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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Preliminary Analysis of Advanced Reactors Storage, Transportation, and Disposal

Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP[1]) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes the efforts by the Spent Fuel & Waste Science and Technology (SFWST) in evaluating characteristics and packaging options for advanced reactor spent nuclear fuel forms. The fuel forms were categorized into three types: (1) tristructural isotropic (TRISO), (2) metallic, and (3) fuel salt. This work emphasized TRISO and metallic SNF and waste streams because of the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations. Preliminary information for the spent-fuel salts discharged from molten-salt reactors (MSRs) is also examined to provide a baseline for future efforts. All calculations and assumptions used publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. The paper also includes a literature review and analysis on the storage, transportation, and disposal evaluations and experiences from reactors with similar fuel forms. Advanced-reactor vendors cite past experiences with Fort St. Vrain for TRISO and the Experimental Breeder Reactor II (EBR-II) for metallics that have major influences on fuel design. Finally, the paper includes preliminary concepts of operation for advanced-reactor SNF. This encompasses storage, transportation, potential treatment, and disposal activities from both a per-canister and systems-integration perspective.

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DOE OSTI2022

Annual status update for OWL

This report represents completion of milestone deliverable M2SF-23SN010309082 Annual Status Update for OWL due on November 30, 2022. It provides the status of fiscal year 2022 (FY2022) updates for the Online Waste Library (OWL).

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DOE OSTI2024

Literature Review Investigating Adsorbents for Mercury Removal from Salt Batch Feed Material

The Savannah River Site (SRS) Liquid Waste System (LWS) contains mercury that originated from decades of canyon processing where it was used as a catalyst for dissolving aluminum cladding of reactor fuel. Since the start of radioactive operations, the Salt Waste Processing Facility (SWPF) has experienced processing difficulties due to the presence of mercury-laden solids. In response to these challenges, Savannah River National Laboratory (SRNL) has been requested by Savannah River Mission Completion (SRMC)/Salt Waste Processing Facility (SWPF) personnel to perform a literature review to assess the feasibility of chemical strikes to salt waste-containing tanks at the SRS as an option to reduce mercury introduction into SWPF. The literature review should evaluate known technologies and chemistries for decreasing mercury concentration and the anticipated efficacy of each technology on various mercury species known to exist in SRS waste.

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DOE OSTI2024

Transportation and Systems Analysis Collaborations in Support of a Federal Consolidated Interim Storage Facility

The U.S. Department of Energy’s Integrated Waste Management (IWM) program under the Office of Nuclear Energy is planning for the future transportation, storage, and eventual disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from nuclear power plant sites across the United States. To better enable informed decision-making regarding the back end of the nuclear fuel cycle, the IWM program has been sponsoring the development and application of system analysis tools capable of analyzing various options for managing SNF and HLW. With these tools, integrated waste management system (IWMS) architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). System analyses and assessments typically use these modeling and simulation tools to investigate implications of changes in various assumptions and parameters such as acceptance rates, receipt logic, facility capacities and capabilities, use of standardized canisters, start and stop dates of facilities, etc. The Next Generation System Analysis Model (NGSAM) is an agent-based simulation toolkit that is used for a system-level simulation and analysis focused on SNF management in the United States. An analyst using NGSAM has the ability to define several factors like the number of storage facilities, capacity at each facility, transportation schedules, shipment rates, and other conditions. NGSAM’s primary purpose is to provide a system analyst with capabilities to model the IWMS and gain insights into SNF and HLW management alternatives including the impact of system choices, associated cost estimates, and development of integrated yet flexible approaches. IWM is also developing the Stakeholder Tool for Assessing Radioactive Transportation (START). START is a web-based geospatial tool developed to provide visualization and initial evaluation of transportation options associated with future SNF and HLW shipment planning and operations. This includes characterizing safety, economic, and environmental conditions on and in proximity to shipment origins as well as along prospective transportation routes. Information from the START tool can be presented/shared as maps, graphics, geospatial files, and tabular form to enable easy data representation and export functionality. The system analysis, NGSAM, and START teams had been working closely for several years before formalizing this collaboration. START provides the SNF routing information for use in NGSAM. System analysts use the NGSAM tool to generate results (schedule, costs, infrastructure acquisitions, shipment rates, etc.). Subject-matter experts and system analysts work together to inform how NGSAM should model the waste management system. Specifically, this paper discusses the collaborations between the system analysis, NGSAM, and START teams and their accomplishments over the past year. Some examples of these efforts include calibrating and adding data to the START output files to meet NGSAM needs, gaining a better understanding of START data used in NGSAM, as well as how updates in START data could result in an improved NGSAM analysis. Detailed examples of various tasks that have been performed by the team will be discussed in the full paper. Continued efforts in this direction are expected in the coming years.

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DOE OSTI2022

Overview of System Integration Analysis Activities for Integrated Waste Management

Spent nuclear fuel (SNF) generated by the current fleet of commercial nuclear reactors is being stored at reactor sites in spent fuel pools (SFPs) and in dry independent spent fuel storage installations (ISFSIs). The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Integrated Waste Management (IWM) program is examining a suite of IWM system options and conducting supporting analyses to enable future informed choices. The IWM program is currently organized into the following four major areas: (1) IWM facilities and equipment concepts and development, (2) transportation capability analysis and support, (3) information technology solutions and support, and (4) system integration analysis and support. This paper focuses on the activities ongoing in the IWM system integration analysis and support area. Two main research activities in this area are: data and tools development, validation, and maintenance; and special studies, analyses, and assessments.

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DOE OSTI2020

A Database for Radwaste Management in Germany-30 Years of Experience by TUV NORD - 20461

From the very beginning in the late 1980's TUV NORD EnSys is involved as an independent expert in quality assurance processes concerning interim storage facilities and the disposal for radioactive waste with negligible heat generation in Germany. A database has been developed since these times by TUV NORD EnSys. The development of this database over a long time will be discussed as well as the lessons learned within this process. (authors)

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DOE OSTI2023

Considerations for Managing DOE Standard Canisters within an Over-canister as Part of an Integrated Waste Management System

To better enable informed decision making regarding the back end of the nuclear fuel cycle, the Integrated Waste Management Program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the research into a comprehensive integrated waste management system (IWMS) that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). An important aspect of the IWMS is DOE-managed SNF. DOE and its predecessor agencies have generated, transported, received, stored, and reprocessed SNF at DOE facilities nationwide, and DOE is responsible for managing the SNF currently in its possession. These fuels come from a wide range of reactor types that employ various cladding materials, fuel materials, and enrichments. To enable interim, road-ready dry storage (RRDS) of the wide variety of SNF types found in the DOE inventory, a standardized canister system (i.e., the DOE Standard Canister) was proposed. This robust, welded canister system is designed to confine radionuclides, prevent criticality by precluding content moderation, and satisfy other requirements as part of a larger storage, transportation, and disposal system. While SNF has yet to be loaded into a DOE Standard Canister, DOE Standard Canister designs were included in past storage facility and disposal facility design licensing endeavors. In a renewed effort to evaluate packaging SNF at Idaho National Laboratory (INL) in a RRDS configuration, researchers are planning the RRDS Packaging Demonstration. This demonstration is supplemented by analytical structural, criticality, and material compatibility evaluations that support management of SNF in DOE Standard Canisters, taking advantage of past analysis work to the extent possible. One of the largest differences between the current Packaging Demonstration and past analytical evaluations is the inclusion of an over-canister containing multiple DOE Standard Canisters. For the Packaging Demonstration, DOE Standard Canisters loaded with SNF are planned to be placed in a larger diameter over-canister. The sealed over-canister could then be placed in a storage overpack for onsite storage, or in a transportation overpack for shipment to an offsite storage location or disposal site once one becomes available. This paper examines the relevant considerations and provides a preliminary evaluation of integrating the over-canister configuration into the storage, transportation, and disposal processes of the overall waste management system. For storage and transportation, the over-canister can be considered analogous to a multi-purpose canister (MPC) for commercial SNF. For disposal, the DOE Standard Canisters could be removed from the over-canister and placed in a co-disposal waste package with canisters containing vitrified HLW similar to configurations examined previously, or the sealed over-canisters might be capable of direct disposal in a waste package.

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DOE OSTI2022

Verification and Validation of the Stakeholder Tool for Assessing Radioactive Transportation (START) – 22323

The U.S. Department of Energy (DOE) Office of Integrated Waste Management is planning for the eventual transportation, storage, and disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from nuclear power plant and DOE sites. The Stakeholder Tool for Assessing Radioactive Transportation (START) is a web-based, geospatial decision-support tool developed for evaluating routing options and other aspects of transporting SNF and HLW, covering rail, truck, barge, and intermodal infrastructure and operations in the continental United States. The verification and validation (V&V) process is intended to independently assess START to provide confidence in the ability of START to accurately provide intended results. The V&V process checks the START tool using a variety of methods, ranging from independent hand calculations to comparison of START performance and results to those of other codes. The V&V activity was conducted independently from the START development team with opportunities to provide feedback and collaborate throughout the process. The V&V analyzed attributes of transportation routes produced by START, including route distance and both population and population density captured within buffer zones around routes. Population in the buffer zone, population density in the buffer zone, and route distance were all identified as crucial outputs of the START code and were subject to V&V tasks. Some of the improvements identified through the V&V process were standardizing the underlying population data in START, changing the projection of the population raster data, and changes to the methodology used for population density to improve its applicability for expected users. This collaboration also led to suggested improvements to some of the underlying shape file segments within START. A formal series of test routes went through the V&V process from START version 3.2.1 following the implementation of the recommended improvements to START. The V&V of the buffer zone population and distance reported in START were achieved by utilizing a custom workflow developed in QGIS. This platform was chosen because it is fully independent of START. The buffer zone population and route distances in START showed excellent agreement to independent V&V test results. Over 200 route test cases were run; in all cases, the percent difference in the population within the buffer zone was less than +/- 5%, and the majority of cases were below +/- 1%. These cases were also duplicated in ArcMap, which is Environmental Systems Research Institute’s (ESRI) GIS desktop application. The ArcMap results also show good agreement; however, they cannot be considered independent of START because START also uses ESRI software. The V&V tasks are currently addressing population density and estimated radiation dose in START as well as implementing further automation of the V&V process to allow larger suites of test routes to be run more easily and quickly by less experienced GIS users. These improvements to the V&V process will make it a valuable tool for the development team and the stakeholders so that each new iteration of START can easily be checked for quality and consistency. This paper and presentation will describe the V&V process, findings and implications, and ongoing V&V activities.

START, DOE, Transporation, dose, GIS↗
DOE OSTI2022

OPP-B332-032 Rev AD

The purpose of this procedure is to provide the approved method for converting transuranic (TRU) contaminated liquid waste into acceptable TRU waste forms.

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DOE OSTI2022

OPP-B332-035 Rev AD

This procedure provides instructions for converting transuranic (TRU) contaminated liquid waste into acceptable TRU waste forms.

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DOE OSTI2022

PTSFA QA Review of Container Procurements: Information Use Procedure

The purpose of this Packaging and Transportation Safety Functional Area (PTSFA) procedure is to describe the process in which the PTSFA Quality Assurance Engineer (QAE) or designee reviews and approves the documentation associated with the procurements of containers to be used for onsite transfer or off-site transportation of hazardous materials and waste.

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DOE OSTI2020

Life-Cycle Baseline Customization for the Formerly Utilized Sites Remedial Action Program - 20304

One of the responsibilities of the federal government is to estimate sound and defensible life-cycle baseline costs for use in federal budget estimates and to meet federal financial reporting requirements. To accomplish this, the US Department of Energy (DOE) Office of Legacy Management (LM) and the US Army Corps of Engineers (USACE) have partnered together to ensure that the liabilities documented in each Formerly Utilized Sites Remedial Action Program (FUSRAP) site's life-cycle baseline are specifically tailored. FUSRAP was created in the mid-1970's to clean up radiological contamination resulting from the early development of nuclear weapons. DOE was responsible for FUSRAP until October 1997, when Congress transferred the administration and execution of FUSRAP site cleanups to USACE. By 1997, DOE had completed the cleanup of 25 of the 46 sites that were active within the program and had begun cleanup at 13 additional sites. USACE was assigned responsibility for the cleanup of the 21 remaining FUSRAP sites, and at 8 additional sites that had since been referred for cleanup. The LM mission for the FUSRAP sites is to perform long-term surveillance and maintenance (LTS and M). Currently, LM provides long-term stewardship for 34 completed FUSRAP sites. Another 20 sites are under active remediation by USACE. Within the last 5 years, USACE has completed the cleanup at five FUSRAP sites and the LTS and M responsibility has been transferred to LM. By 2029, USACE will compete remediation at eight additional sites and LTS and M for those sites will transfer to LM. Because responsibilities for the FUSRAP sites transfer between USACE and LM upon completion of remedial actions, both agencies maintain life-cycle baselines for different stages of the project and both must have a strong understanding of the needs and requirements for each site. This understanding ensures that the life-cycle baselines form a complete and accurate picture of what is required for the site and for the FUSRAP program. LM focuses on several things when customizing the life-cycle baseline estimates for the FUSRAP sites, including (1) Understanding the unique requirements for each site. By reviewing site-specific documents prepared by USACE, such as Feasibility Studies and Records of Decision, and partnering with USACE to gain additional insight about site conditions and requirements for stewardship, as well as potential risks, LM can better develop the life-cycle baselines. (2) Developing site-specific labor breakdowns. This ensures the required labor mix is baselined for specific activities by comparing the labor mix required to perform activities at (a) other LM-managed FUSRAP sites and (b) non-FUSRAP LM sites and (c) by USACE at active FUSRAP sites. (3) Taking a tiered approach to life-cycle baseline planning. Estimates for sites transferring to LM in the near term (5 years) are more definitive than for sites transferring in the out-year period. Remedial actions at the near-term sites are at or near completion, providing LM a strong understanding of the LTS and M requirements and remaining liabilities. This in turn allows for site-specific customization of the baseline. (4) Using a robust risk management approach to ensure that liabilities specific to each site are identified, evaluated by probability and severity, and documented in relation to the impact to cost or schedule. To ensure the most accuracy within all the baselines, the FUSRAP life-cycle baselines are updated as needed to support program, project, and contract management needs. (authors)

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DOE OSTI2025

Colloidal Behavior of Plutonium Oxide in Concentrated Electrolyte Solutions

The Hanford Site in Washington State manages legacy high-level radioactive waste streams that display major chemistry and engineering challenges, including the high salt levels and pH values that correspond to conditions under which many classical concepts describing chemical reactivity cannot be applied. One particular challenge that needs to be tackled at Hanford is that Pu concentrations, [Pu], in the soluble phases of the tank wastes are higher than expected based on the solubility of crystalline PuO2, which is widely accepted to be caused by the formation of PuO2 colloid, consisting of nano- to submicron-sized particles (PuO2 NPs). Fundamental research underpinning the behavior of PuO2 NPs under conditions not only relevant to the Hanford tank waste but at high ionic strength in general is needed to reliably predict the chemical reactivity of PuO2 NPs and develop engineering solutions to safely and efficiently process high-level radioactive waste into forms suitable for long-term storage. In this work, we study the behavior of PuO2 NPs (particle size ~100 nm) under high ionic strength conditions by reacting it with highly concentrated (up to 5 M) salt solutions. We explore different electrolyte compositions to elucidate the impact of different anions (NO3-, Cl-, ClO4-, SO42-, C2O42-, CO32-) on the stability of PuO2 NP in the acidic and alkaline pH regime. PuO2 NP aggregation and precipitation as function electrolyte concentration is tracked by a combination of liquid scintillation counting, dynamic light scattering for determination of particle size distributions, and zeta potentials as a proxy for particle charge. At acidic pH, electrolytes containing non-coordinating anions, such as NaNO3, NaCl, and NaClO4 mostly stabilize PuO2 NPs over a large electrolyte concentration range, showing only subtle differences in their reactivity. Other electrolyte anions show a more pronounced effect on the PuO2 NP stability: SO42-, binds directly to the particles’ surface, reverses the particle charge, and precipitates the PuO2 NPs efficiently even at intermediate sulfate concentrations (>0.1 M). In contrast, C2O42- is found to lead to high [Pu] in solution, in the milli-molar range, even at mildly acidic pH (~4). Thermodynamic modeling of the dissolved Pu concentrations using PHREEQC is unable to predict the observed [Pu] in the acidic pH regime, supporting the influence of colloids in maintaining elevated [Pu]. It is noteworthy that the current thermodynamic databases do not include constants for colloidal Pu phases and cannot accurately predict many of the high ionic strength solutions relevant to this work. The mechanisms and models responsible for these observations will need further investigation in the future. At high pH values (~12), PuO2 NPs exhibits classical sol-gel chemistry, meaning that upon destabilization of the colloidal sol, for example by addition of concentrated NaOH, highly porous and viscid PuO2 coagulates are formed that consist of a three-dimensional network likely held together by physical interactions. The PuO2 NP coagulate shows no significant reversibility of the aggregation when contacted with concentrated brines; however, PuO2 NPs can be efficiently resuspended in solution by addition of diluted electrolytes, alkaline solutions containing high amounts of carbonate, or simple addition of water. Especially carbonate is shown to stabilize PuO2 NPs in solution at high pH, characterized by stable colloidal suspensions that are resistant against sedimentation during centrifugation. Thermodynamic modeling of the carbonate system was able to predict an increasing dissolved Pu concentration with increasing carbonate concentration. However, the model was profoundly sensitive to the fixed redox potential and does not include any thermodynamic constants for colloidal Pu species.

Neumann, Julia↗
DOE OSTI2022

On-Line Waste Library V4.0 Supporting Information

The On-Line Waste Library is a website that contains information regarding United States Department of Energy-managed high-level waste, spent nuclear fuel, and other wastes that are likely candidates for deep geologic disposal, with links to supporting documents for the data. This report provides supporting information for the data for which an already published source was not available.

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