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At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Efficient Mesofluidic Separation of Large Particles in nuclear slurries - 20408

Efficient and effective particle separation is essential to cleanup of many nuclear wastes. For example, particle separation may be used to accelerate settle-decant bottlenecks that limit the throughput necessary to achieve the U.S. Department of Energy waste cleanup mission in a timely manner. Particle separation may be used to protect downstream processing equipment from pulses of solids that may be encountered near (within a few feet of) liquid-solid interfaces during waste transfer operations, thereby preventing work stoppages. Effective particle separation may permit efficient sludge washing. These are only a few of the ways in which particle separation is important. Yet, efficient and effective particle separation in nuclear processing environments remains challenging. For example, settle-decant operations permit larger and heavier particles to settle, leaving smaller, lighter particles suspended. However, settle-decant operations are slow, convection currents may resuspend solids, hindered settling of multicomponent slurries remains incompletely understood, and predictive models for settling in graduated cylinders fail to match limited observations of settling in large waste tanks. Additionally, pumping operations can cause turbulent resuspension of particles when the pump intake is close to the settled solids layer. Filtration techniques that use membranes or partially permeable barriers retain larger particles, permitting smaller particles to permeate. However, filters, including dead-end filtration, are prone to clogging and caking, operate at elevated pressures due to minimal void volume, and increase in pressure during operations, which require more control systems. Therefore, the need for high throughput particle separation techniques that operate with modest pressure drops persists. A novel mesofluidic separator presents the opportunity to effectively and efficiently accelerate the waste cleanup mission. The separator separates large particles from process streams across a broad range of particle sizes and has no moving parts or media to replace, regenerate, or clean. This separator design has an unusually large void volume, permitting operation at much higher flow rates (and lower pressures) than traditional filtration (e.g., dead-end filtration). Industrial-scale flow rates have been demonstrated. In performance testing, mesofluidic separators operate at might flow rates (>90 gpm (0.0006 m{sup 3}/s) in piping 3 inches (0.08 m) in diameter; Re>10{sup 5}), with modest pressure drop (∼25 psi (170 KPa) in testing). In complex, aggregating waste simulants, the separator loses <25% of flow projected over months without back pulsing or chemical cleaning. Mesofluidic separation presents infrastructure advantages, reduces risk, and provides mission impact. Infrastructure advantages include implementation within existing transfer systems, plug and play without facility modifications to safety systems, and operation at low pressures. The separator is flexible in location, may be positioned within or outside of waste tanks, and may replace or augment dead-end and cross-flow filters. Mesofluidic separation reduces risk by minimizing waste (no media to replace or regenerate and no cleaning chemicals to handle and dispose) and by minimizing filter change-out consistent with as low as reasonably achievable (ALARA) exposure to workers. The potential mission impact of these separators is substantial. The separator has potential to support, simplify, and accelerate in-farm transfers and waste feed delivery. Furthermore, the separator can free up settling and holding tanks in the Direct Feed Low-Activity Waste (DFLAW) mission by reducing or eliminating post-transfer settling and wait times with the potential to free up 1 Mgal (4000 m{sup 3}) of double-shell tank space. This paper quantitatively compares mesofluidic separation to dead-end filtration, discusses scale-up results, and considers the separator's potential to efficiently and effectively reduce the long-term environmental ability of particulate-rich nuclear wastes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI1991

Nuclear Safety [Vol. 32, No. 1, January-March 1991]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 The Safety Review and Approval Process for Space Nuclear Power Sources, G. L. Bennett; 19 Report on the American Nuclear Society International Topical Meeting: "The Safety, Status, and Future of Non-Commercial Reactors and Irradiation Facilities", E. G. Silver; 35 Book Review: Fission Product Transport Processes in Reactor Accidents, Proceedings to the International Centre tor Heat and Mass Processes T. S. Kress; 38 Fast Reactor Technology in the 1990s: A Summary of the 1990 International Fast Reactor Safety Meeting, A. E. Levin; ACCIDENT ANALYSIS: 56 Effects of Chemical Phenomena on LWR Severe Accident Fission Product Behavior, A. P. Malinauskas and T. S. Kress; CONTROL AND INSTRUMENTATION: 65 Technical Note: Safety Parameter Display Systems—10 Years Later, R. J. Eckenrode; 68 Potential Application of Neural Networks to the Operation of Nuclear Power Plants, R. E. Uhrig; DESIGN FEATURES: 80 Twenty-First DOE/NRC Nuclear Air-Cleaning Conference, R. R. Bellamy, D. W. Moeller, and M. W. First; 91 Impact of an Apparent Radiation Embrittlement Rate on the Life Expectancy of PWR Vessel Supports, R. D. Cheverton, G. C. Robinson, W. E. Pennell, and R. K. Nanstad; ENVIRONMENTAL EFFECTS: 103 Technical Note: The Impact of Offsite Factors on the Safety Performance of Small Nuclear Power Plants, Yu. D. Baranaev and A. N. Viktorov; WASTE AND SPENT FUEL MANAGEMENT: 109 Activities Related to Waste Management, Compiled by E. G. Silver; OPERATING EXPERIENCES: 118 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 121 Selected Safety-Related Events, Compiled by G. A. Murphy; 123 Operating U.S. Power Reactors, Compiled by E. G. Silver; RECENT DEVELOPMENTS: 140 General Administrative Activities, Compiled by E. G. Silver; 150 Reports, Standards, and Safety Guides, D. S. Queener; 155 Status of Power-Reactor Licensing Activities, Compiled by E. G. Silver; 157 Proposed Rule Changes as of Sept. 30, 1990; ANNOUNCEMENTS: 79 MIT Offers Summer Program on Nuclear Power Reactor Safety; 108 Harvard School of Public Health Offers Several Short Courses; 139 Short Course and Workshop on Nuclear Criticality Safety at University of New Mexico; 160 SCK/CEN Announces Training Course on Emergency Planning and Response; 161 The Authors; 164 Indexes to Nuclear Safety, Volume 31

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Feedback on Forty-year Long Clean-up Operations of a Contaminated Soil for Environmental Purpose - 20026

One of the CEA facilities in France is the place of miscellaneous mapping and clean-up operations since the 1980's. The final purpose of the conducted work deals with the environmental remediation of the building. In France, the absence of a regulatory framework for the management of sites polluted by radioactive substances, especially as the lack of release thresholds, has led to develop different approaches of soil radiological characterization. The geostatistical approach is part of them. Preliminary investigations are an unavoidable step prior to clearance and remediation. First, historical analysis enables to define site perimeters to investigate as well as to precisely locate expected contaminated parts. For radioactive pollutants, the media (water, soil, air) and transfer routes participate in the definition of the investigation perimeter and thus also need to be identified through a geological study. At the same time, functional analysis leads to choose the best measurement means which could provide useful information when combined with visual inspection. An equipment well adapted to easy-to-measure radionuclides as gamma emitters, such as gamma probe or on-site gamma spectrometry, is usually selected as a non-destructive assay. Coupled with spatial positions, processed on-site radiological data can be mapped. Data processing consists in mapping and kriging before a geostatistical analysis in order to identify the zones of interest to be targeted. Selected zones are not only linked to high level count rates but also to low level ones so that a pollution-free reference is known in the area. This data processing also enables to categorize waste types according to their origin and contamination levels. An expected volume of different waste then follows as well as the waste characterization equipments. In-depth investigations, e.g. core drillings and samplings, naturally ensue to aim at remediation optimization through the assessment of environmental impact. After clean-up operations, non-destructive assays as well as destructive samplings enable to check the effective clearance. Following this methodology in the present remediation work, results of the first non-destructive assay campaign highlighted a long-lived transuranic radionuclides contamination in the concrete flagstone. Successive non-destructive assays were then performed in the defined zones of interest. Count rates measured with a surface probe were first mapped. The best localization of destructive assays, i.e. core drillings at a 1 meter depth, came from this surface mapping followed by a geostatistical analysis. The concrete flagstone was totally removed. Then, the depth of soil to remove under the flagstone was optimized from the results of radiological activities measured in core drillings samples. A new surface mapping was then drawn after this partial soil excavation. The sand and demolition rubble samples, homogeneously constituted from the flagstone fragments, were analyzed by gamma spectrometry. Meanwhile, waste drums and other containers were also measured by gamma spectrometry and passive neutron measurement devices. All these results have brought miscellaneous pieces of information. In this context, major issues appear in terms of physical constraints such as the premises tininess as well as of soil sampling techniques and in terms of measurement performances assessment and the representativeness of homogeneous samples. The choice of radiological soil and waste characterization devices is described in the light of enhanced performances by keeping in mind radiation protection requirements and a willing to always optimize waste categories (very low level waste) and volume. The improvement of the characterization methods appears through this forty-year long work while sharing the feedback of encountered difficulties. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Impacts of Fast Critical Assembly Fuel Discards on Liquid Waste Processes

The Savannah River Nuclear Solutions (SRNS) Fast Critical Assembly (FCA) mission is reestablishing the electrolytic dissolver for processing of Pu and Pu-U materials clad in stainless steel (SS). H-Canyon is planning to dissolve and neutralize FCA fuel without recovering the special nuclear material (i.e., Pu) prior to discarding to the Concentration, Storage, and Transfer Facilities (CSTF) operated by the Savannah River Mission Completion (SRMC) Liquid Waste (LW) Organization. The FCA discards will be combined with sludge in the CSTF after Low Temperature Aluminum Dissolution (LTAD) if needed. The combined waste will be washed, concentrated, and vitrified at the Defense Waste Processing Facility (DWPF). The high level waste canisters produced will be temporarily stored in the Glass Waste Storage Buildings prior to transfer to a future federal repository. Decants from LTAD and sludge washing will be combined with DWPF recycle and dissolved salt cake to prepare salt batches for processing in the Salt Waste Processing Facility. The resulting decontaminated salt solution will be processed in the Saltstone Production Facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Modeling Stress-Induced Pore Water Pressures in The Vadose Zone Beneath a Composite-Lined Landfill - 20029

A finite-element model was developed to evaluate mechanisms contributing to positive pore pressures measured with sealed pressure transducers in the geological buffer beneath the Environmental Management Waste Management Facility, a composite-lined mixed waste disposal facility operated by the US Department of Energy. The geological buffer is a 3-m-thick engineered fine-textured layer directly beneath the Environmental Management Waste Management Facility's composite liner, and above the groundwater table. The model accounts for changes in pore water pressure resulting from (i) loading imposed by waste placed on the overlying liner, (ii) moistening of the geological buffer due to equilibration with the underlying geological materials, and (iii) fluctuations in the elevation of the underlying groundwater table. Pore water pressures predicted by the model are in good agreement with pore water pressures measured in the field. The predictions confirm that positive pore water pressures recorded by the sealed pressure transducers in the geological buffer are excess pore water pressures induced by the vertical normal stress imposed by waste placed on the liner, and are not due to a rise in the groundwater table. Simulations also showed that two additional years of filling would further increase the pore water pressure without any change in elevation of the groundwater table. The geological buffer remained unsaturated during the simulation, with a B-coefficient (parameter indicative of the degree of saturation) similar to that computed from the field-measured pore water pressures and waste filling records. Larger increases in pore water pressure were observed when the geological buffer was assumed to have higher initial saturation, as was observed in the field data. Incorporating seasonal fluctuations in the groundwater table beneath the geological buffer in the model resulted in predictions of small seasonal oscillation in the pore water pressure at the measurement location, similar to seasonal oscillations observed in the field. Predictions made with the model indicate that the dissipation of the excess pore water pressures will occur over decades due to the low hydraulic conductivity of the geological buffer material. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

A Rapid Sequential Separation Method for Determination of Actinides and TENORM in Fracking Wastes - 20223

With recent advances in unconventional drilling technology in the US and around the world, the risks of environmental contamination and exposure due to technologically enhanced naturally occurring radioactive materials (TENORM) from fracking wastes have also surfaced. Although the Permian Basin (southeastern New Mexico, western Texas) has long been a modest producer of oil, the advent of fracking technology a decade ago turned it into hot property for producers. A 2018 assessment of undiscovered, technically recoverable continuous oil and gas resources by the U.S. Geological Survey estimates an average 46.3 billion barrels of oil and 281 trillion cubic feet of gas. This is likely to increase the use of unconventional drilling and TENORM generation. TENORM concentrations in oil and gas exploration and production waste can be several hundred to several thousand times higher than background TENORM concentrations. Both the thorium and uranium decay chains contribute to airborne radionuclides arising from the radon gas escaping the ground and subsequently decaying as airborne particulates. Additionally, radon decay products {sup 210}Pb and {sup 210}Po can build up in scale on the internal surfaces of oil and gas handling pipes and in sludge in refineries, becoming potential inhalation and ingestion hazards for workers. Furthermore, southeastern NM is also home to world's only licensed and operating transuranic nuclear waste repository, the DoE's Waste Isolation Pilot Plant (WIPP). Plutonium isotopes ({sup 239+240}Pu) and {sup 241}Am, are expected to account for more than 99% of the total radioactivity scheduled for disposal in the WIPP repository. Thus, an improved understanding of the environmental fate and transport of TENORM, liberated by unconventional drilling, is essential to assess how best to protect individuals and the environment. In this context, accurate measurement of TENORM and actinides in environmental and biological samples is essential. In this presentation, a new sequential method for the separation and pre-concentration of actinides (Pu, Am, Np) and TENORM (Po, U and Th) derived from oil and gas exploration is proposed. The TEVA method involves a rapid co-precipitation step to remove matrix interferences from the samples, followed by plutonium oxidation state adjustment to Pu (IV) and an incubation period of ∼ 1 hour at 50-60 deg. C to allow the resultant Po (II) to oxidize into Po(IV). The polonium, neptunium, thorium and plutonium are then separated on a TEVA column, while americium and uranium are separated on a TRU column. After separation, the alpha counting source was prepared by micro-precipitation with copper sulfide (CuS) for polonium and neodymium fluoride (NdF{sub 3}) for actinides. The efficiency and reliability of the procedures were tested by analyzing filter, drinking water and frack sand samples. The method is simple, robust and can be performed quickly with excellent removal of interferences, high chemical recovery and very good alpha peak resolution. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

In-Situ Uranium Source Treatment Performance Assessment Elements - 20247

Field-scale in situ injection of an aqueous polyphosphate amendment was conducted to decrease uranium leaching from a vadose zone source to groundwater near the Columbia River at the U.S. Department of Energy's Hanford Site. After an initial phase of application in November 2015, this treatment was applied as an enhanced attenuation approach in September 2018, targeting uranium in a subsurface zone where water table periodically rises in response to the seasonal high river stage. Uranium within this zone is aqueous, adsorbed, and in low- and high-solubility precipitates and is mobilized into the groundwater at high river stage, resulting in a persistent groundwater plume. The polyphosphate amendment creates phosphate coatings on uranium (potentially including low solubility uranium-phosphate precipitates) that decrease the uranium leached into groundwater when the zone is periodically rewetted to improve the ability of natural attenuation to meet groundwater concentration objectives. For treatment, the amendment was injected into a network of wells within the periodically rewetted source area. Multiple lines of evidence are being applied to evaluate the performance of the phosphate treatment. Three elements of this performance assessment include 1) identifying the distribution of injected phosphate amendment to the targeted treatment zone using cross-borehole electrical resistivity tomography and comparisons of phosphate precipitates between pre- and post-treatment samples, 2) demonstrating the functional reduction in uranium mobility and leachability using laboratory tests with pre- and post- treatment sediments from the source zone, and 3) evaluating the presence of low-solubility uranium phosphate minerals and/or coatings associated with reduced uranium mobility. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

The Influence of Alkalinity on the Uptake of Cs{sup +} and Sr{sup 2+} by Cation-substituted Natisites in Sodium-bearing Conditions - 20336

The titanosilicate natisite (Na{sub 2}TiSiO{sub 5}) is a kinetic phase of the mineral sitinakite (Na{sub 2}Ti{sub 2}O{sub 3}SiO{sub 4}.2H{sub 2}O), which is a reference material in the removal of Cs and Sr from radioactive high-level waste. Natisite is disregarded in the literature as a candidate sorbent for Cs and Sr, despite being more thermally stable than sitinakite, which is a critical property for this application. Replacing portion of the Ti in natisite by other metals is believed to enhance natisite sorption properties. In nuclear waste remediation, Cs and Sr are contained in high-salinity liquid wastes that can either be highly acidic or alkaline. In the present study, Al-, Sn-, and Zr-natisites were synthesized, and compared to pure natisite and sitinakite in batch experiments. Five concentrations of NaOH and NaNO{sub 3} (i.e. source of Na{sup +}) mimicking conditions of high alkalinity, and competing Na{sup +} ions, respectively, were evaluated. Sorption results demonstrate that sitinakite is generally more effective than all four natisites in removing Cs and Sr. However, sitinakite uptake mechanism seems to deteriorate at increasing concentrations of base and Na{sup +}. Although less selective than sitinakite, all natisite sorbents provided a less dramatic decline in Cs and Sr uptake through increasing molarities. Overall, there was an improvement in the sorption of Cs and Sr by the metal-substituted materials over natisite and, in neither of the testing solutions, natisite outperformed its substituted variants. In highly alkaline solutions, Sn- and Zr-Natisite provided for exceptional removal for Sr, removing more than twice the amount sorbed by sitinakite at 0.1 M NaOH. The results obtained for Sr uptake by all four natisite materials in alkaline solutions are promising, especially considering the highly alkaline nature of wastes from the nuclear industry. Future studies should investigate the sorption mechanisms responsible for natisite selectivity for Sr at high pH. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Thermal Process Technology for Nuclear Applications - 20379

ANSTO's Synroc technology has been developed to provide a safe, secure matrix for the immobilization and final disposal of radioactive waste. Synroc technology will be used to manage radioactive wastes from the production of the radioisotope Molybdenum-99 (Mo-99). This paper shall outline various stages of the process development with specific reference to the thermal treatment technology of calcination. Calcination is a key step in the Synroc process [1-2].The rotary thermal processing system includes: an advanced heating element design for increased robustness and ease of remote operation and maintenance, an enhanced modular design of components for ease of remote maintenance in a hot cell and in compliance with hot cell radioactive environment requirements for safety, reliability and maintainability. In addition to thermal treatment of waste from nuclear medicine production, this technology provides solutions for a variety of nuclear materials processing applications including sintering UO{sub 2} pellets for reactor fuel rods, oxidation of UO{sub 2} pellets, swarf, and powder to U{sub 3}O{sub 8}, de-nitration of Uranyl nitrate and hydrofluorination of UO{sub 2} pellets. The paper will also discuss thermal processing solutions for a range of nuclear applications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Testing of Crystalline Silicotitanate to Support Tank-Side Cesium Removal System Operations - 20458

The direct feed of waste from the Hanford tank farms to the Low-Activity Waste Facility at the Hanford Waste Treatment and Immobilization Plant requires an intermediate treatment step that prepares the waste for vitrification. Washington River Protection Solutions (WRPS) selected a near-tank treatment system known as the Tank-Side Cesium Removal (TSCR) system to perform the required intermediate treatment functions. The approach in the TSCR system is to filter waste supernatant with a sintered metal dead-end filter and then use a series of ion exchange columns to remove cesium; this system has several similarities to the Tank Closure Cesium Removal system that has been deployed on the Savannah River Site. The ion exchange media proposed for use is crystalline silicotitanate (CST), which is a non-elutable media with a high affinity for cesium. Because the media is non-elutable, loaded columns will be blown down with compressed air, moved into interim storage, and replaced with new columns when cesium capacity is reached. The assumed extent and rate of drying that could be achieved in the TSCR column geometry lacked confirmation by experimental data. In addition, once the columns are loaded with cesium, they generate flammable gases (primarily hydrogen) via radiolysis acting on any resident liquids, i.e., moisture, remaining in the media bed. An assessment of available information on CST concluded that gas generation data that bounds expected operational conditions were needed to support the TSCR system safety basis and planned operations. Working with WRPS, Pacific Northwest National Laboratory (PNNL) designed and conducted testing with CST media to address the need for (a) representative in-column drying data, and (b) bounding flammable gas generation data. Drying testing was performed using a full-height column with a diameter of approximately 2 inches with injected air at temperatures of 18 deg. C and 30 deg. C. Gas generation testing was conducted in an engineered bunker that allowed simultaneous irradiation of up to eight samples at a time. These two experimental approaches used at PNNL for the CST testing are summarized and the major outcomes are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Modeling Efforts to Gain Insight into Historical Leak Events from the Single-Shell Tank A-105 in the Hanford 241-A Tank Farm - 20113

The 241-A Tank Farm is a single-shell tank (SST) farm constructed to store process waste from Hanford nuclear operations. Millions of gallons of nuclear waste were stored in the 241-A Tank Farm SSTs and some of the SSTs leaked in the past. In addition, spills and pipeline leaks during transfers and storage and intentional discharges to cribs and trenches resulted in releasing waste to the ground. Tank liner leaks are referred to as 'leaks' and all other discharges to the soil are referred to as 'releases.' Liquid waste that could be removed by pumping has been removed from all of the SSTs to reduce the potential for future leaks. The most significant historical leak event in 241-A Tank Farm occurred in 1965 when tank A-105 experienced a pressurized steam event. This event resulted in damage to the inner steel liner flooring of A-105, which was separated from the sidewalls over part of its circumference and buckled up to 2.5 m, vertically. During this event up to 7,570 L of contaminated liquid may have leaked to the adjacent soil. This leak occurred under extreme temperature and pressure conditions. Modeling efforts to better understand the nature and extent of historical leaks from subsurface SSTs that were used to store highly radioactive, self-boiling liquid wastes at 241-A Tank Farm have recently been undertaken. These efforts have compiled the available historical information to support a preliminary non-isothermal, multiphase flow and chemical transport modeling effort to re-create the conditions under which the leaks occurred and to formulate a conceptual model as to the extent and distribution of leaked radioactive contaminants in the adjacent soil material. The focus of this work was on developing an understanding of the key features, processes and bounding conditions related to tank A-105 leak events that occurred in the 1960's. The activities include estimation of leak composition, non-isothermal multiphase flow and transport modeling, and geochemical modeling. The STOMP{sup C} Water-Air-Energy modeling code was utilized to implement the three-dimensional representation of the subsurface tank and the surrounding flow field. Once constructed, the model was used to evaluate the conceptual understanding of leaks originating from different parts of the tank under the elevated temperature boundary conditions that were imposed by surrounding tanks at the time of the historical leaks. The chemical evolution of the liquid waste was also evaluated as it leaked from the tank under transient pressure and temperature gradients. The results of the preliminary evaluation suggested the presence of a heat-pipe effect beneath tank A-105 in which water vapor at an elevated temperature is driven away from the base of the tank to a position where water vapor cools, condenses and is then drawn back toward the tank by the strong capillary attraction of the dry soil. A non-sorbing contaminant (Tc-99) was introduced into this flow field to better understand the potential distribution patterns of leaked contaminants that may have occurred during the historical leak events. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Structural Effects of Postulated Hydrogen Explosions in Process Piping and Vessels - 20274

At Savannah River Site (SRS), preparations to startup new liquid waste operations and recent advances in understanding thermolytic hydrogen generation have led to several recent studies on the structural effects of postulated hydrogen explosions. This paper reports on seven representative piping and vessel cases that are evaluated for deflagrations, detonations, and deflagration-to-detonation transition (DDT) assuming the initial mixture is stoichiometric hydrogen and air. The evaluations are based on the approach used by the Department of Defense for analyzing structures subjected to accidental explosions and include inelasticity, dynamic structural amplification, and material strain-rate effects. Evaluation criteria consider the potential for repeated explosions causing fatigue but also consider that some degree of permanent plastic deformation is acceptable as long as rupture is prevented. Specific conclusions depend on the assumed pre-ignition pressure. If the pre-ignition pressure is atmospheric and if the components are designed to a code from the American Society of Mechanical Engineers, the components evaluated are expected to plastically deform but not rupture under the explosion scenarios considered. (authors)

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

Follow-on Report of Analysis of Approaches to Supplemental Treatment of Low–Activity Waste at the Hanford Nuclear Reservation (Volumes I & II)

The Hanford Site, in southeast Washington State, is preparing to disposition approximately 56,000,000 gallons (56 Mgal) of radioactive and chemically hazardous wastes currently stored in underground tanks at the site. Tank wastes will be divided into a high-activity fraction and a low-activity fraction for subsequent treatment and disposition. A waste processing and treatment facility, the Waste Treatment and Immobilization Plant (WTP), will include the high-level waste (HLW) vitrification facility (WTP HLW Vitrification Facility) for immobilizing the high-activity fraction and a low-activity waste (LAW) vitrification facility (WTP LAW Vitrification Facility) for immobilizing the low-activity fraction. Both facilities will use vitrification technology to immobilize the Hanford tank wastes in a glass waste form.

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

A Historical Review of the Safe Transport of Spent Nuclear Fuel

This report meets the requirements of M3 milestone M3SF-21OR020302066 for the Office of Integrated Waste Management (IWM) and is a revision to M3 milestone M3FT-16OR090402028 for the former Nuclear Fuels Storage and Transportation Planning Project (NFST), “Safety Record of SNF Shipments.” The US Department of Energy (DOE) has since established the IWM, which builds on the work begun by NFST, to develop an integrated waste management system for spent nuclear fuel (SNF), including the development of a large-scale transportation system for the safe transport of SNF to storage or disposal facilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Overview of Upcoming Process Improvement Efforts at the Saltstone Facilities at the Savannah River Site - 20474

The Saltstone Facilities at the Savannah River Site (SRS) process and dispose of low activity salt solution waste generated by other liquid waste facilities. Today in the Liquid Waste System (LWS), approximately 125 million liters of salt waste remain to be processed. The Saltstone Production Facility (SPF) receives the decontaminated salt waste, which is mixed with dry feeds consisting of cement, fly ash and slag to produce the saltstone. The mixed slurry is pumped to the Saltstone Disposal Units (SDUs) for safe disposal in the Saltstone Disposal Facility (SDF). To support continuous Salt Waste Processing Facility (SWPF) operations at annual processing volumes many times higher than present, significant modifications are required to ensure that the SPF has the capacity to support the throughput of salt waste. These changes include several infrastructure upgrades that are being worked through the Enhanced Low Activity Waste Disposal (ELAWD) project. In addition to infrastructure upgrades, Savannah River Remediation (SRR), the Savannah River Site's Liquid Waste Contractor for the U.S. Department of Energy, has identified a flowsheet improvement opportunity of implementing a cement-free saltstone formulation to eliminate cement from the current three component formulation. This two-component formulation would simplify the procurement, transport, off-loading, and storage of the premix materials, and increase the storage capacity for the individual components. This flowsheet, when implemented, will reduce operational risk and improve throughput necessary to achieve disposal rates necessary for SWPF operations. Research has been completed to confirm the feasibility of this formulation, with the new cement-free flowsheet expected to be implemented in the next few years. Improvements have also been identified downstream of the SPF to increase SDU utilization. Specifically, a concentrated effort has taken place to raise the fill height in SDU 6 to the full disposal height, 13 meters, to maximize the usage of each SDU in support of the liquid waste disposal mission. This improvement has been made through additional thermal modeling and improved process understanding to create a SDU flammability model and control strategy more representative of facility operations. This work will allow for the completion of construction and operational readiness for SDU 7 while filling SDU 6 to ensure there is no lapse in available space once SWPF starts up. A summary of these improvements as they relate to increased throughput will be provided along with associated status. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

A Preliminary Radiological Risk Assessment Model for Disposition of Remote-Handled Transuranic Wastes at Los Alamos National Laboratory Area G - 20116

The U.S. Department of Energy (DOE) operates a low-level radioactive waste (LLW) disposal site at Material Disposal Area G, in Los Alamos, New Mexico, USA. Area G has been the primary LLW disposal site for Los Alamos National Laboratory (LANL) since the 1960's. In addition to LLW, Area G is host to a variety of other wastes, the disposition of which must be determined before closure of the site. A probabilistic Radiological Risk Assessment (RRA) for Area G is used in order to support decision making regarding some wastes that are not addressed in the extant Area G Performance Assessment (PA) and Composite Analysis (CA). Between 1979 and 1987, 33 special shafts were augered into the Bandelier Tuff at Area G. This volcanic tuff is present across Pajarito Plateau on the eastern slopes of the Jemez Mountains, and varies widely in its consistency, from weakly indurated non-welded layers to welded layers that uphold the mesa cliffs of the Plateau. These mesas are home to LANL, Area G, and the townsites of Los Alamos and White Rock, with residences about 1400 m from Area G. The 33 Shafts were lined with steel casing, and contain remote-handled (RH) transuranic wastes (TRU) resulting from experiments and analysis performed in special glove boxes at the Chemistry and Metallurgy Research (CMR) facility at LANL. Some of these wastes originated as used nuclear fuel. The purpose of the Area G RRA is to evaluate the potential future risk to humans and the environment from the RH TRU in the 33 Shafts in the context of the risk associated with the surrounding wastes at Area G. The analysis is responsive to expectations outlined in DOE Order 458.1, Radiation Protection of the Public and the Environment, and is informed by the Manual and Guidance accompanying DOE O 435.1, Radioactive Waste Management. Because the waste meets the definition of TRU, the regulatory context necessarily takes into consideration the regulation governing the disposal of TRU from the U.S. Environmental Protection Agency (EPA): 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes. Given the broader regulatory context for the RRA, the analysis is subject to different assumptions from those made in the existing DOE O 435.1 PA and CA, such as allowing for future occupation of the site. The analysis begins with a comprehensive evaluation of features, events, processes, and exposure scenarios (FEPS) for Area G and the wastes it contains. These FEPSs are screened to eliminate from further consideration those of extremely low probability and/or consequence, and a conceptual site model (CSM) is subsequently developed. The scope and structure of the Area G RRA Model is informed by this CSM, and the Area G RRA Model is developed using the GoldSim systems analysis modeling platform. This paper presents the initial version of a defensible, transparent, and reasonably realistic model, which is based on the state of knowledge of the wastes, the site, and the FEPSs that govern contaminant transport from wastes into the environment and subsequent exposures to humans and other biota. Probabilistic model input distributions represent uncertainties inherent in the real and modeled systems. The results of the Area G RRA Model inform decisions regarding the disposition of the RH TRU in the 33 Shafts. (authors)

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DOE OSTI2022

SB14 Source Reduction Evaluation Review and Plan for the DOE California Sites

The Department of Energy (DOE) is the owner of multiple facilities in Northern California. The facilities include Lawrence Livermore National Laboratory (LLNL), Lawrence Berkeley National Laboratory (LBNL), Sandia National Laboratories/California (SNL/CA) and SLAC National Accelerator Laboratory (SLAC) among other sites. Through their operations, the facilities generate hazardous waste and, thereby, are subject to the requirements of Chapter 31 of the Title 22 California Code of Regulations, Waste Minimization. The Northern California sites are primarily research and development facilities in the areas relating to national security, high-energy physics, bioscience and the environment.

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