Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Waste Repository”

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
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Radiation Induced Corrosion of Copper in Deep Geological Repositories - 19493

This paper presents an overview of the Cu corrosion research being conducted by the Radiation Assisted Materials Performance Science (RAMPS) group at Western University, in collaboration with the Canadian Nuclear Waste Management Organization (NWMO) and Canada Natural Science and Engineering Research Council (NSERC). This paper presents an overview of our research in three areas to develop a high-fidelity copper corrosion model. Copper is an important candidate material for corrosion protection of used nuclear fuel containers (UFC). The UFC and its environment will be exposed to a continuous flux of γ-radiation. The three areas are chemical kinetics of γ-radiolysis of humid air and groundwater, the effects of solution parameters on Cu corrosion dynamics without γ-radiation and the effects of γ-radiation on Cu corrosion dynamics. These studies show that because of the strong feedback loops between processes occurring during copper corrosion, the overall corrosion mechanism may not follow linear dynamics. Presented here are our key findings, their implications for corrosion science in general, and the potential implications of these results for predictions of the adequacy of a UFC copper corrosion allowance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

System-level logistics modeling of DPC direct disposal - 15511

The U.S. Department of Energy (DOE), Office of Used Nuclear Fuel Disposition, has been conducting a study of the technical feasibility of direct disposal of existing dual-purpose canisters (DPCs) loaded with commercial spent nuclear fuel (SNF). The system-level logistical modeling described here is a part of this study with the primary goals to project the inventory of DPCs in the future and to understand the timing of DPC disposal constrained by thermal power limits for disposal and other key parameters of the waste management system. The logistics simulations also allow for estimating fuel age at emplacement, an important consideration if storage time is limited. This analysis evaluated the effects on DPC direct disposal from uncertainty in repository emplacement power limits, the repository start date, and whether or when the nuclear industry could transition to loading SNF into the smaller capacity multi-purpose canisters (MPCs). MPCs implemented at a future date would likely contain the youngest fuel with highest burnup, but smaller MPCs would cool to the emplacement power limits relatively quickly. The results of this analysis demonstrate that DPC direct disposal could be complete by about 2070 to 2170 depending on the repository emplacement power limit. These completion dates can be expedited by approximately 10 to 40 years in some scenarios through the transition to MPCs. The greatest average SNF age at emplacement is 88 years in the scenarios with the smallest repository thermal power limit in which no transition to MPCs is considered. Transition to MPCs could decrease the average age at emplacement, to as low as 42 years in some scenarios. (authors)

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

Modelling uranium dioxide corrosion under repository conditions: A pore-scale study of the chemical and thermal processes

This paper investigates the corrosion behaviors of spent uranium dioxide (UO 2 ) fuel when placed in geological repositories. We performed pore-scale reactive transport simulations of the UO 2 corrosion process in a defective fuel rod with different orientations of breach or fracture on its clad. It is found that the corrosion rate has strong dependency on the breach orientation. The highest corrosion rate of UO 2 is calculated when the angle between the groundwater flow direction and the clad fracture reaches 180°. UO 2 corrosion simulations with different flow rates are compared. The results show the higher flow rate accelerates the corrosion of UO 2 fuel. The effect of pH on the corrosion process is also determined. It is indicated that higher pH can limit the corrosion of UO 2 by reducing the reaction rate. The dependence of reaction rate and time on reactive surface area is explored. Spent fuel with lower surface area demonstrates longer lifetime under corrosion conditions. A numerical model coupling thermal conduction and chemical reactions is developed to assess the impact of temperature on the process of UO 2 corrosion. The results show that higher temperature leads to larger corrosion rates for UO 2 . Here, the predicted reaction rates are higher in comparison with the isothermal results. Hence, our results can help improve the fundamental understanding of UO 2 corrosion in geological repositories for long-term storage of spent nuclear fuels and provide guidance for the safe operations and selection of appropriate repositories.

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

Storage, transportation and disposal system for used nuclear fuel assemblies

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

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

Storage, transportation and disposal system for used nuclear fuel assemblies

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

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

Potential Cost Implications of an Interim Storage Facility for Commercial SNF+ - 17132

The question of whether or not consolidated interim storage of commercial spent nuclear fuel (SNF) should be part of the federal waste management system as an intermediate step before permanent disposal has been debated for more than four decades. This paper summarizes an evaluation of the cost implications of incorporating a consolidated interim storage facility (ISF) into the waste management system (WMS). In this study, the order-of-magnitude estimates of total system costs were calculated and tabulated. The analyzed scenarios involve shipment of SNF from reactors in dual-purpose storage and transportation canisters (DPCs) currently being used by utilities for dry storage at reactor sites. A number of pertinent conclusions can be drawn from this evaluation, including: - Delay in repository availability increases total system costs. Any delay in opening a repository increases total system costs, regardless of whether the system has an ISF or not. This is due to the increased cost associated with an extended duration for storage, whether at multiple independent spent fuel storage installations (ISFSIs) or at an ISF, until waste can be disposed of.- There is a potential total system lifecycle cost avoidance in realistic scenarios with an ISF when compared to scenarios with no ISF. However, most of the cost avoidance occurs several decades after the ISF is opened. The total WMS cost differential over the long term is mainly attributed to the reduced operational costs of storing the fuel in a consolidated facility rather than at individual reactor sites. - Earlier establishment of an ISF allows for more avoidance of post-shutdown at-reactor storage costs for any repository opening date. An ISF allows earlier acceptance of fuel from reactors, which reduces at-reactor costs from a total system perspective. - Transportation costs have little impact on a WMS with or without an ISF. These impacts range from 3-11% of the total cost in all scenarios. Therefore, transporting the fuel twice does not appear to be a significant cost concern driver relative to other system costs. In conclusion, designing, licensing, and constructing an ISF will require a large near-term financial investment. However, reduced annual operating and maintenance costs for an integrated waste management system with an ISF, compared to a system without an ISF, could eventually result in system-wide (integrated) cost avoidance due to reductions in at-reactor storage costs in the long term. The largest cost avoidances would not occur for several decades, so assumptions about inflation, escalation, and discount rates will have a significant effect on potential economic impacts of an ISF. It should be noted that other benefits associated with the investment of an ISF-such as earlier removal of SNF from shutdown reactor sites, earlier prospects for site re-utilization, and additional overall integrated waste management system flexibility-were not addressed in this study. (authors)

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

DOE SFWST Campaign R&D Roadmap Update

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel and Waste Disposition (SFWD) is conducting research and development (R&D) on deep geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). R&D addressing the disposal of SNF/HLW in the U.S. is currently generic (i.e., "non-site-specific") in scope, following the suspension of the Yucca Mountain Repository Project in 2010. However, to prepare for the eventuality of a repository siting process, the former Used Fuel Disposition Campaign (UFDC) of DOE-NE, which was succeeded by the SFWST Campaign, formulated an R&D Roadmap in 2012 outlining generic R&D activities and their priorities appropriate for developing safety cases and associated performance assessment (PA) models for generic deep geologic repositories in several potential host-rock environments in the contiguous United States. This 2012 UFDC Roadmap also identified the importance of re-evaluating priorities in future years as knowledge is gained from the DOE's ongoing R&D activities.

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

Annual Transuranic Waste Inventory Report – 2018 (Data Cutoff Date 12/31/2017)

The purpose of this Annual Transuranic Waste Inventory Report – 2018 (ATWIR-2018) is to document the inventory estimate of transuranic (TRU) waste reported by the TRU waste generator sites as of December 31, 2017. This report also notes major changes to the inventory used for the Annual Transuranic Waste Inventory Report – 2017 (ATWIR-2017) (U.S. DOE 2017a) that had a cutoff date of December 31, 2016. This updated inventory information is provided to the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) and is available to the DOE TRU waste complex, Waste Isolation Pilot Plant (WIPP) stakeholders, and regulators. The TRU waste inventory information is used for strategic planning, and supports DOE/CBFO’s input into documents (e.g., WIPP Documented Safety Analysis and National Environmental Policy Act analyses), planned changes, and other design changes that may take place in the repository. The DOE/CBFO has determined this report will provide the basis for the Performance Assessment Inventory Report for development of the 2019 Compliance Recertification Application (CRA) deferred performance assessment (PA) (Shrader 2018).

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

Microstructural Characteristics of U-Zr Ingot Slug and Atomized Powder for SFR metal Fuel

Rod-type fuel slugs and spherical powder of U- 10wt.%Zr alloy was also fabricated and characterized in microstructure by injection casting, gravity casting and centrifugal atomization process. U grains became finer from ∼30μm in conventionally ingot rod to ∼2μm in atomized powder. Laminar structure became so fine from ∼0.2μm in conventionally cast ingot rod to 0.05∼0.1μm in atomized powder. Metallic fuels such as U-Zr and U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR), related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980's. Metallic fuel slugs of U-Pu-Zr alloys for SFR have conventionally been fabricated by an injection casting method. U-Pu-Zr alloy melts are injected and cast into quartz tube molds during injection casting. Metallic fuel slugs are collected by fracturing quartz tubes from the molds. Demolded quartz tubes are treated with radioactive wastes. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. The metallic fuel for SFR should be remotely fabricated under a radiation shielded environment such as a hot cell, because it contains highly radioactive MA such as Np, Am and Cm. The addition of MA in the fuel composition introduces new challenges in the manufacturing of a metallic fuel slug. Am has a high vapor pressure under a vacuum atmosphere at above a melting temperature of U-Zr or U-Pu-Zr. Used graphite crucibles and quartz molds cannot be recycled easily because of difficult decontamination of radioactive residues on their surfaces. It is necessary to minimize the vaporization of Am and volume of radioactive wastes. Atomized metallic fuel does not need these quartz molds during fabrication, as uranium alloys are prepared as a particle form. It is possible to reduce the formation of radioactive wastes by applying the fabrication process of a particulate metallic fuel. Metallic fuel having a fine microstructure, which originates from a rapid solidification effect during atomization process, results in a higher FG release rate during irradiation. In the case of applying a new forming method such as vibrocompaction in Zr sheath, the sheath can act as a reliable FCCI (fuel/cladding chemical interaction) barrier and a re-constraint for external elongation. In addition, atomized fuel powder, dispersed in a Na liquid metal having a high thermal conductivity, may have a lower temperature distribution during irradiation. In order to develop an efficient fabrication process of metallic fuel for SFR, U-Zr ingot rods were fabricated in a lab-scale with various casting methods. Rod-type U-Zr fuel slugs were fabricated by gravity casting and injection casting. Spherical powder of U-Zr alloy was also fabricated by centrifugal atomization. And then, the microstructural characteristics were investigated by using electron microscopes and energy-dispersive electron X-ray spectroscopy. Atomized U-10Zr powder generally showed spherical morphology of below 200 μm in particle size. U grains became finer from ∼30 μm in conventionally ingot rod to ∼2 μm in atomized powder. Laminar structure became so fine from ∼0.2 μm in conventionally cast ingot rod to 0.05∼0.1 μm in atomized powder. Atomized particulate metallic fuel for SFR has a fine microstructure, resulting in a higher fission gas release rate during irradiation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Panel reports session

This section brings together the reports of the 31 panel sessions listed below: - Avoiding the Human Capital Cliff in the DOE (PS22); - Implementing Technically-Based Cleanup; US DOE Balancing Regulatory and Fiscal Complexity (PS24); - United Kingdom - Remediation of Sellafield Legacy Ponds and Silos (PS25); - Aligning Decision-Making to Deliver Timely Results in Large Clean-up Programs (PS38); - Transition to Government Owned-Contractor Operated Site at Canadian Nuclear Laboratories (PS39); - Emerging Middle East Nuclear States' Status and Plans (PS40); - NPP Panel on Waste Management (PS41); - Robotics - International Innovations in Robotic Decommissioning Technology (PS43); - Fukushima Panel (1/2): Japan Fukushima Daiichi Decommissioning Update (PS54); - Fukushima Panel (2/2): Update on the Revitalization of Fukushima and the Surrounding Area (PS55); - US DOE LLW On-Site Disposal: LLW Disposal Facility Federal Review Group Review Process (PS56); - Challenges in DOE HLW Tank Management: Facility Start-up and Waste Disposition (PS57); - Waste Management Planning for Utility Small Modular Reactors (PS58); - US Nuclear Power Plant Waste Management - LLW Disposal Issues (PS59); - DOE-EM Robotics and Emerging Technologies Road-map (4/8) (PS60); - Graduating Students and New Scientists and Engineers - Wants and Needs - Are Companies Even Listening? (PS63); - U.S. Nuclear Robotics: Yesterday, Today, Tomorrow (5/8) (PS74); - US DOE Mixed Waste: Proposals for Dealing with Problematic Waste Streams and Policy Changes (PS76); - Current and Emerging US NRC LLW Regulatory Topics (PS77); - Agency Perspectives (US and International) on the Use of Risk and Dose Assessment Tools (PS80); - Progress on Deep Repository Programs Around the World (PS81); - US DOE Procurement and Contracting Opportunities (PS82); - Collaboration across Borders to Deliver Cleanup and Decommissioning Challenges (PS93a); - DOE Mission Success through Application of Practices and Solutions (PS94); - Challenges in Cask Certification - International Perspectives (PS96); - Robotics: Leveraging Innovations in Science and Technology to Benefit Safety Performance (PS98); - DOE Panel: DOE EM Featured Office - Portsmouth/Paducah Project Office (PS109); - Robotics: Lessons Learned from Industrial Robotics Deployment that can be Applied to the Nuclear Field (PS110a); - Roundtable: DOE Radioactive Management - Energy Facility Contractors Group - Waste Management Working Group (EFCOG WMWG) (PS118); - Thursday Topical Session 'Management of Used Nuclear Fuel: Will the United States Reengage and Align with International Efforts' (PS124); - Remotely Operated Vehicles for Emergency, Disaster, and Operational Upset Response (PS125)

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

ORNL/TM-2021/2046 Appendix data repository

Repository of data files used to generate plots for Post-Closure Nuclear Criticality Safety Evaluations for Disposition of Criticality Control Overpacks at the Waste Isolation Pilot Plant (ORNL/TM-2021/2046). See the readme for information contained in the individual files. See the report for description of the models used to generate the data.

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

Safeguards for Reactors and Spent Fuel

It is crucial for safeguards that the amount of plutonium going to the reprocessing plant or repository is what we believe it to be.

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

Flow Rate Dependency of Permeability Changes with the Deposition of CSH under the Highly Alkaline Ca-rich Condition - 17175

The formation of calcium silicate hydrate (CSH) around the geological repository is one of the notable factors for the assessment of the radionuclide migration. This study discussed the clogging effect of CSH in flow path with the change in the flow rate based on the flow experiments using micro flow cell. In the experiment, the change of the permeability with the alteration of granite under the highly alkaline and Ca-rich condition was examined. As a result, the permeability in flow path gradually decreased with the periodical increase and decrease of the permeability. In addition, under the condition of the lower flow rate (0.06 ml/h), the permeability became much smaller than those for other conditions of the higher flow rates. These suggest that the clogging of flow path will progress with the local deposition/moving of CSH, in particular, under the condition of the lower flow rate. Retardation effect of radionuclide migration will be expected because it is prospected that flow rate of groundwater is lower than that of this study in the actual condition of repository. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

MP197HB; First licensed transportation package with canistered high burnup fuel after dry storage - 15580

AREVA TN has recently received the NRC 10 CFR Part 71 transportation license CoC 9302 Revision 7 for the MP197HB transportation package that allows transportation of canistered high burnup fuel assemblies. The Nuclear Horizontal Modular Storage (NUHOMS{sup R}) system canisters containing high burnup PWR and BWR fuel assemblies are now licensed for transportation as a payload in MP197HB after these canisters were in dry storage at the utility ISFSIs. Enhancements have been made to the MP197HB transportation package to improve thermal performance and also provide superior shielding capabilities. Design features of existing AREVA TN transportation packages such as radial fins, thermal shield, neutron resin and optimized gamma and neutron shielding were used to accommodate transportation of high burnup fuel assemblies with heat loads up to 32 kW. The transportation package is designed for exclusive use by rail, truck or marine transport. State-of-the-art analyses methodologies were used to demonstrate compliance with regulatory requirements of 10 CFR 71.55 for the high burnup fuel after dry storage at any ISFSI. Some of the issues addressed in the license application related to the transportation of high burnup fuel included potential degradation of fuel cladding, uncertainty associated with high burnup fuel cladding material properties, and larger radiation and thermal sources due to higher burnups. Bounding conservative assumptions were made where limited test data was available. AREVA TN's licensing approach did not rely on the cladding integrity of high burnup fuel cladding. Instead, the criticality, thermal and shielding analyses were carried out assuming high burnup fuel reconfiguration. With the approval of this transportation package for PWR and BWR canistered high burnup fuel from dry storage at ISFSIs, the utilities can now transport these canisters which are in dry storage at their sites. This is also a major step forward for the canistered fuel that is currently in dry storage at the shutdown plants because they can use this AREVA TN MP197HB transportation package to ship that fuel to any interim storage facility (ISF) or to a repository when they become available. (authors)

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

Millennial Time Capsules as a Promising Means for Preserving Records for Future Generations - 16542

When we deal with preservation of Records, Knowledge and Memory (RK and M) for any long-term, non-inspected facility such a deep geological repository, there is no single technical or cultural provision that can be relied upon to do alone the preservation job 100%. Rather we should increase our chances by implementing a combination of approaches based on different components that provide redundancy and/or pointers to one another. Time capsules are no exception. However, they seem well suited to support national archives and other preservation elements, in order to preserve RK and M as long as possible or interesting. Time capsules are a ready to go, workable concept, with many examples of implementation of large-size, millennial time capsules at small depths based on science and lessons to be learned. A proposal is also made for considering and developing small-size time capsules placed strategically deep underground at repository level. (authors)

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

To the 'Safety related?' Question: What is Related to Safety? - 18430

Most, if not all, national programs for radioactive waste management (RWM) pledge their overall commitment to safety or - in the case of radioactive waste disposal - to long-term safety. Therefore, it may be somewhat surprising to find that the term 'safety' is hardly defined in these programs. The same holds for some of the core international guidance literature on the deep geological repository (DGR) 'safety case' concept. With respect to stakeholder concern over the safety of geological disposal, it seems, however, advisable to seek common ground in the understanding of the idea of 'safety'. In this paper, we review and discuss the most relevant international guidance literature's provision (or non-provision) of definitions of 'safety' in the context of radioactive waste disposal. Against this background, we develop a tentative, practical, glossary-style definition of 'safety' in the DGR context. As a seed of discussion, it may help to eventually expose possible mismatches in the base assumptions of safety experts and other stakeholders. (authors)

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

Estimation of Deposition Behavior of Supersaturated Silicic Acid around Geological Disposal Facility - 19137

For safety evaluation about the geological disposal system, the leakage of radionuclides with groundwater flow is estimated by considering the groundwater flow through the manmade and natural barriers. In such an estimate, we should also consider the influence of cement materials used for the construction of the disposal facility system. Cement material including high alkalinity alters groundwater condition up to 13 in pH and dissolves numerous silicate materials, which release silicic acid to groundwater. Since this groundwater mixes with natural groundwater downstream, supersaturated silicic acid will be produced with the decrease of pH to 8 around the repository system. Thus, in this study, the behavior of supersaturated silicic acid in a flow system was examined by using the column packed with amorphous silica particles under the condition of constant hydraulic head. The packed length is 20 mm or 40 mm, and the supersaturated silicic acid solution was continuously injected into the column. Here, the supersaturated silicic acid was adjusted by mixing with distilled water of pH 8 just before the solution was injected into the column. In the results, the supersaturated silicic acid quickly deposited in the packed column, and colloidal silicic acid was not detected at the outlet of the column. The deposition amount of supersaturated silicic acid in the column decreased with time, and became larger in the longer column. Besides, the flow rate decreased with the deposition of supersaturated silicic acid. Thus, this study monitored the decrease rate of permeability with the supply amount of silicic acid. In the results, the change rate did not depend on packed length. This suggested that even with a short packed length, by considering the amount of precipitation of supersaturated silicic acid, it is possible to estimate the change in permeability similarly to the case of a long packed length. Such a decrease effect in permeability may contribute to retarding the migration of radionuclides around the repository. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗