Shirley Basin South, Wyoming, Disposal Site
This fact sheet provides information about the Shirley Basin South, Wyoming, Disposal Site
Source record↗
SEARCH · Engineering Papers
Results for “DISPOSAL”
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.
Slick Rock, Colorado, Processing Sites and Disposal Sites
This fact sheet provides information about the Slick Rock, Colorado, Processing Sites and Disposal Sites
Source record↗
Spook, Wyoming, Disposal Site
This fact sheet provides information about the Spook, Wyoming, Disposal Site
Source record↗
Tuba City, Arizona, Disposal Site
This fact sheet provides information about the Tuba City, Arizona, Disposal Site
Source record↗
Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS)
This will be licensed via RSICC. A new, integrated data and analysis system has been designed to simplify and automate the performance of accurate and efficient evaluations for characterizing the input to the overall nuclear waste management system -UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database within UNF-ST&DARDS provides a standard means by which UNF-ST&DARDS can succinctly store and retrieve modeling and simulation (M&S) parameters for specific spent nuclear fuel analysis. A library of various analysis model templates provides the ability to communicate the various set of M&S parameters to the most appropriate M&S application. Interactive visualization capabilities facilitate data analysis and results interpretation. UNF-ST&DARDS current analysis capabilities include (1) assembly-specific depletion and decay, (2) and spent nuclear fuel cask-specific criticality and shielding. Currently, UNF-ST&DARDS uses SCALE nuclear analysis code system for performing nuclear analysis.
Banerjee, Kaushik↗
Packaging design for storage, transportation, and disposal of disused radiological sources
A system for encapsulating and storing disused radiological sources in sealed capsules is provided, the system having a basket to removably position capsules relative to each other, the capsules containing the radiological sources; a containment vessel for receiving the basket; and a cask reversibly encapsulating or otherwise housing the containment vessel. Also provided is a method for packaging, transporting and storing disused radiological sources, the method having the steps of transporting sealed capsules containing radiological sources from water pools to baskets; placing the basket in a containment vessel and sealing the vessel with helium backfill; placing the vessel in a cask and reversibly capping the cask; surrounding the cask with personnel a shield and crumple zones to create a construct; and transporting and storing the construct until its final disposal at a geological repository or a deep borehole, all without repackaging of the disused radiological sealed capsules.
Liu, Yung Y.↗
Hygroscopic cooling tower for waste water disposal
In various embodiments, the present invention relates to heat dissipation systems including a hygroscopic working fluid integrating waste water as makeup water. The present invention also relates to methods of using the same. The present invention also relates to hygroscopic cooling systems adapted to dispose of waste water by combining the waste water with a hygroscopic working fluid, precipitating impurities and evaporating the remaining water.
Martin, Christopher Lee↗
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↗
Experimental Investigation of Time-Dependent Deformation of Fluid Disposal Reservoir Caprocks
For approximately the last two decades low-permeability formations have become a prominent research area because these are the target of unconventional oil shale exploration, and also the primary trapping mechanism for CO2 sequestration and other such subsurface storage. The objective of this research is focused on the sealing ability and induced seismic risk of such tight formations. Disposal of fluid into the subsurface alters the stress state in both the target reservoir and the surrounding formations from their initial in situ state. Therefore, caprocks must be able to deform and accommodate the altered stress and strain state while maintaining the integrity of the seal. The deformation induced in caprock formations often results in a decrease in the normal stress on faults and fractures, which may induce shear slip and therefore create the potential for induced seismicity and leakage. Deformations induced by fluid injection are generally not instantaneous, but typically occur gradually as a pressure front advances through the reservoir and caprock formations. As such, an injection program design can influence both the rate and magnitude of deformation and associated changes in stress. Many common caprock formation types, such as shales and evaporates, are prone to undergo time-dependent deformation such as viscoelastic and/or viscoplastic response. Both the magnitude of the time-dependent deformation and the associated characteristic time scale are important considerations for selecting or operating an injection site. This paper describes tests intended to characterize these effects using caprock samples from a variety of current or potential storage sites. Triaxial and load relaxation test data are presented and used to quantify the time-dependent deformation behavior of rocks with different textures and compositions.
Edelman, Eric↗
Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) 4.x
A new, integrated data and analysis system has been designed to simplify and automate the performance of accurate and efficient evaluations for characterizing the input to the overall waste management system-UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database has been assembled to provide a standard means by which UNF-ST&DARDS can succinctly store and retrieve M&S parameters for specific SNF analysis. Users can import SNF data into this database for user-specific analysis. A library of various analysis model templates provides the ability to communicat the various set of M&S parameters to the most appropriate M&S application. A process manager facilitates performance of actual, as-loaded, assembly-specific and cask-specific evaluations. Interactive visualization capabilities facilitate data analysis and results interpretation.
Banerjee, Kaushik↗
Hygroscopic cooling tower for waste water disposal
In various embodiments, the present invention relates to heat dissipation systems including a hygroscopic working fluid integrating waste water as makeup water. The present invention also relates to methods of using the same. The present invention also relates to hygroscopic cooling systems adapted to dispose of waste water by combining the waste water with a hygroscopic working fluid, precipitating impurities and evaporating the remaining water.
Martin, Christopher Lee↗
Superfund Policy Statements and Guidance Regarding Disposition of Radioactive Waste in Non-NRC Licensed Disposal Facilities - 13407
This talk will discuss EPA congressional testimony and follow-up letters, as well as letters to other stakeholders on EPA's perspectives on the disposition of radioactive waste outside of the NRC licensed disposal facility system. This will also look at Superfund's historical practices, and emerging trends in the NRC and agreement states on waste disposition. (author)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Modular, disposable 3D printed microfluidic membrane system for separation and purification
The present disclosure relates to a membrane extraction apparatus for extracting a component from a first liquid. The apparatus may incorporate a housing comprised of first and second mating housing halves, with each housing half having an open faced channel formed therein such that the channels at least partially overlay one another when the two housing halves are secured together. A membrane filter is disposed between the two housing halves to overlay the open faced channels. The membrane filter extracts the component from the first liquid and transfers the component into the second liquid as the first and second liquids flow through the first and second housing halves.
Servis, Anna G.↗
Performance assessment for the geological disposal of Deep Burn spent fuel using TTBX
The behavior of Deep Burn Modular High Temperature Reactor Spent Fuel (DBSF) is investigated in the Yucca Mountain geological repository (YMR) with respect to the annual dose (Sv/yr) delivered to the Reasonably Maximally Exposed Individual (RMEI) from the transport of radionuclides released from the graphite waste matrix. Transport calculations are performed with a novel computer code, TTBX which is capable of modeling transport pathways that pass through heterogeneous geological formations. TTBX is a multi-region extension of the existing single region TTB transport code. Overall the peak annual dose received by the RMEI is seen to be four orders of magnitude lower than the regulatory threshold for exposure, even under pessimistic scenarios. A number of factors contribute to the favorable performance of DBSF. A reduction of one order of magnitude in the peak annual dose received by the RMEI is observed for every order of magnitude increase in the waste matrix lifetime, highlighting the importance of the waste matrix durability and suggesting graphite's utility as a potential waste matrix for the disposal of high-level waste. Furthermore, we see that by incorporating a higher fidelity far-field model the peak annual dose calculated to be received by the RMEI is reduced by two orders of magnitude. By accounting for the heterogeneities of the far field we have simultaneously removed unnecessary conservatisms and improved the fidelity of the transport model. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Development of Wireless Data Transmission System for the Monitoring in Geological Disposal of Radioactive Waste - 12063
The authors have been developing a wireless data transmission system to monitor the performance of a geological disposal system for radioactive waste. The system's concepts, advantages, and a recent development focused on reducing transmitter size to suit narrow spaces such as bentonite buffers and boreholes. A wireless transmitter with a built-in temperature sensor and a connector for external sensors has been developed, measuring 130 mm in length and 50 mm in diameter. The capability of the transmitter was confirmed by transmission tests on the ground and in a bentonite block. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Modeling the Hydrogeochemical Transport of Radionuclides through Engineered Barriers System in the Proposed LLW Disposal Site of Taiwan - 12082
A proposed site for final disposal of low-level radioactive waste located in Daren Township of Taitung County along the southeastern coast has been on the selected list in Taiwan. The geology of the Daren site consists of argillite and meta-sedimentary rocks. A mined cavern design with a tunnel system of 500 m below the surface is proposed. Concrete is used as the main confinement material for the engineered barrier. To investigate the hydrogeochemical transport of radionuclides through engineered barriers system, HYDROGEOCHEM5.0 model was applied to simulate the complex chemical interactions among radionuclides, the cement minerals of the concrete, groundwater flow, and transport in the proposed site. The simulation results showed that the engineered barriers system with the side ditch efficiently drained the ground water and lowered the concentration of the concrete degradation induced species (e.g., hydrogen ion, sulfate, and chloride). The velocity of groundwater observed at side ditch gradually decreased with time due to the fouling of pore space by the mineral formation of ettringite and thaumasite. The short half-life of Co-60, Sr-90 and Cs-137 significantly reduced the concentrations, whereas the long half-life of I-129(1.57x10{sup 7} years) and Am-241(432 years) remain stable concentrations at the interface of waste canister and concrete barrier after 300 years. The mineral saturation index (SI) was much less than zero due to the low aqueous concentration of radionuclide, so that the precipitation formation of Co-60, Sr-90, I-129, Cs-137 and Am-241 related minerals were not found. The effect of adsorption/desorption (i.e., surface complexation model) could be a crucial geochemical mechanism for the modeling of liquid-solid phase behavior of radionuclide in geochemically dynamic environments. Moreover, the development of advanced numerical models that are coupled with hydrogeochemical transport and dose assessment of radionuclide is required in the future. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
On the influence of the power plant operational history on the inventory and the uncertainties of radionuclides relevant for the final disposal of PWR spent fuel - 15149
A mandatory requisite for official granting safe final disposal approval of SNF is the comprehensive specification and declaration of the nuclear inventory in SNF by the NPP-operators. The whole radiochemical characterization of each spent fuel assembly hides many difficulties such as very high radiation burden, destruction method, the inhomogeneous burn-up of fuel assembly, very high cost. Thus, non-destructive characterization is usually applied on industrial scale. In this case only several nuclides can be measured, and the rest is calculated using simulations and typical irradiation history. Like any other German rad-waste acceptance requirements all these declarations will have to be checked and verified by an independent inspector. In this verification process not only values of radionuclide (RN) activities are important, but also their uncertainties. The simple simulation using burn-up calculations based on typical and generic reactor operational parameters do not answer the question about possible uncertainties observed in real reactor operations. At the same time, the details of irradiation history are often unknown, which complicates the assessment of declared RN inventories. Therefore, the development of dedicated generic proof tools, which can be applied to different types of waste, often of unknown origin and history is worth a serious consideration. Here, we have compiled a set of burn-up calculations considering the operational history of 339 published or anonymized real PWR fuel assemblies. This gives us the possibility to calculate the realistic variation or spectrum of RN inventories. These histories were used as a basis for 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's), i.e. initial enrichment (IE), specific power (SP), irradiation time (IT), downtime (DT), boric acid concentration (BA), moderator and fuel density (MD and FD) and temperatures (MT and MT) used as input parameters for the simulation models. Then, burn-up calculation codes provide the anticipated results for the entire nuclear inventory such as characteristic activity values and the corresponding minimum and maximum limits as a function of the fuel burn-up. These results are being validated with the experimental data from various publications and the online database - SFCOMPO and subsequently transferred and summarized into our own database. Ultimately, this database will provide the basis of proof tools for the operator independent verification of the declared nuclear inventories for the quality and safety assessment of a radioactive waste compound and bounding values for certain RN as a function of burn-up. The RN inventory can hence be verified without taking credit of the whole of the operational history of the SNF. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
A study of transfer of UNF from non-disposable canisters - 15388
A study was performed by AREVA and URS for the Department of Energy (DOE) [1] to establish optimal methods for the opening of dry storage canisters (DSC) and the subsequent repackaging of used nuclear fuel (UNF) from the DSCs into generic standard transportation, aging, and disposal canisters (STADs) in wet or dry environments at reactor, consolidated storage, or repository sites. Eight dry and wet cutting technologies were evaluated: (1) skiving; (2) milling; (3) arc gouging/plasma torch; (4) cutting wheel; (5) diamond cable cutting; (6) laser cutting; (7) sawing; and (8) electrical discharge machining. The evaluated UNF repackaging technologies options included: (1) wet transfer in a pool; (2) dry transfer in a dry transfer system (DTS); (3) dry transfer in a fixed hot cell (FHC); (4) dry transfer in a mobile hot cell (MHC), and (5) a hybrid approach utilizing wet transfer in a pool and dry transfer in a FHC. A systems engineering evaluation was utilized for ranking the eight viable cutting options and the five different repackaging approaches for each of the sites. The systems engineering evaluation took into account six criteria (i.e., life cycle costs, worker dose, operational efficiency, technical maturity, ease of licensing, and waste generation) and performed a qualitative pairwise ranking of their relative importance to one another. Then quantitative or qualitative factors associated with each criterion (e.g., cost in millions of dollars, relative ease of licensing, etc.) for each option at each location (reactor, consolidated storage facility [CSF], repository) were developed based on historical experience, existing design studies, current 'state-of-the art', and/or engineering judgment. Using this input, the systems engineering evaluation identified dry skiving as the preferred cutting technology, with the wheel cutting technology in either a dry or wet environment identified as a promising technology that, with some RD and D, could mature to a better option (depending also on the disposition of the used DSC). For a reactor and a CSF/repository site, the systems engineering evaluation ranked the repackaging option using the skiving dry cutting technique and wet transfer of UNF in a pool as the highest option. However for a CSF and a repository site, the team recommended the hybrid option using the skiving dry cutting technique and both dry and wet transfer of UNF in a pool and FHC as a potentially better option (providing specific benefits such as higher throughput and more flexibility in operation). (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗