Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “RADIOACTIVE WASTE PROCESSING”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,639 records · Page 91
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Update on Investigations of Viability of Cold Spray and FSW as a Spent Nuclear Fuel Dry Storage Canister Mitigation Tool

The purpose of this work is to investigate the use of cold spray and friction stir welding or processing (FSW/P) for repair and mitigation of chloride-induced stress corrosion cracking (CISCC) in dry cask storage system (DCSS) canisters to ensure integrity far beyond their original license period. This report provides a viability analysis of cold spray and friction stir welding for repair and fabrication of DCSS canisters. This report presents experimental work and a research strategy to optimize and evaluate cold spray and FSW/P for CISCC resistance. The literature reports that the following three conditions must exist for CISCC to occur: tensile stress, corrosive environment, and susceptible material. These conditions must be understood and accounted for while evaluating CISCC susceptibility. Furthermore, non-obvious factors, such as surface conditions and geometric confinement, can dramatically increase the intensity of one or more of the three listed conditions for CISCC. Arc welding should be considered for either temporary or permanent repair when combined with a mitigation technology. It is known that alloys used in the existing inventory of spent nuclear fuel storage canisters become sensitized in areas around fusion welds. Technologies, such as cold spray or FSW, can mitigate the sensitization caused by fusion welds. FSW/P and cold spray, potential alternatives for fabrication and repair, are solid-phase processes. No melting occurs and the energy input can be controlled such that no harm is done to the base metal during these processes. FSW/P should be considered for use in new fabrication of canisters and repair of existing canisters. Low heat input and grain refinement result in material with improved mechanical properties and comparable generalized corrosion performance relative to base metal. FSW/P offers the potential of through-thickness CISCC resistance. Cold spray should be considered for production, mitigation, and repair of canisters. Cold spray has demonstrated the ability to deposit stainless steel, Inconel alloys, and other alloys on stainless steel 304L such that galvanic potential is matched and resistance to pitting is improved. The technical basis for cold spray mitigation and repair within the overpack using remote robotic equipment is established. A non-obvious application to the research community is the sealing of crevices. All reported instances of CISCC in the fielded nuclear applications were associated with crevice corrosion. This can be avoided by cold spraying to seal crevices. The bulk of the work planned for FY 2020 will be generating coupons for CISCC testing by collaborative projects at Pacific Northwest National Laboratory and Sandia National Laboratories. First an exploratory study will be executed that will include screening studies and process optimization. This task will identify the effect of process parameters on microstructure and CISCC resistance. Optimized parameters will be used to produce a smaller number of coupons for full testing and evaluation. Test methods for evaluating CISCC resistance must meet the following criteria: (1) normalized surface geometry and exposed process microstructure, (2) account for residual stresses, (3) be in a controlled and appropriate test environment, and (4) be numerically quantifiable and repeatable. Failure to meet these criteria could lead to results that cannot be used for comparative analysis of processes and materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

DOE's Environmental Management Site-Specific Advisory Board: The Roles, Work, and Assessment of the Constituent Local Boards - 13587

The charter for the Department of Energy's Environmental Management (EM) Site-Specific Advisory Board (SSAB) was approved under the Federal Advisory Committee Act (FACA) in 1994. With a unique mandate to provide public input on issues associated with the cleanup of nuclear legacy sites in the U.S., the EM SSAB comprises eight local boards, which are based at major EM sites. While each board is unique to the community in which it is located and reflects the diversity of the local population, the boards are governed by FACA, related regulations, and DOE policies that are intended to standardize agency advisory board operations. The EM SSAB local boards are made up of a diverse group of citizens who want to understand the mission and goals of the EM program and to help EM achieve those goals for the benefit of their communities. Some are quite passionate about their mission; others need to be coaxed into active participation. Maintaining productive relationships and a supportive environment for effective board operations is the challenge of board management for DOE EM and the board members themselves. DOE draws on research findings and best practices literature from academics and practitioners in the field of public involvement in its board management practices. The EM SSAB is also evaluated annually under the law to ensure that the investment of taxpayer dollars in the board is warranted in light of the contributions of the board. Further evaluation takes place at the agency and site levels in order to identify what aspects of board functioning the agency and board members find important to its success and to address areas where improvement is needed. Board contributions, compliance factors, and measurable outcomes related to board products and process areas are key to agency commitment to ongoing support of the boards and to participant satisfaction and thus continued member involvement. In addition to evaluation of these factors in improving board effectiveness, the agency draws on the experience of members to create best practices for the EM SSAB, as a unique form of public involvement. Four areas that have been identified by local board Chairpersons as important to their local board operations are - Enhancing communication between technical and non-technical board members; - Building on common ground toward recommendations; - Public involvement in EM SSAB local board activities; - The EM SSAB annual work plan process. The first three areas are addressed below by current or former chairpersons of the EM SSAB: Ralph Phelps, former Chairperson of the Northern New Mexico Citizens' Advisory Board; Susan Leckband, former Chairperson and current Vice Chairperson of the Hanford Advisory Board; and Val Francis, Vice Chairperson of the Portsmouth (PORTS) SSAB. In addition, Eric Roberts, facilitator of the PORTS SSAB, has contributed to the section on public involvement. In a separate paper for this session, Ralph Young, Chairperson of the Paducah Citizens' Advisory Board addresses the EM SSAB annual work plan process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Comparison of U.S. Special Permit/Exemption/Authorization Processes for Licensing Type B Radioactive Material Packages - 17513

In the United States, there are three regulators (agencies) associated with licensing Type B radioactive material packages for shipping. They are the U.S. Department of Energy (DOE), the United States Nuclear Regulatory Commission (NRC), and the United States Department of Transportation (DOT). Each regulator has requirements and guidance which implement United States Code of Federal Regulations 10 CFR 71 - Packaging and Transportation of Radioactive Material and 49 CFR 171 through 180 - Hazardous Materials Regulations. Periodically, select packages may not clearly fit within the requirements or guidance. In such cases, the planned shipper can petition the applicable agency for a special permit/exemption/authorization approving the shipment with the specified deviations from the requirements or guidance. If equivalent safety is successfully demonstrated to the applicable regulator, such authorization may be obtained. This paper compares the: 1) DOE exemption process; 2) the NRC special package authorization process; and, 3) the DOT special permit process. (authors)

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

Lessons Learned from FUSRAP - 16200

The US DOE Office of Legacy Management (LM) is the long-term steward for 90 sites remediated under numerous regulatory regimes including the Formerly Utilized Sites Remedial Action Program (FUSRAP) sites. In addition, LM holds considerable historical information, gathered in the 1970's, to determine site eligibility for remediation under FUSRAP. To date, 29 FUSRAP sites are in LM's inventory of sites for long-term surveillance and maintenance (LTS and M), and 25 are with the US Army Corps of Engineers (USACE) for remediation or in the process of being transitioned to LM. It is forecasted that 13 FUSRAP sites will transfer from the USACE to LM over the next 10 years; however, the timing of the transfers is strongly dependent upon federal funding of the ongoing remedial actions. Historically, FUSRAP sites were generally cleaned up for unrestricted industrial use or remediated to the cleanup standards at that time, and their use remained unchanged. Today, these sites as well as the adjacent properties are now changing or envisioned to have changes in land use, typically from industrial to commercial or residential uses. The implication of land-use change affects DOE's LTS and M responsibility for the sites under LM stewardship as well as the planning for the additional sites scheduled to transition in time. Coinciding with land-use changes at or near FUSRAP sites is an increased community awareness of these sites. As property development increases near FUSRAP sites, the general public and interested stakeholders regularly inquire about the sufficiency of cleanups that impact their neighborhoods and communities. LM has used this experience to address a series of lessons learned to improve our program management in light of the changing conditions of our sites. We describe these lessons learned as (1) improved stakeholder relations, (2) enhanced LTS and M requirements for the sites, and (3) greater involvement in the transition process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Development of Advanced Remote Cutting Technology for Reactor Internals Segmentation - 18274

Since the Three Mile Island decommissioning in the 1980's, Westinghouse has been involved in dismantling and segmentation of commercial reactors. With dozens of components segmented and packaged across the globe, Westinghouse has continued to evaluate the state of the art cutting technologies for their applications in the nuclear industry. While there have been well documented issues of certain cutting technologies in field applications, there are advancements of these technologies which prove them applicable for specific areas of the decommissioning process. During 2017, Westinghouse embarked on a revision of its plasma arc cutting (PAC) technology as part of an extensive innovation program for reactor internals segmentation. While these systems were used in Yankee Rowe and Three Mile Island, new plasma arc cutting systems improve the versatility and have much better precision of cutting. While bringing this technology into the future, Westinghouse is adapting the mounting and application so that it is compatible with the existing mechanical cutting portfolio. This means that in the event that a mechanical cutting technique is unable to perform as intended, a plasma cutting alternative will be available with minimal time to switch cutting methods. A test program for underwater PAC was performed, and the test gave a good reference of cutting speed and handling in different plate thicknesses which can be used when evaluating the use of the new cutting technology for future projects. PAC technology has proven that it could have a future in reactor vessel internals segmentation projects as a complement to the existing mechanical cutting methods under water. Thicknesses up to at least 25 mm can be cut with a much higher speed than the existing mechanical cutting methods in use. This and other cutting technology developments as part of the innovation program provide exciting new possibilities for dismantling and segmentation projects. While each of the techniques presents definite benefits for operators and, more importantly, customers, there are risks that need to be mitigated through engineered processes and controls. (authors)

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

Engaging Canadian Indigenous Communities and Public Stakeholders Through a US Site Benchmarking Trip - 19358

Canadian Nuclear Laboratories (CNL) is proposing a new approach to decommission the WR-1 reactor at its Whiteshell site near Pinawa, Manitoba Canada. To tackle WR-1 CNL is proposing in situ decommissioning, this new approach is subject to a Canadian federal environmental assessment. As part of this CNL is engaging with Indigenous communities and public stakeholders, both regionally and across the country. Nuclear in situ decommissioning is new to Canada but has been used numerous times in the United States. Throughout the environmental assessment process CNL has considered the feedback public and Indigenous groups have provided and has tried to enhance the way information is shared and respond where possible and reasonable. Feedback from some communities has been related to a greater understanding of what the in situ method looks like, how it is monitored, will it perform, and anxiety about having it in the community. CNL sought to respond to this feedback by coordinating a benchmarking trip to an in situ reactor in Hallam, Nebraska. The visit was coordinated around an annual inspection. Several regional stakeholder and Indigenous communities joined CNL on the benchmarking trip. The group witnessed and learned what a decommissioned in situ reactor looked like, how it was performing and the effect on the surrounding communities. The group learned important lessons during the visit, key among them that the entombment is working as designed and has had zero significant detectable effects to the surrounding community. All attendees are now better informed on what CNL's proposal for WR-1 represents and are in a position to better inform their communities and the process moving forward. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Testing and Deployment of Remote Systems to Remove Highly Radioactive Soil at Hanford's 324 Building - 18043

U.S. Department of Energy (DOE) contractor CH2M HILL Plateau Remediation Company (CHPRC) is in the process of procuring, testing and deploying remotely operated tools to safely remove highly contaminated soil from beneath the 324 Building on the Hanford Site in southeastern Washington State. The 324 Building operated for more than 40 years, completing a variety of high-dose engineering development and demonstration activities inside six hot cells that shielded the workers from the highly radioactive materials. Activities to decommission, decontaminate and demolish the facility began in 2008. In 2010, soil contamination was found in the soil under one of the hot cells, and building deactivation was paused to evaluate the soil contamination. Sampling efforts, using borehole technology, revealed significant under-building contamination with dose rates measuring has high as 130 Sv/hr (13,000 R/hr). The highly-contaminated soil has to be removed before traditional open air deactivation activities can continue. CHPRC completed proof of concept testing and built a full-scale test facility to demonstrate the effectiveness of three primary tools that will accomplish the floor and soil removal. The tools include a commercially available concrete saw, and a commercially available excavator arm that have been modified to operate remotely in the high-radiation environment within the hot-cell. In order to reduce the wear on the hot-cell overhead cranes, a remote operated rail and cart system has been designed to transport soil and debris out of the hot-cell to an airlock area where the waste will be separated, according to its level of radioactivity. A camera and light system has been designed to provide visual and audio coverage of the hot-cell area that will allow the operations to proceed from remote control stations. This paper will describe the primary remote operated equipment designed for the project and the challenges and processes for developing, testing and deploying the equipment necessary to excavate the soil from below the 324 Building. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Upgrade to Ion Exchange Modeling for Removal of Technetium from Hanford Waste Using SuperLig ® 639 Resin

This report documents the development and application of computer models to describe the sorption of pertechnetate [TcO 4 - ], and its surrogate perrhenate [ReO 4 - ], on SuperLig ® 639 resin. Two models have been developed: 1) A thermodynamic isotherm model, based on experimental data, that predicts [TcO 4 - ] and [ReO 4 - ] sorption as a function of solution composition and temperature and 2) A column model that uses the isotherm calculated by the first model to simulate the performance of a full-scale sorption process. The isotherm model provides a synthesis of experimental data collected from many different sources to give a best estimate prediction of the behavior of the pertechnetate-SuperLig ® 639 system and an estimate of the uncertainty in this prediction. The column model provides a prediction of the expected performance of the plant process by determining the volume of waste solution that can be processed based on process design parameters such as column size, flow rate and resin physical properties.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Upgrade to Ion Exchange Modeling for Removal of Technetium from Hanford Waste Using SuperLig ® 639 Resin

This report documents the development and application of computer models to describe the sorption of pertechnetate [TcO 4 ⁻], and its surrogate perrhenate [ReO 4 ⁻], on SuperLig ® 639 resin. Two models have been developed: 1) A thermodynamic isotherm model, based on experimental data, that predicts [TcO 4 ⁻] and [ReO 4 ⁻] sorption as a function of solution composition and temperature and 2) A column model that uses the isotherm calculated by the first model to simulate the performance of a full-scale sorption process. The isotherm model provides a synthesis of experimental data collected from many different sources to give a best estimate prediction of the behavior of the pertechnetate-SuperLig ® 639 system and an estimate of the uncertainty in this prediction. The column model provides a prediction of the expected performance of the plant process by determining the volume of waste solution that can be processed based on process design parameters such as column size, flow rate and resin physical properties.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Upgrade to Ion Exchange Modeling for Removal of Technetium from Hanford Waste Using SuperLig ® 639 Resin

This report documents the development and application of computer models to describe the sorption of pertechnetate [TcO 4 ⁻], and its surrogate perrhenate [ReO 4 ⁻], on SuperLig ® 639 resin. Two models have been developed: 1) A thermodynamic isotherm model, based on experimental data, that predicts [TcO 4 ⁻] and [ReO 4 ⁻] sorption as a function of solution composition and temperature and 2) A column model that uses the isotherm calculated by the first model to simulate the performance of a full-scale sorption process. The isotherm model provides a synthesis of experimental data collected from many different sources to give a best estimate prediction of the behavior of the pertechnetate-SuperLig ® 639 system and an estimate of the uncertainty in this prediction. The column model provides a prediction of the expected performance of the plant process by determining the volume of waste solution that can be processed based on process design parameters such as column size, flow rate and resin physical properties.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Waste management strategy for cost effective and environmentally friendly NPP decommissioning

Decommissioning of nuclear power plants generates large volumes of radioactive or potentially radioactive waste. The proper management of the dismantling waste plays an important role for the time needed for the dismantling phase and thus is critical to the decommissioning cost. An efficient and thorough process for inventorying, characterization and categorization of the waste provides a sound basis for the planning process. As part of comprehensive decommissioning studies for Nordic NPPs, Westinghouse has developed the decommissioning inventories that have been used for estimations of the duration of specific work packages and the corresponding costs. As part of creating the design basis for a national repository for decommissioning waste, the total production of different categories of waste packages has also been predicted. Studsvik has developed a risk based concept for categorization and handling of the generated waste using six different categories with a span from extremely small risk for radiological contamination to high level waste. The two companies have recently joined their skills in the area of decommissioning on selected market in a consortium named ndcon to further strengthen the proposed process. Depending on the risk for radiological contamination or the radiological properties and other properties of importance for waste management, treatment routes are proposed with well-defined and proven methods for on-site or off-site treatment, activity determination and conditioning. The system is based on a graded approach philosophy aiming for high confidence and sustainability, aiming for re-use and recycling where found applicable. The objective is to establish a process where all dismantled material has a pre-determined treatment route. These routes should through measurements, categorization, treatment, conditioning, intermediate storage and final disposal be designed to provide a steady, un-disturbed flow of material to avoid interruptions. Bottle-necks in the process causes increased space requirements and will have negative impact on the project schedule, which increases not only the cost but also the dose exposure to personnel. For these reasons it is critical to create a process that transfers material into conditioned waste ready for disposal as quickly as possible. To a certain extent the decommissioning program should be led by the waste management process. With the objective to reduce time for handling of dismantled material at site and to efficiently and environmental-friendly use waste management methods (clearance for re-use followed by clearance for recycling), the costs for the plant decommissioning could be reduced as well as time needed for performing the decommissioning project. Also, risks for delays would be reduced with a well-defined handling scheme which limits surprises. Delays are a major cost driver for decommissioning projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Waste Management Strategy for Dismantling Waste to Reduce Costs for Power Plant Decommissioning - 13543

Decommissioning of nuclear power plants generates large volumes of radioactive or potentially radioactive waste. The proper management of the dismantling waste plays an important role for the time needed for the dismantling phase and thus is critical to the decommissioning cost. An efficient and thorough process for inventorying, characterization and categorization of the waste provides a sound basis for the planning process. As part of comprehensive decommissioning studies for Nordic NPPs, Westinghouse has developed the decommissioning inventories that have been used for estimations of the duration of specific work packages and the corresponding costs. As part of creating the design basis for a national repository for decommissioning waste, the total production of different categories of waste packages has also been predicted. Studsvik has developed a risk based concept for categorization and handling of the generated waste using six different categories with a span from extremely small risk for radiological contamination to high level waste. The two companies have recently joined their skills in the area of decommissioning on selected market in a consortium named 'ndcon' to further strengthen the proposed process. Depending on the risk for radiological contamination or the radiological properties and other properties of importance for waste management, treatment routes are proposed with well-defined and proven methods for on-site or off-site treatment, activity determination and conditioning. The system is based on a graded approach philosophy aiming for high confidence and sustainability, aiming for re-use and recycling where found applicable. The objective is to establish a process where all dismantled material has a pre-determined treatment route. These routes should through measurements, categorization, treatment, conditioning, intermediate storage and final disposal be designed to provide a steady, un-disturbed flow of material to avoid interruptions. Bottle-necks in the process causes increased space requirements and will have negative impact on the project schedule, which increases not only the cost but also the dose exposure to personnel. For these reasons it is critical to create a process that transfers material into conditioned waste ready for disposal as quickly as possible. To a certain extent the decommissioning program should be led by the waste management process. With the objective to reduce time for handling of dismantled material at site and to efficiently and environmental-friendly use waste management methods (clearance for re-use followed by clearance for recycling), the costs for the plant decommissioning could be reduced as well as time needed for performing the decommissioning project. Also, risks for delays would be reduced with a well-defined handling scheme which limits surprises. Delays are a major cost driver for decommissioning projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Performance Assessment of a New Gamma Spectroscopy Toolkit: Peak Doctor and the Spectral Nondestructive Assay Platform (SNAP) - 17397

High resolution gamma spectroscopy (HRGS) is a widely used technique for non-destructive characterization of nuclear materials, and assay of radioactive waste. The ability to accurately evaluate raw spectral data is an important step in producing high-quality results. A spectral analysis application known as Peak Doctor has been developed under a Cooperative Research and Development Agreement (CRADA) between LANL and Pajarito Scientific Corporation (PSC). The software has undergone several years of testing and validation by the Radioactive Waste Characterization Team of the LANL Waste Management Services Group (WM-SVS). PeakDoctor is designed to work with the Spectral Nondestructive Assay Platform (SNAP{sup TM}) to provide a complete HRGS analysis engine. SNAP{sup TM} performs model specific corrections for attenuation and geometry. These applications are the major components of a new toolkit designed primarily for portable far-field non-destructive assay. Peak Doctor performs a very accurate peak fit analysis with the goal of improving upon the capabilities of existing commercially available routines. It is designed to be user-configurable and can be customized to any given type of detector and any given type of spectrum. Peak Doctor was developed to resolve many of the inherent complexities in gamma ray spectra including overlapping peaks, complex multiplet regions, large step peaks, backscatter regions and Compton edges. Unlike most spectral analysis platforms, Peak Doctor does not rely on the energy calibration and detector resolution parameters that are saved by the operator in the spectrum file. Instead, key parameters are pre-defined for individual detectors. The software fits a continuum using a series of splines between a series of junctions or 'knots'. Fitting techniques are used to join the splines together to create a near-continuous multi-element curve. Optimizing the location of the knots is one of Peak Doctor's most powerful features. The knots are initially configured using a pre-defined list and are removed or added automatically during the fitting process. Knots are also created when new features are identified such as Step Peaks and Compton Edges. The user can also manually specify knots in order to improve the continuum fit. The initial fit is improved on by removing any outlier regions (which are converted to peaks) and the continuum fitting process is repeated. At this stage, an experienced user can work interactively with the fit by manually converting peak regions to continuum or vice versa. The final step is to fit a series of peaks above the continuum based on detector parameters. Single and multiplet peaks are then processed by either the tailing (piecewise) or non-tailing (Gaussian) Levenberg Marquardt non-linear peak fitting routine depending on the user defined selection. Only a rough knowledge of the detector characteristics (energy calibration and resolution) is required for the fitting process to proceed with good accuracy. Any peaks that are missed can be manually added by the user and will also be fitted. Discontinuities in the peak / continuum fit can be rapidly identified by reviewing the 'residuals scale' plot. For waste streams that are well characterized, it is often the case that routinely encountered spectra will have similar peaks and continuum features. For these cases a handy 'One Step' method is available that proceeds in an automated fashion without user interaction. Peak Doctor has been configured with a set of default spectrum types including plutonium / transuranic waste (including Weapons Grade Pu, Heat Source Pu, Am-241, Am-243 and Np-237), depleted or enriched uranium, nuclear power plant / intermediate level waste (with fission and activation products such as Cs-137 and Co-60) as well as Naturally Occurring Radioactive Materials (NORM). The user can easily add new spectrum types for specific waste streams or other types of radiological / special nuclear materials. The output from PeakDoctor (including net peak area, centroid energy and a list of minimum detectable activity levels at pre-defined regions) is used by SNAP{sup TM} to complete the data analysis process. Isotope activities are calculated using configurable template-driven geometry and attenuation correction methods. The software accounts for matrix and container wall effects, collimator / angular corrections, detector efficiency (based on calibration curves) and stand-off distance. SNAP{sup TM} uses published algorithms incorporating a series of decision points during its operation to allow the analyst to provide logical input to guide the software during the analysis of each unique spectrum. The Peak Doctor / SNAP{sup TM} toolkit has been extensively tested and validated under a variety of conditions using 'benchmark' reference spectra. Test scenarios reported include HPGe-based measurements of well characterized uranium and plutonium standards and radioactive sources with a variety of different geometries and matrix materials including surrogate drums and boxes. (authors)

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

ANALYSIS OF THE LEACHING EFFICIENCY OF INHIBITED WATER AND TANK SIMULANT IN REMOVING RESIDUES ON THERMOWELL PIPES

A key component for the accelerated implementation and operation of the Salt Waste Processing Facility (SWPF) is the recovery of Tank 48H. Tank 48H is a type IIIA tank with a maximum capacity of 1.3 million gallons. Video inspection of the tank showed that a film of solid material adhered to the tank internal walls and structures between 69 inch and 150 inch levels. From the video inspection, the solid film thickness was estimated to be 1mm, which corresponds to {approx}33 kg of TPB salts (as 20 wt% insoluble solids) (1). This film material is expected to be easily removed by single-rinse, slurry pump operation during Tank 48H TPB disposition via aggregation processing. A similar success was achieved for Tank 49H TPB dispositioning, with slurry pumps operating almost continuously for approximately 6 months, after which time the tank was inspected and the film was found to be removed. The major components of the Tank 49H film were soluble solids - Na{sub 3}H(CO{sub 3}){sub 2} (Hydrated Sodium Carbonate, aka: Trona), Al(OH){sub 3} (Aluminum Hydroxide, aka: Gibbsite), NaTPB (Sodium Tetraphenylborate), NaNO{sub 3} (Sodium Nitrate) and NaNO{sub 2} (Sodium Nitrite) (2). Although the Tank 48H film is expected to be primarily soluble solids, it may not behave the same as the Tank 49H film. There is a risk that material on the internal surfaces of Tank 48H could not be easily removed. As a risk mitigation activity, the chemical composition and leachability of the Tank 48H film are being evaluated prior to initiating tank aggregation. This task investigated the dissolution characteristics of Tank 48H solid film deposits in inhibited water and DWPF recycle. To this end, SRNL received four separate 23-inch long thermowell-conductivity pipe samples which were removed from the tank 48H D2 risers in order to determine: (1) the thickness of the solid film deposit, (2) the chemical composition of the film deposits, and (3) the leaching behavior of the solid film deposit in inhibited water (IW) and in DWPF recycle simulant (3).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Investigation of Coupled Processes and Impact of High Temperature Limits in Argillite Rock

Shale and clay-rich geological formations have been considered as potential host rock for geological disposal of high-level radioactive waste throughout the world, because of their low permeability, low diffusion coefficient, high retention capacity for radionuclides, and capability to self-seal fractures. Permeabilities arising from the primary porosity of these rock types are low because of the extremely fine-grained constituents and the very small pores associated with the fine-grained material. Permeability values are typically less than 10 microdarcies (10-17 m2 or about 10-10 m/s equivalent hydraulic conductivity) and are often two to four orders of magnitude lower. While fractures can occur, clay and shale often demonstrate the tendency to self-seal fractures, which reduces the effects of fractures on bulk permeability. This occurs as a result of swelling from the increased water potential in fractures, the generation of fracture infilling materials and mineral precipitation, and plastic deformation of the rock (Mazurek et al., 2003). The low permeability of clay and shale rock are well-known in the hydrogeology community where these rock types represent aquitards that severely limit groundwater movement, and in petroleum geology, where they act as caprocks limiting the rise of buoyant petroleum fluids. Other favorable characteristics of clay/shale rock are the strong sorptive behavior for many radionuclides and (in saturated systems) the low flow rates, which typically lead to reducing conditions because of the lack of oxygen transport from the surface. Clay and shale rock also act to chemically buffer the effects of materials introduced through repository construction, operation, and emplaced materials (Arcos et al., 2008).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Public involvement on closure of Asse II radioactive waste repository in Germany

From 1967 to 1978, about 125,800 barrels of low- and intermediate level waste were disposed of - nominally for research purposes - in the former 'Asse' salt mine which had before been used for the production of potash for many years. Since 1988 an inflow of brine is being observed which will cause dangers of flooding and of a collapse due to salt weakening and dissolution if it should increase. Since several years the closure of the Asse repository is planned with the objective to prevent the flooding and collapse of the mine and the release of radioactive substances to the biosphere. The first concept that was presented by the former operator, however, seemed completely unacceptable to regional representatives from politics and NGOs. Their activities against these plans made the project a top issue on the political agenda from the federal to the local level. The paper traces the main reasons which lead to the severe safety problems in the past as well as relevant changes in the governance system today. A focus is put on the process for public involvement in which the Citizens' Advisory Group 'A2B' forms the core measure. Its structure and framework, experience and results, expectations from inside and outside perspectives are presented. Furthermore the question is tackled how far this process can serve as an example for a participatory approach in a siting process for a geological repository for high active waste which can be expected to be highly contested in the affected regions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Image Classification Using Convolutional Neural Networks to Automate Visual Inspection of CCO Containers (Rev. 1)

The Savannah River Site is automating the receipt and inspection of Criticality Control Overpack drums to aid in nuclear waste disposal. The Advanced Engineering group is automating this receipt and inspection process to assist the site’s goals. Part of this process includes visually inspecting CCO drums to ensure that no defects or security risks are present. This research focuses on developing a machine learning model to automatically classify drums as passing or failing inspection. The machine learning model was implemented using the open-source Tensorflow library and uses a Convolutional Neural Network architecture to extract features and differentiate between images. The model was able to achieve an average of 85% accuracy on a dataset of 288 drums. This project’s goal was to prove the validity of computer vision in an automation process and provide a starting place for continued research.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Processing and Disposition of Special Actinide Target Materials - 17071

The DOE manages an inventory of materials that contain a range of long-lived radioisotopes that were produced from the 1960's through the 1980's by irradiating targets in high-flux reactors at the SRS to produce special heavy isotopes for DOE programmatic use, scientific research, and industrial and medical applications. Among the products were Cf-249, -250, -251, and -252; heavy curium (=50% Cm-246); and Pu-242 and -244. The irradiation of the Mark-18A (Mk-18A) targets at extremely high fluxes produced the world's supply of heavy curium and Pu-244. Approximately one fourth of the Mk-18A targets were processed in the 1970's to recover and enrich the material that has been used as the global source of Pu-244 and heavy curium by industry and researchers. These isotopes are still in demand, and the available supply of purified material is dwindling. A program has been initiated to recover heavy curium and Pu-244 from the remaining stored unprocessed Mk-18A targets. Improved flowsheets to process the materials to recover Pu-244 and heavy curium are being developed. Evaluations are also underway to determine whether additional materials present in the targets in much smaller quantities, such as long-lived fission product (LLFP) isotopes, can be recovered for research and development applications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗