Engineering Papers⌕ Search

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.

At least 847 records · Page 47

Review of corrosion interactions between different materials relevant to disposal of high-level nuclear waste

Abstract This review covers the corrosion interactions between different materials that are relevant to the disposal of high-level nuclear waste, in particular the waste forms and containers. The materials of interest are borosilicate glass, crystalline ceramics, metal alloys, and any corrosion products that might form. The available data show that these interactions depend on the structure, chemistry, thermodynamic history, and proximity of the materials in contact, as well as the environmental attributes, such as temperature, solution chemistry, and radiation. Several key mechanisms that govern these interactions are highlighted. Scientific gaps and open questions are summarized and discussed.

Materials Science↗

Retrieval, Disposal, and Disposition of Legacy 252 Cf Sealed Sources

In this article, we describe work performed on behalf of the US Department of Energy (DOE) to disposition 467 government-owned, sealed californium-252 ( 252 Cf) sources located at 52 academic and government research sites. These sources are the legacy of the DOE 252 Cf Loan/Lease Program, which operated from the late 1960s until 2008, when the loan program closed abruptly. Although the DOE took decades to fabricate and distribute the sources, best safety and nuclear accountability practices demanded retrieval or disposition of all loaned sources as soon as achievable after the loan program was closed. An adaptable disposition approach was developed, and disposition of the sources was pursued. The approach involved a combination of source retrievals, disposal actions, and source ownership transfers. A new shielded shipping container was designed and deployed to aid in source retrievals. In a couple cases, incomplete source fabrication and shipping records prevented shipment, and those sources were left indefinitely with the customer in a safe configuration. Lastly, all the sources in the inventory were safely dispositioned in some way, and lessons learned from this effort are provided.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Interaction of Sr-90 with site candidate soil for demonstration disposal facility at Serpong

Interaction of radiostrontium (Sr-90) with site candidate soil for demonstration disposal facility to be constructed in the near future at Serpong has been done. This activity is to anticipate the interim storage facility at Serpong nuclear area becomes full off condition, and show to the public how radioactive waste can be well managed with the existing technology. To ensure that the location is save, a reliability study of site candidate soil becomes very importance to be conducted through some experiments consisted some affected parameters such as contact time, effect of ionic strength, and effect of Sr{sup +} ion in solution. Radiostrontium was used as a tracer on the experiments and has role as radionuclide reference in low-level radioactive waste due to its long half-live and it's easy to associate with organism in nature. So, interaction of radiostrontium and soil samples from site becomes important to be studied. Experiment was performed in batch method, and soil sample-solution containing radionuclide was mixed in a 20 ml of PE vial. Ratio of solid: liquid was 10{sup −2} g/ml. Objective of the experiment is to collect the specific characteristics data of radionuclide sorption onto soil from site candidate. Distribution coefficient value was used as indicator where the amount of initial and final activities of radiostrontium in solution was compared. Result showed that equilibrium condition was reached after contact time 10 days with Kd values ranged from 1600-2350 ml/g. Increased in ionic strength in solution made decreased of Kd value into soil sample due to competition of background salt and radiostrontium into soil samples, and increased in Sr ion in solution caused decreased of Kd value in soil sample due to limitation of sorption capacity in soil samples. Fast condition in saturated of metal ion into soil samples was reached due to a simple reaction was occurred.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Zika virus detection using antibody-immobilized disposable cover glass and AlGaN/GaN high electron mobility transistors

Zika virus detection was demonstrated using antibody-functionalized cover glasses externally connected to the gate electrode of an AlGaN/GaN high electron mobility transistor (HEMT). A pulsed bias voltage of 0.5 V was applied to an electrode on the region of the cover glass region functionalized with antibody, and the resulting changes of drain current of the HEMT were employed to determine the presence of Zika virus antigen concentration ranging from 0.1 to 100 ng/ml. The dynamic and static drain current changes as a function of Zika virus concentration were modeled with a spring-like elastic relaxation model and the Langmuir extension model, respectively. Excellent fits to the data were found with relaxation time constants of antibody and antigen molecules in the range of 11 μs and 0.66–24.4 μs, respectively, for the concentration range investigated. The ratio of antibody bound with antigen to the total available antibody on the functionalized contact window was in the range of 0.013–0.84 for the Zika antigen concentration range of 0.1–100 ng/ml. Since the HEMT is not exposed to the bio-solution, it can be used repeatedly. Here, the functionalized glass is the only disposable part in the detection system, showing the potential of this approach for hand-held, low cost sensor packages for point-of-care applications.

59 BASIC BIOLOGICAL SCIENCES↗

Introduction to special section: Wastewater disposal and CO2 transport in the subsurface

This special section presents interpretation, reservoir characterization, modeling, monitoring, and geoengineering of the multidisciplinary aspects of wastewater disposal and CO2 transport in the subsurface. In striving for zero net greenhouse gas emissions, the oil and gas industry needs transformative technology to efficiently produce and use fossil fuels in a safe and responsible manner.

Brown, Christopher F.↗

Summary of Disposable Debris Shields (DDS) Analysis for Development of Solid Debris Collection at NIF

Collection of solid debris from the National Ignition Facility (NIF) is being developed both as a diagnostic tool and as a means for measuring nuclear reaction cross sections relevant to the Stockpile Stewardship Program and nuclear astrophysics. The concept is straightforward; following a NIF shot, the debris that is produced as a result of the capsule and hohlraum explosion would be collected and subsequently extracted from the chamber. The number of nuclear activations that occurred in the capsule would then be measured through a combination of radiation detection and radiochemical processing followed by mass spectrometry. Development of the catcher is challenging due to the complex environment of the NIF target chamber. The collector surface is first exposed to a large photon flux, followed by the debris wind that is produced. The material used in the catcher must be mechanically strong in order to withstand the large amount of energy it is exposed to, as well as be chemically compatible with the form and composition of the debris. In addition, the location of the catcher is equally important. If it is positioned too close to the center of the target chamber, it will be significantly ablated, which could interfere with the ability of the debris to reach the surface and stick. If it is too far away, the fraction of the debris cloud collected will be too small to result in a statistically significant measurement. Material, geometric configuration, and location must all be tested in order to design the optimal debris collection system for NIF. One of the first ideas regarding solid debris collection at NIF was to use the disposable debris shields (DDS), which are fielded over the final optics assemblies (FOA) 7 m away from the center of the target chamber. The DDS are meant to be replaced after a certain number of shots, and if the shields could be subsequently analyzed after removal, it would serve as a mechanism for fielding a relatively large collection area through the use of a part meant to be replaced regularly. The solid angle covered by one of the shields is roughly 10 -4 of 4π. If several shields were analyzed at once, it would increase the solid angle of the collection area accordingly. The glass shields consist of ammonia hardened silica with a sol gel coating and kapton tape around the edge. The square sheets are 14-inch on each side. The original shields were 1 mm thick, but it was determined that a thicker shield (3.3 mm) was more effective in preventing debris from reaching the FOA. The Solid Radchem group received two sets of DDS as part of our evaluation of the potential use of the DDS as solid debris collectors. The first set consisted of two 3.3 mm shields, one each from the top and bottom of the chamber (the '3mm set'). The second set consisted of four 1mm shields, one from the top of the chamber and the other three from the bottom (the 'IFSA set'). For each set, the shields were cut into smaller subsamples, which were then imaged using scanning electron microscopy (SEM) followed by chemical leaching and mass spectrometry. The purpose was to evaluate both the quantity and identity of the debris that was present on the DDS surfaces, and to determine if any of the capsule debris was reaching the chamber walls. In addition, potential enhancement due to gravity in the chamber was evaluated by directly comparing shields fielded in the top and bottom of the chamber. Based on the results, the use of the DDS as debris collectors would be evaluated. The results from both sets were presented to the DDS Working Group. The slides are attached to this document. The 3mm set results are presented first, followed by the results from the IFSA set. In both cases it was determined that a small fraction of the overall debris field was collected on the DDS. This means that the debris that is formed during a NIF shot is condensing out of the plasma and depositing on surfaces closer to the target chamber center, or else it is simply falling to the bottom of the chamber. In either case, it was determined that using the DDS, or fielding a debris collector at the chamber wall, was not feasible for solid debris collection at NIF due to the small amount of debris that had been collected. In addition, since the glass shields suffered quite a bit of damage from particles impacting the surface, glass was ruled out as a collection medium.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Portsmouth On-Site Disposal Cell (OSDC) High Density Polyethylene (HDPE) Geomembrane Longevity

It is anticipated that high density polyethylene (HDPE) geomembranes will be utilized within the liner and closure cap of the proposed On-Site Disposal Cell (OSDC) at the Portsmouth Gaseous Diffusion Plant. The likely longevity (i.e. service life) of HDPE geomembranes in OSDC service is evaluated within the following sections of this report: (1) Section 2.0 provides an overview of HDPE geomembranes, (2) Section 3.0 outlines potential HDPE geomembranes degradation mechanisms, (3) Section 4.0 evaluates the applicability of HDPE geomembrane degradation mechanisms to the Portsmouth OSDC, (4) Section 5.0 provides a discussion of the current state of knowledge relative to the longevity (service life) of HDPE geomembranes, including the relation of this knowledge to the Portsmouth OSDC, and (5) Section 6.0 provides summary and conclusions relative to the anticipated service life of HDPE geomembranes in OSDC service. Based upon this evaluation it is anticipated that the service life of HDPE geomembranes in OSDC service would be significantly greater than the 200 year service life assumed for the OSDC closure cap and liner HDPE geomembranes. That is, a 200 year OSDC HDPE geomembrane service life is considered a conservative assumption.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Safety Design Strategy for the Remote Handled Low-Level Waste Disposal Project

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3A, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3A and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Remote-Handled Low-Level Waste Disposal Project.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Site characterization methodology for deep borehole disposal

Deep Borehole Disposal (DBD) for radioactive waste has many clear advantages over mined repositories: 1) the possibility of incremental construction and loading at multiple locations, 2) the enhanced natural barriers in the deep continental crystalline basement, and 3) reduced site characterization. This report identifies characterization methods relevant to DBD of spent nuclear fuel or vitrified high-level waste (HLW). A systematic process based on performance assessment methodology and in particular an analysis of features, events, and processes (FEPs) is used to focus the selection of characterization methods.

Vaughn, Palmer↗

Deep borehole disposal of nuclear waste summary (Final Report)

Research and development activities carried out at Sandia National Laboratories from 2009-2012 have built the technical baseline for perform ance of a deep bor ehole for disposal of nuclear waste. Early work established the coupled thermo-hydrological-mechanical-chemical forces likely to control radionuclide m ovement from a deep borehole. This report emphasizes more recent borehole science and engineering activities including: 1. Establishing a reference borehole design; 2. Developing the design for borehole seals; and 3. Identifying the nature and extent of site characterization needed for deep borehole siting.

Brady, Patrick Vane↗

RH-LLW Disposal Facility Project CD-2/3 to Design/Build Proposal Reconciliation Report

A reconciliation plan was developed and implemented to address potential gaps and responses to gaps between the design/build vendor proposals and the Critical Decision-2/3 approval request package for the Remote-Handled Low Level Waste Disposal Facility Project. The plan and results of the plan implementation included development of a reconciliation team comprised of subject matter experts from Battelle Energy Alliance and the Department of Energy Idaho Operations Office, identification of reconciliation questions, reconciliation by the team, identification of unresolved/remaining issues, and identification of follow-up actions and subsequent approvals of responses.

99 GENERAL AND MISCELLANEOUS↗

Towards a defensible safety case for deep geologic disposal of DOE HLW and DOE SNF in bedded salt

The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. A safety case also provides the necessary structure for organizing and synthesizing existing knowledge in order to help DOE prioritize its future research and development (R&D) activities. We conclude that a defensible initial safety case for potential licensing could be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), work on other repository development programs, and the work published through international efforts in salt repository programs such as in Germany.

Mackinnon, Robert James↗

Considerations of human intrusion in U.S. programs for deep geologic disposal of radioactive waste

Regulations in the United States that govern the permanent disposal of spent nuclear fuel and high-level radioactive waste in deep geologic repositories require the explicit consideration of hypothetical future human intrusions that disrupt the waste. Specific regulatory requirements regarding the consideration of human intrusion differ in the two sets of regulations currently in effect in the United States; one defined by the Environmental Protection Agency's 40 Code of Federal Regulations part 197, applied only to the formerly proposed geologic repository at Yucca Mountain, Nevada, and the other defined by the Environmental Protection Agency s 40 Code of Federal Regulations part 191, applied to the Waste Isolation Pilot Plant in New Mexico and potentially applicable to any repository for spent nuclear fuel and high-level radioactive waste in the United States other than the proposed repository at Yucca Mountain. This report reviews the regulatory requirements relevant to human intrusion and the approaches taken by the Department of Energy to demonstrating compliance with those requirements.

Swift, Peter N.↗

Recommended Method To Account For Daughter Ingrowth For The Portsmouth On-Site Waste Disposal Facility Performance Assessment Modeling

A 3-D STOMP model has been developed for the Portsmouth On-Site Waste Disposal Facility (OSWDF) at Site D as outlined in Appendix K of FBP 2013. This model projects the flow and transport of the following radionuclides to various points of assessments: Tc-99, U-234, U-235, U-236, U-238, Am-241, Np-237, Pu-238, Pu-239, Pu-240, Th-228, and Th-230. The model includes the radioactive decay of these parents, but does not include the associated daughter ingrowth because the STOMP model does not have the capability to model daughter ingrowth. The Savannah River National Laboratory (SRNL) provides herein a recommended method to account for daughter ingrowth in association with the Portsmouth OSWDF Performance Assessment (PA) modeling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

All-Pathways Dose Analysis for the Portsmouth On-Site Waste Disposal Facility (OSWDF)

A Portsmouth On-Site Waste Disposal Facility (OSWDF) All-Pathways analysis has been conducted that considers the radiological impacts to a resident farmer. It is assumed that the resident farmer utilizes a farm pond contaminated by the OSWDF to irrigate a garden and pasture and water livestock from which food for the resident farmer is obtained, and that the farmer utilizes groundwater from the Berea sandstone aquifer for domestic purposes (i.e. drinking water and showering). As described by FBP 2014b the Hydrologic Evaluation of Landfill Performance (HELP) model (Schroeder et al. 1994) and the Surface Transport Over Multiple Phases (STOMP) model (White and Oostrom 2000, 2006) were used to model the flow and transport from the OSWDF to the Points of Assessment (POAs) associated with the 680-ft elevation sandstone layer (680 SSL) and the Berea sandstone aquifer. From this modeling the activity concentrations radionuclides were projected over time at the POAs. The activity concentrations were utilized as input to a GoldSimTM (GTG 2010) dose model, described herein, in order to project the dose to a resident farmer over time. A base case and five sensitivity cases were analyzed. The sensitivity cases included an evaluation of the impacts of using a conservative inventory, an uncased well to the Berea sandstone aquifer, a low waste zone uranium distribution coefficient (Kd), different transfer factors, and reference person exposure parameters (i.e. at 95 percentile). The maximum base case dose within the 1,000 year assessment period was projected to be 1.5E-14 mrem/yr, and the maximum base case dose at any time less than 10,000 years was projected to be 0.002 mrem/yr. The maximum projected dose of any sensitivity case was approximately 2.6 mrem/yr associated with the use of an uncased well to the Berea sandstone aquifer. This sensitivity case is considered very unlikely because it assumes leakage from the location of greatest concentration in the 680 SSL in to the Berea sandstone aquiver over time and does not conform to standard private water well construction practices. The bottom-line is that all predicted doses from the base case and five sensitivity cases fall well below the DOE all-pathways 25 mrem/yr Performance Objective.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Sensitivity Analysis for Saltstone Disposal Unit Column Degradation Analyses

PORFLOW related analyses supporting a Sensitivity Analysis for Saltstone Disposal Unit (SDU) column degradation were performed. Previous analyses, Flach and Taylor 2014, used a model in which the SDU columns degraded in a piecewise manner from the top and bottom simultaneously. The current analyses employs a model in which all pieces of the column degrade at the same time. Information was extracted from the analyses which may be useful in determining the distribution of Tc-99 in the various SDUs throughout time and in determining flow balances for the SDUs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229 Th, 230 Th, 232 U, 233 U, and 234 U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Special Analysis for the Disposal of the Neutron Products Incorporated Sealed Source Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

The purpose of this special analysis (SA) is to determine if the Neutron Products Incorporated (NPI) Sealed Sources waste stream (DRTK000000056, Revision 0) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The NPI Sealed Sources waste stream consists of 850 60Co sealed sources (Duratek [DRTK] 2013). The NPI Sealed Sources waste stream requires a special analysis (SA) because the waste stream 60Co activity concentration exceeds the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗