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At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

CLOSURE REPORT FOR CORRECTIVE ACTION UNIT 577 AREA 5 CHROMIUM CONTAINING WASTE DISPOSAL CELLS

Report will be submitted to the state of Nevada Corrective Action Unit (CAU) 577, “Area 5 Chromium Containing Waste Disposal Cells,” includes five low-level waste cells at the Nevada National Security Site (NNSS) Area 5 Radioactive Waste Management Site (RWMS) where buried waste received from Nuclear Fuel Services, Inc. (NFS), was subsequently determined to contain chromium that exceeded the Toxicity Characteristic Leaching Procedure regulatory limit, which would require the waste to carry hazardous waste code D007. CAU 577 was created to satisfy the requirements of the Settlement Agreement (SA) executed between the Nevada Division of Environmental Protection (NDEP) and the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office on April 25, 2019. The SA required that the chromium-containing waste received from NFS would be addressed following the closure process laid out in the Federal Facility Agreement and Consent Order (FFACO). The FFACO process ensures proper closure of the chromium-containing waste cells and documentation of that closure through FFACO-type documents (NDEP 2019). As stipulated in Paragraph 5C of the SA, a Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) was prepared that identified and evaluated risks and corrective action alternatives (CAAs) for CAU 577, provided the plan for implementing the recommended CAA, and presented the post-closure plan (U.S. Department of Energy [DOE] Environmental Management [EM] Nevada Program 2021).

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

Systems and methods for enhancing isolation of high-temperature reactor containments

A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

Peterson, Per F.↗
DOE OSTI2022

Additively manufactured tamper evident container

A tamper evident container includes a container body having walls surrounding a cavity. The walls are composed of a non-electrically conductive material and at least one integrity sensor that is embedded in the non-electrically conductive material. A monitoring circuit is configured to monitor the tamper evident container using the at least one integrity sensor and determine a state-of-health of the tamper evident container.

Bernardin, John D.↗
DOE OSTI2018

Performance Evaluation of Decontamination Foam Containing Mesoporous silica NPs - 18453

It is necessary to secure decontamination technology that is easy to apply and minimizes the generation of secondary waste for large-scale equipment and large-area decontamination, which are difficult to be applied structurally when massive repairs or the decommissioning of a nuclear facility are required. In this respect, decontamination foam has a potentially wide application in the removal of contaminants from metallic walls, overhead surfaces, and elements of complex components and facilities. The decontamination efficiency can be enhanced by improving the contact time between the chemical decontamination agents and the contaminated surface through the introduction of stabilizers into the decontamination foam. In this study, the stability of a decontamination foam containing mesoporous silica NPs as a stabilizer and the decontamination performance of radioactive isotope contaminated specimens were evaluated. A process for waste liquid treatment after foam decontamination was also proposed. Decontamination performance tests using RI contaminated specimens were carried out in a decontamination foam prepared only with a solution containing only 1 wt% EM100 surfactant, and in a decontamination foam prepared with a solution containing 1 wt% EM100 surfactant and 1 wt% of various silica NPs. The stability of the decontamination foam was increased owing to the addition of various types of silica NPs. In particular, the liquid film thickness of the decontamination foam containing mesoporous silica NPs was the largest. As a result, the decontamination coefficient for RI contaminated specimens increased as the thickness of the foam liquid film increased owing to the longer contact time between the RI contaminated specimen and the chemical decontamination agent as the thickness of the foam liquid film increased. (authors)

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

Radionuclide Behaviour During Thermal Treatment of LLW and ILW Using In-Container Vitrification - 19239

Thermal treatment is being considered in the UK for the treatment of a wide range of low and intermediate level wastes (LLW and ILW). To support this objective, a range of in-container vitrification trials were conducted to test a wide range of simulated waste streams. The vitrified products have been visually inspected and samples removed for chemical analysis. This paper presents preliminary results of this analysis with some comment on the spatial and chemical heterogeneity of the products. Treatment of a sludge simulant in a steel skip failed to fully melt the metal and produced a glass encapsulated waste-form with clearly visible porosity. Clinoptilolite produced a dense uniform glass but with some chemical heterogeneity. Some operational issues were encountered when treating water-rich sludge, but these were successfully resolved and resulted in a broadly homogenous vitrified product. Melting simulated mixed decommissioning waste produced a chemically homogenous product, despite the heterogenous nature of the waste stream. Uranium and concrete contained in a steel 'top hat' presented significant treatment issues. The top hat failed to fully melt, and this appears to have prevented the waste contained within the top hat from mixing with the glass external to the top hat. As with the steel skip containing sludge, this again left a glass encapsulated, rather than fully vitrified, product. Attempts to melt bulk PVC (an analogue of soft PCM) initially led to significant operational issues with the melt being terminated early and the waste apparently combusting rather than melting. This was resolved by pre-treating the PVC by shredding it and mixing this with the frit to give a larger contact area for melting. This approach produced a homogenous well mixed glass. Overall it appears the properties of the melt product are highly dependent on the physical and chemical properties of the waste stream being treated. (authors)

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

Treatment of Problematic Reactive Metal Wastes Using the GeoMelt{sup R} In-Container Vitrification (ICV{sup TM}) Process - 20326

Decommissioning of sodium-cooled reactors and fast reactor technologies has generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV{sup TM} melts that processed elemental sodium into stable non-reactive glass. INL is currently implementing GeoMelt{sup R} technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate sodium cooled reactor components into waste forms that comply with existing disposition pathways. Reactive metal wastes require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics to comply with land disposal restrictions. GeoMelt{sup R}, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions first demonstrated the effectiveness of the GeoMelt{sup R} ICV{sup TM} process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glass formers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt{sup R} ICV{sup TM} at a higher waste loading relative to previously demonstrated work performed in 2016; the higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. Additionally, follow-on demonstration testing in 2018 and 2019 focused on more complex shapes and other reactive-metals (mocked up Experimental Breeder Reactor II [EBR-II] subassembly, sodium filled heat exchanger, and a can containing sodium potassium alloy) which were all performed at engineering scale. Veolia Nuclear Solutions designed, installed, and commissioned in September 2018, at Perma-Fix Northwest in Richland Washington, a 10-metric ton full-scale GeoMelt unit (GeoMelt{sup R} Richland) for the treatment of reactive metal wastes. As of September 2019, over 900 55-gallon drums containing a total of around 3,500 lb of sodium with low levels of radioactivity have been treated at GeoMelt{sup R} Richland, with resulting glass monoliths disposed at the Nevada National Security Site (NNSS). A full-scale radiological demonstration melt on an actual EBR-II subassembly has also been performed using the full-scale melter in 2019. The GeoMelt{sup R} technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt{sup R} technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt{sup R} can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results generated as a part of the 2018-2019 demonstration program are presented in the paper. (authors)

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

Double-walled containment cell

Double-walled containment cells are described as may be used for storage, transport, and examination of a sample held within the cell by use of optical analysis techniques. A double-walled containment cell can include multiple types of windows that can be located as desired on the containment cells and thereby provide for optical access to a sample for multiple optical analysis techniques. Disclosed containment cells can be sized and designed for use with existing optical analysis systems, e.g., laser ablation, X-ray diffraction, spectral analysis (e.g., Raman spectroscopy, infrared spectroscopy, laser-induced breakdown spectroscopy, etc.), imaging analysis, etc.

Villa-Aleman, Eliel↗
DOE OSTI2019

Closure Device For a Containment Vessel in a Small Gram Quantity Shipping Package

SRNL has been chartered by DOE Environmental Management (DOE-EM) to develop a new Type B shipping package for Small Gram Quantities (SGQ) of radioactive material. Design a containment vessel (CV) closure device that will reduce the operation time and improve the ease-of- use for Operations Personnel. The Containment Vessel (CV) includes a closure lid cap with spring shaft and locking tabs, multiple disc springs, and vessel with integral locking ring. The CV is assembled by aligning the tabs on the lid cap with the openings in the locking ring. Next, the lid cap is pushed down to compress the disc springs. Finally, the lid is rotated 45-degrees to locking location. When released, the tabs are secured against the locking ring. Disc springs (e.g. Belleville washers, coned-disc springs, Belleville spring, etc.) exert axial loads on the mating parts when compressed. The orientation of the disc springs can be configured to change the deflection and axial loads for specific designs. In addition to securing the lid cap to the vessel, the compressed disc springs can be used to apply a load to a containment seal (e.g. o-ring). A prototype of the CV closure was 3D printed to evaluate the locking mechanism design and functionality. A test plan will be developed to evaluate the mechanical properties (i.e. compression forces) and functionality of the CV closure design.

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

Electrification Analysis: Container Ports' Cargo Handling Equipment

This one-page highlight details the key takeaways from a project that utilized NREL's Fleet Research, Energy Data, and Insights (FleetREDI) data analysis pipeline, the Electrification Analysis of Container Ports' Cargo Handling Equipment project. This project created a scalable solution to model energy demand per shipping container moved (kWh/TEU) for an all-electric cargo handling equipment fleet located at a maritime port. The model allows stakeholders to understand energy demand at each electric vehicle (EV) equipment level and is easily scalable to container demand and EV adoption rate projections.

ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2017

Advances in containment methods and plutonium recovery strategies that led to the structural characterization of plutonium(IV) tetrachloride tris-diphenylsulfoxide, PuCl 4 (OSPh 2 ) 3

In an attempt to further advance the understanding of plutonium coordination chemistry, we report a robust method for recycling and obtaining plutonium aqueous stock solutions that can be used as a convenient starting material in plutonium synthesis. This approach was used to prepare and characterize plutonium(IV) tetrachloride tris-diphenylsulfoxide, PuCl 4 (OSPh 2 ) 3 , by single crystal X-ray diffraction. The PuCl 4 (OSPh 2 ) 3 compound represents a rare example of a 7-coordinate plutonium(IV) complex. Structural characterization of PuCl 4 (OSPh 2 ) 3 by X-ray diffraction utilized a new containment method for radioactive crystals. The procedure makes use of epoxy, polyimide loops, and a polyester sheath to provide a robust method for safely containing and easily handling radioactive samples. Lastly, the described procedure is more user friendly than traditional containment methods that employ fragile quartz capillary tubes. Additionally, moving to polyester, instead of quartz, lowers the background scattering from the heavier silicon atoms.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Vortex breakdown in closed containers with polygonal cross sections

The vortex breakdown bubble in the confined flow generated by a rotating lid in closed containers with polygonal cross sections was analysed both experimentally and numerically for the height/radius aspect ratio equal to 2. The stagnation point locations of the breakdown bubble emergence and the corresponding Reynolds number were determined experimentally and in addition computed numerically by STAR-CCM+ CFD software for square, pentagonal, hexagonal, and octagonal cross section configurations. The flow pattern and the velocity were observed and measured by combining the seeding particle visualization and the temporal accuracy of laser Doppler anemometry. The vortex breakdown size and position on the container axis were determined for Reynolds numbers, ranging from 1450 to 2400. The obtained results were compared with the flow structure in the closed container of cubical and cylindrical configurations. It is shown that the measured evolution of steady vortex breakdown is in close agreement with the numerical results.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

Development of sealed sample containers and high resolution micro-tomography

A sample container and a high resolution micro-tomography system have been developed at BL47XU at SPring-8. The container is made of a SiN membrane in a shape of truncated pyramid, which makes it possible to exclude oxygen and moisture in the air. The sample rotation stage for tomography is set downward to keep the sample in the container without any glue. The spatial resolution and field of view are 300 nm and 110 μm using a Fresnel zone plate objective with an outermost zone width of 100 nm at 8 keV, respectively. The scan time is about 20 minutes for 1800 projections. A 3-D image of an asteroid particle was successfully obtained without adhesive and contamination.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

An experimental study on the uptake factor of tungsten oxide particles resulting from an accidentally dropped storage container

A test procedure was developed and verified to measure the airborne concentrations of particles of different sizes (0.5–20 μm) within the vicinity of a dropped container when a significant portion of the tungsten oxide powder (simulating uranium oxide) is ejected from the container. Tests were carried out in a full-scale stainless steel environmental chamber with an interior volume of 24.1 m 3 . Thirty-two drop tests were performed, covering variations in dropping height, room air movement, landing scenario, and lid condition. Assuming a breathing rate of 1.2 m 3 /hr, the uptake factor during the first 10 min was calculated to be between 1.13 × 10 –9 and 1.03 × 10 –7 in reference to the amount loaded; or between 6.44 × 10 –8 and 3.55 × 10 –4 in reference to the amount spilled. Results provide previously unavailable data for estimating the exposure and associated risk to building occupants in the case of an accidental dropping of heavy powder containers. The test data show that for spills larger than 0.004 g, the power-law correlation between the spill uptake factor and the spilled mass (i.e., SUF = 2.5 × 10 –5 × Spill_Mass –0.667 ) established from the test data is smaller and a more accurate estimate than the constant value of 10 –3 assumed in the Department of Energy Nuclear Material Packaging Manual. [Supplementary materials are available for this article. Go to the publisher's online edition of Journal of Occupational and Environmental Hygiene for the following free supplementary resource: an online supplementary table of all cumulative uptake amounts at 10 min for all test data.]

42 ENGINEERING↗
DOE OSTI2020

Intubation Containment System for Improved Protection From Aerosolized Particles During Airway Management

Worldwide efforts to protect front line providers performing endotracheal intubation during the COVID-19 pandemic have led to innovative devices. Authors evaluated the aerosol containment effectiveness of a novel intubation aerosol containment system (IACS) compared with a recently promoted intubation box and no protective barrier. In a simulation center at the authors’ university, the IACS was compared to no protective barrier and an intubation box. Aerosolization was simulated using a commercial fog machine and leakage of aerosolize mist was visually assessed. The IACS appeared to contain the aerosolized mist, while the intubation box allowed for mist to contact the laryngoscopist and contaminate the clinical space through arm port holes and the open caudal end. Both devices protected the laryngoscopist better than no protective barrier. The IACS with integrated sleeves and plastic drape appears to offer superior protection for the laryngoscopist and assistant providers from aerosolized particles.

aerosols↗
DOE OSTI2021

Numerical Simulation of Container Breach and Airborne Release of Solids Due to Mechanical Insults

Throughout U.S. Department of Energy (DOE) complexes, safety engineers employ the five-factor formula to calculate the source term (ST) that includes parameters of airborne release fraction (ARF), respirable fraction (RF) and damage ratio (DR). Limited experimental data on fragmentation of solids, such as ceramic pellets (i.e., PuO2), and container breach due to mechanical insults (i.e., drop and forklift impact), can be supplemented by modeling and simulation using high fidelity computational tools to estimate these parameters. This paper presents the use of Sandia National Laboratories' SIERRA solid mechanics (SM) finite element code to investigate the behavior of the widely utilized waste container (such as 7A Drum) subject to a range of free fall impact and puncture scenarios. The resulting behavior of the container is assessed, and the estimates are presented for bounding DRs from calculated breach areas for the various accident conditions considered. This paper also describes a novel multiscale constitutive model recently implemented in SIERRA/SM that simulates the fracture of brittle materials such as PuO 2 and determines ARF during the fracture process. Furthermore, comparisons are made between model predictions and simple bench-top experiments.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2013

Key differentiators influencing the choice of robust shielded containers

The NDA's Upstream Optioneering project has examined the potential implications of using an alternative type of waste package and its influence on the waste management lifecycle across the NDA estate. Robust Shielded Containers (RSCs) are radioactive waste packages that provide integral radiological shielding, reducing the need for remote handling. The robustness of the container could remove the need to immobilise waste by cement encapsulation. RSCs are routinely used to package ILW for interim storage in Germany and have fairly recently been considered for use in the UK because RSCs have the potential to simplify the waste management lifecycle and enable risk and hazard reduction. In particular, the current baseline (included in Magnox Limited lifetime plans) follows the strategy set out in the Magnox Operating Decommissioning Programme (MODP), in which Type II MOSAIK and Type VI Ductile Cast Iron Containers (DCICs) are used to package (in non-encapsulated form) most Magnox ILW arising prior to 2050. By considering representative types of Intermediate Level Waste (ILW) and exploring potential packaging options for these wastes, this paper identifies the factors that could differentiate between cases in which RSCs would, or would not, be an appropriate option. The potential role of RSCs across the waste management lifecycle is examined, from retrieval of waste through to emplacement at a Geological Disposal Facility (GDF), including consideration of other potential uses of RSCs, such as temporary storage of raw wastes for which appropriate treatment and conditioning measures have yet to be developed. (authors)

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

Leachate Treatment for Radioactive Waste Containment Mound: Pilot Test and Design - 19098

Canadian Nuclear Laboratories plans to construct an engineered containment mound (ECM) at its Chalk River Laboratories site in Ontario for safe management of radioactive and mixed waste at the facility. The ECM is designed for containment of 1,000,000 cubic meters of waste, and will be operated over a period of 50 years. Contact storm water, leachate, and other wastewater will be generated during operation of the ECM, and must be treated prior to discharge to the environment. The wastewater is projected to contain low concentrations of radionuclides and metals, as well as other cations and anions, and low concentrations of organic chemicals. Through comparison with surface water discharge criteria, radionuclides cesium-137 and strontium-90, and metals including aluminum, barium, copper, iron, lead, and manganese were projected to be present in the wastewater at concentrations that may exceed surface water discharge criteria, and were therefore targeted for treatment. Several technologies, including chemical precipitation, membrane filtration, ion exchange, and reverse osmosis were reviewed and tested on a laboratory scale using simulated wastewater prepared to replicate the concentrations of both target and non-target constituents. Based upon the results of the laboratory-scale tests, a six-week pilot test was conducted to simulate the performance of the technologies under continuous-flow conditions and determine full-scale design criteria. This paper describes the methodology and results of the six-week pilot test, and how the test results were used to inform the full-scale wastewater treatment system design. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Corrosion of Containment Alloys in Molten Salt Reactors and the Prospect of Online Monitoring

The aim of this review is to communicate some essential knowledge of the underlying mechanism of the corrosion of structural containment alloys during molten salt reactor operation in the context of prospective online monitoring in future MSR installations. The formation of metal halide species and the progression of their concentration in the molten salt do reflect containment corrosion, tracing the depletion of alloying metals at the alloy salt interface will assure safe conditions during reactor operation. Even though the progress of alloying metal halides concentrations in the molten salt do strongly understate actual corrosion rates, their prospective 1 st order kinetics followed by near-linearly increase is attributed to homogeneous matrix corrosion. The service life of the structural containment alloy is derived from homogeneous matrix corrosion and near-surface void formation but less so from intergranular cracking (IGC) and pitting corrosion. Online monitoring of corrosion species is of particular interest for molten chloride systems since besides the expected formation of chromium chloride species CrCl 2 and CrCl 3 , other metal chloride species such as FeCl 2 , FeCl 3 , MoCl 2 , MnCl 2 and NiCl 2 will form, depending on the selected structural alloy. The metal chloride concentrations should follow, after an incubation period of about 10,000 hours, a linear projection with a positive slope and a steady increase of <1 ppm per day. During the incubation period metal concentration show 1 st order kinetics and increasing linearly with time. Ideally, a linear increase reflects homogeneous matrix corrosion, while a sharp increase in the metal chloride concentration could set a warning flag for potential material failure within the projected service life, e.g. as result of intergranular cracking or pitting corrosion. Continuous monitoring of metal chloride concentrations can therefore provide direct information about the mechanism of the ongoing corrosion scenario and offer valuable information for a timely warning of prospective material failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗