plastic composition. The heat melt compactor takes advantage of the low melting point of plastics to compact plastic materials using a combination of heat and pressure. The US Navy has demonstrated successful development of a similar unit for shipboard application. Ames is building upon the basic approach demonstrated by the Navy to develop an advanced heat melt type compactor for space mission type wastes.

Fisher, John↗
DOE OSTI2015

Analysis of the likelihood of post-closure criticality in a geological disposal facility - 14349

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal, controls will ensure that all waste packages are sub-critical, and the containment provided by the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is prevented for some time after disposal. However, as waste packages begin to degrade, fissile and other materials may be mobilised, which could affect the potential for criticality. Therefore, the possibility for the evolution of conditions in a GDF to result in criticality requires assessment. RWM established the 'Likelihood of Criticality' research project to develop qualitative and quantitative arguments to demonstrate that criticality after closure of a GDF is unlikely. The project included consideration of the criticality safety of disposal concepts for intermediate and high level waste, spent fuel, separated plutonium and highly-enriched uranium in different generic geological settings. The results are underpinned by the construction of post-closure criticality scenarios that have been evaluated using high-level judgments about scenario credibility as well as modelling of evolving conditions in a GDF. In the latter case, a probabilistic modelling approach has been undertaken to address uncertainty in the calculation of concentrations and masses of fissile material that could occur at potential accumulation sites. The potential for criticality has been judged based on comparisons of the calculated fissile masses with minimum values required for criticality at different locations in the barrier system. This paper explains the novel approach taken to evaluating the likelihood of criticality in a GDF based on the probabilistic modelling of waste package evolution and plutonium and uranium migration. The role of this research in RWM's work to demonstrate the criticality safety of a GDF in the UK is presented in a companion paper, which also draws on the results of a parallel research project that demonstrates that, even if a criticality were to occur, then its consequences would be insignificant to GDF performance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

A WM Facility for the Processing of Radioactive Materials Arising from the Decommissioning Activities of the Belgian Nuclear Power Plant of Doel - 16042

Doel Nuclear Power Plant is located in the northern part of Belgium in the harbour of Antwerp on the Scheldt River. On this 80-hectare site Electrabel, the Belgian utility, operates four PWRs with a total generation capacity of 2911 MWe. Those nuclear power units were commissioned in 1975 (Doel 1 and 2 - twin units), in 1982 (Doel 3) and in 1985 (Doel 4). In light of the Belgian governments decision to definitively shutdown the oldest units of the site (Doel 1 and 2 - 433 MWe each) in 2015 at the end of their 40-year design lifetime, Electrabel launched a D and D program in partnership with Tractebel Engineering in July 2012. During the D and D preparation phase prior to the final shutdown, studies performed by Tractebel Engineering showed the added benefit of a purpose-built Waste Management Facility (WMF) specifically equipped to process contaminated and activated components generated by the in-situ dismantling activities. The WMF is a key infrastructure contributing to a large extent to the Waste Minimization Policy within the global D and D program. The present paper aims at presenting its strategic, logistic, infra-structural, and design principles. At the end of 2014, the Belgian government and Electrabel entered into discussions to envisage a 10-year lifetime extension of Doel 1 and 2 units. Therefore, Electrabel decided to put the D and D program on hold. Nevertheless, the preparatory studies for the future D and D activities on Doel site are set on solid assumptions and pave the way towards a safe, cost-effective and waste minimizing decommissioning process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Comparing urban solid waste recycling from the viewpoint of urban metabolism based on physical input-output model: A case of Suzhou in China

Highlights: Black-Right-Pointing-Pointer Impacts of solid waste recycling on Suzhou's urban metabolism in 2015 are analyzed. Black-Right-Pointing-Pointer Sludge recycling for biogas is regarded as an accepted method. Black-Right-Pointing-Pointer Technical levels of reusing scrap tires and food wastes should be improved. Black-Right-Pointing-Pointer Other fly ash utilization methods should be exploited. Black-Right-Pointing-Pointer Secondary wastes from reusing food wastes and sludge should be concerned. - Abstract: Investigating impacts of urban solid waste recycling on urban metabolism contributes to sustainable urban solid waste management and urban sustainability. Using a physical input-output model and scenario analysis, urban metabolism of Suzhou in 2015 is predicted and impacts of four categories of solid waste recycling on urban metabolism are illustrated: scrap tire recycling, food waste recycling, fly ash recycling and sludge recycling. Sludge recycling has positive effects on reducing all material flows. Thus, sludge recycling for biogas is regarded as an accepted method. Moreover, technical levels of scrap tire recycling and food waste recycling should be improved to produce positive effects on reducing more material flows. Fly ash recycling for cement production has negative effects on reducing all material flows except solid wastes. Thus, other fly ash utilization methods should be exploited. In addition, the utilization and treatment of secondary wastes from food waste recycling and sludge recycling should be concerned.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Transuranic Waste Burning Potential of Thorium Fuel in a Fast Reactor - 12423

Westinghouse Electric Company (referred to as 'Westinghouse' in the rest of this paper) is proposing a 'back-to-front' approach to overcome the stalemate on nuclear waste management in the US. In this approach, requirements to further the societal acceptance of nuclear waste are such that the ultimate health hazard resulting from the waste package is 'as low as reasonably achievable'. Societal acceptability of nuclear waste can be enhanced by reducing the long-term radiotoxicity of the waste, which is currently driven primarily by the protracted radiotoxicity of the transuranic (TRU) isotopes. Therefore, a transition to a more benign radioactive waste can be accomplished by a fuel cycle capable of consuming the stockpile of TRU 'legacy' waste contained in the LWR Used Nuclear Fuel (UNF) while generating waste which is significantly less radio-toxic than that produced by the current open U-based fuel cycle (once through and variations thereof). Investigation of a fast reactor (FR) operating on a thorium-based fuel cycle, as opposed to the traditional uranium-based is performed. Due to a combination between its neutronic properties and its low position in the actinide chain, thorium not only burns the legacy TRU waste, but it does so with a minimal production of 'new' TRUs. The effectiveness of a thorium-based fast reactor to burn legacy TRU and its flexibility to incorporate various fuels and recycle schemes according to the evolving needs of the transmutation scenario have been investigated. Specifically, the potential for a high TRU burning rate, high U-233 generation rate if so desired and low concurrent production of TRU have been used as metrics for the examined cycles. Core physics simulations of a fast reactor core running on thorium-based fuels and burning an external TRU feed supply have been carried out over multiple cycles of irradiation, separation and reprocessing. The TRU burning capability as well as the core isotopic content have been characterized. Results will be presented showing the potential for thorium to reach a high TRU transmutation rate over a wide variety of fuel types (oxide, metal, nitride and carbide) and transmutation schemes (recycle or partition of in-bred U-233). In addition, a sustainable scheme has been devised to burn the TRU accumulated in the core inventory once the legacy TRU supply has been exhausted, thereby achieving long-term virtually TRU-free. A comprehensive 'back-to-front' approach to the fuel cycle has recently been proposed by Westinghouse which emphasizes achieving 'acceptable', low-radiotoxicity, high-level waste, with the intent not only to satisfy all technical constraints but also to improve public acceptance of nuclear energy. Following this approach, the thorium fuel cycle, due to its low radiotoxicity and high potential for TRU transmutation has been selected as a promising solution. Additional studies not shown here have shown significant reduction of decay heat. The TRU burning potential of the Th-based fuel cycle has been illustrated with a variety of fuel types, using the Toshiba ARR to perform the analysis, including scenarios with continued LWR operation of either uranium fueled or thorium fueled LWRs. These scenarios will afford overall reduction in actinide radiotoxicity, however when the TRU supply is exhausted, a continued U- 235 LWR operation must be assumed to provide TRU makeup feed. This scenario will never reach the characteristically low TRU content of a closed thorium fuel cycle with its associated potential benefits on waste radiotoxicity, as exemplified by the transition scenario studied. At present, the cases studied indicate ThC as a potential fuel for maximizing TRU burning, while ThN with nitrogen enriched to 95% N-15 shows the highest breeding potential. As a result, a transition scenario with ThN was developed to show that a sustainable, closed Th-cycle can be achieved starting from burning the legacy TRU stock and completing the transmutation of the residual TRU remaining in the core inventory after the legacy TRU external supply has been exhausted. The radiotoxicity of the actinide waste during the various phases has been characterized, showing the beneficial effect of the decreasing content of TRU in the recycled fuel as the transition to a closed Th-based fuel cycle is undertaken. Due to the back-to-front nature of the proposed methodology, detailed designs are not the first step taken when assessing a fuel cycle scenario potential. As a result, design refinement is still required and should be expected in future studies. Moreover, significant safety assessment, including determination of associated reactivity coefficients, fuel and reprocessing feasibility studies and economic assessments will still be needed for a more comprehensive and meaningful comparison against other potential solutions. With the above considerations in mind, the potential advantages of thorium fuelled reactors on HLW management optimization lead us to believe that thorium fuelled reactor systems can play a significant role in the future and deserve further consideration. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Treatment of G1 Baskets at the CEA Marcoule Site - 12027

In the dismantling program for the first-generation French reactors in accordance with the nonproliferation treaty, the CEA is in charge of cleanup and dismantling operations for the facilities at Marcoule, including the decladding units. The G1 decladding was built between 1955 and 1957 in order to de-clad spent fuel elements from the G1 plutonium-producing reactor and prepare them for dissolution. The facility was also used for interim storage of G1, G2 and G3 fuel dissolution baskets, which had been used during plant operation for transfer (from the decladding facility to the UP1 plant) and/or dissolution of spent fuel elements. One of the cleanup projects involves recovery of the baskets, which will be cut up, sorted, and conditioned in metal bins. The bins will be immobilized with cement grout, then transferred to the onsite solid waste conditioning facility (CDS) and to the repository operated by the French National Radioactive Waste Management Agency (ANDRA). The project is now in progress, after special safety permits were issued and measurement stations and dedicated tools were developed to handle all types of baskets (which differed according to their origin and use). The disposal of all the baskets is scheduled to last 2 years and will produce 55 metal waste bins. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

Y-12 Groundwater Protection Program Groundwater And Surface Water Sampling And Analysis Plan For Calendar Year 2011

This plan provides a description of the groundwater and surface water quality monitoring activities planned for calendar year (CY) 2011 at the U.S. Department of Energy (DOE) Y-12 National Security Complex (Y-12) that will be managed by the Y-12 Groundwater Protection Program (GWPP). Groundwater and surface water monitoring performed by the GWPP during CY 2011 will be in accordance with requirements of DOE Order 540.1A and the following goals: (1) to protect the worker, the public, and the environment; (2) to maintain surveillance of existing and potential groundwater contamination sources; (3) to provide for the early detection of groundwater contamination and determine the quality of groundwater and surface water where contaminants are most likely to migrate beyond the Oak Ridge Reservation property line; (4) to identify and characterize long-term trends in groundwater quality at Y-12; and (5) to provide data to support decisions concerning the management and protection of groundwater resources. Groundwater and surface water monitoring during CY 2011 will be performed primarily in three hydrogeologic regimes at Y-12: the Bear Creek Hydrogeologic Regime (Bear Creek Regime), the Upper East Fork Poplar Creek Hydrogeologic Regime (East Fork Regime), and the Chestnut Ridge Hydrogeologic Regime (Chestnut Ridge Regime). The Bear Creek and East Fork regimes are located in Bear Creek Valley and the Chestnut Ridge Regime is located south of Y-12 (Figure A.1). Additional surface water monitoring will be performed north of Pine Ridge along the boundary of the Oak Ridge Reservation. Modifications to the CY 2011 monitoring program may be necessary during implementation. Changes in programmatic requirements may alter the analytes specified for selected monitoring wells or may add or remove wells from the planned monitoring network. All modifications to the monitoring program will be approved by the Y-12 GWPP manager and documented as addenda to this sampling and analysis plan. The following sections of this report provide details regarding the CY 2011 groundwater and surface water monitoring activities. Section 2 describes the monitoring locations in each regime and the processes used to select the sampling locations. A description of the field measurements and laboratory analytes is provided in Section 3. Sample collection methods and procedures are described in Section 4, and Section 5 lists the documents cited for more detailed operational and technical information. The narrative sections of the report reference several appendices. Figures (maps and diagrams) and tables (excluding data summary tables presented in the narrative sections) are in Appendix A and Appendix B, respectively. Groundwater Monitoring Schedules (when issued throughout CY 2011) will be inserted in Appendix C, and addenda to this plan (if issued) will be inserted in Appendix D. Laboratory requirements (bottle lists, holding times, etc.) are provided in Appendix E, and an approved Waste Management Plan is provided in Appendix F.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2012

Design of a Particle Shadow-graph Velocimetry and Size (PSVS) System to Determine Particle Size and Density Distributions in Hanford Nuclear Tank Wastes - 12280

Accurate particle size and density distributions for nuclear tank waste materials are essential information that helps determine the engineering requirements for a host of waste management unit operations (e.g., tank mixing, pipeline transport, and filtration). The most prevalent approach for determining particle size and density distribution is highly laborious and involves identifying individual particles using scanning electron microscope/x-ray diffraction and then acquiring the density of the materials from the technical literature. Other methods simply approximate individual particle densities by assuming chemical composition, rather than obtaining actual measurements of particle density. To overcome these limitations, a Particle Shadow-graph Velocimetry and Size (PSVS) system has been designed to simultaneously obtain particle size and density distributions for a broad range of Hanford tank waste materials existing as both individual particles and agglomerates. The PSVS system uses optical hardware, a temperature-controlled settling column, and particle introduction chamber to accurately and reproducibly obtain images of settling particles. Image analysis software provides a highly accurate determination of both particle terminal velocity and equivalent spherical particle diameter. The particle density is then calculated from Newton's terminal settling theory. The PSVS system was designed to accurately image particle/agglomerate sizes between 10 and 1000 μm and particle/agglomerate densities ranging from 1.4 to 11.5 g/cm{sup 3}, where the maximum terminal velocity does not exceed 10 cm/s. Preliminary testing was completed with standard materials and results were in good agreement with terminal settling theory. Recent results of this method development are presented, as well as experimental design. The primary goal of these PSVS system tests was to obtain accurate and reproducible particle size and velocity measurements to estimate particle densities within 20 percent of direct measured density values. The average percent error observed in particle densities was 4.8, 8.2, and 25.8 percent for S1D3 (high-density soda-lime glass), S3D4 (zirconia), and S1D2 (soda-lime glass) beads, respectively. The maximum measured particle density error observed was 14.5, 21.0, and 58.4, for S1D3, S3D4, and S1D2 beads, respectively. Multiple sources of error can exist that contribute to inaccurate estimates of particle density using the PSVS approach. Two significant influences observed were thermal convective currents in the settling column fluid and non-spherical (irregular) particle shapes in the standard test materials. Experimental results using buoyant beads suggest thermal convective currents can influence the movement of settling particles as much as 4 mm/s. These thermal convective currents are clearly a significant concern for smaller and less-dense particles/agglomerates which have terminal velocities of the same magnitude. Significant effort went into the design and fabrication of a thermal control system for the PSVS settling column. However, small temperature variations (<0.3 deg. C) are difficult to prevent and measure accurately. Therefore, although designing the thermal control system was necessary, it is not considered the only solution to minimize error due to thermal convective currents. The use of small diameter (10 to 27 μm) polyethylene buoyant beads (in low concentrations) during actual settling experiments may be necessary to provide insight into particle-particle interactions and fluid velocity correction factors to account for potential influences from thermal convective currents. Future testing will incorporate the buoyant beads into the analysis technique when the suspension fluid is chemically compatible. The use of the Newton's law to calculate particle/agglomerate densities assumes spherical particle shapes. As the actual particle/agglomerates diverge from spherical, the margin of error increases. The selection of high quality (in density and shape) monodispersed particle standards was a direct attempt to reduce the uncertainties associated with non-spherical or irregular particles and to only observe the PSVS method and analysis error. While investigating the source of data scatter for the S1D3 bead (nominal 700 μm) standard, significant irregular particles where discovered. Yet, a subsequent investigation into the S3D4 bead standard was also conducted and it was determined that 60 percent of the particles had a 0.9 to 0.8 centricity, 32 percent were between 0.8 and 0.7, only 7 percent were between 1.0 and 0.9, and the remaining 1 percent of the particles had centricities between 0.7 and 0.6. Contrast this with the S1D3 bead standard, where 100 percent of particles had centricities between 0.95 and 0.98, upon applying the centricity filter. The influence of the non-spherical shape of the S3D4 bead standard is evident when comparing a 3.9 percent average density error for S1D3 glass bead standard with an 8.3 percent average density error for the S3D4 bead standard. Model corrections for non-spherical shape must be implemented to account for these measurement errors during future method validation and system performance assessments. In addition, further work is needed to resolve the higher than expected velocities for the smaller S1D2 particles. Initial measurements conducted with buoyant beads as fluid velocity tracers indicate that fluid velocities can be subtracted from the increased particle velocities to determine the particle terminal velocity, regardless of the surrounding fluid velocity. However, this approach is in its infancy and will need to be carefully and methodically validated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Processing and disposition of remote-handled transuranic liquid waste generated at Oak Ridge National Laboratory - 15113

Oak Ridge National Laboratory (ORNL) produces transuranic (TRU) isotopes for the US Department of Energy's (US DoE's) heavy-element research program and other programs for medical and industrial applications. As a result, the research facilities generate remote-handled TRU liquid low-level waste (LLLW). This LLLW is presently collected and mixed with larger volumes of legacy and environmental management waste streams for future treatment. Once the legacy waste is disposed of as part of the US DOE Environmental Management program, the large-scale waste treatment processes will be decommissioned, and new capabilities will be needed to characterize, treat, package, and dispose of the newly generated LLLW at ORNL. Various approaches are being considered for addressing treatment and disposal of the future LLLW streams. A variety of processing methods and final waste forms is under consideration. The impacts of the chemical and radionuclide compositions of the waste stream are being evaluated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗