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At least 739 records · Page 41

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DOE OSTI2012

Destructive examination of shipping package 9975-03431

Destructive and non-destructive examinations have been performed on specified components of shipping package 9975-03431. For those attributes that were also measured during the field surveillance, no significant changes were observed. All observations and test results met identified criteria, or were collected for information and trending purposes. Except for modest corrosion of the lead shield (which is typical of these packages following several years service), no evidence of a degraded condition was found in this package.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

High-level waste relocation requires use of new technology in aging nuclear facility - 15269

The West Valley Demonstration Project (WVDP) is preparing to relocate 278 canisters of vitrified high-level waste (HLW) from their current location inside the Main Plant Process Building (MPPB) to an outdoor pad for long-term, on-site storage. The waste was produced at the WVDP between 1996 and 2002 and was originally placed in interim storage inside the MPPB in anticipation of off-site shipment for disposal. Relocation of the canisters to a new on-site location is now necessary as part of WVDP Phase 1 Site Decommissioning work to enable pre-demolition activities inside the MPPB to proceed. The logistics of packaging and moving the HLW canisters within an aging facility to another on-site location has provided numerous technical challenges with subsequent innovations. Significant preparations have taken place to resolve structural inadequacies, including sub-surface soil subsidence issues, antiquated equipment, and facility reconfigurations to support up to 134 metric ton (148 ton) loads. To perform the transfer and storage of the HLW canisters, new equipment and fabrications were required including a robust storage system consisting of stainless steel canister overpacks within steel-reinforced concrete storage casks, three specialty cask transporters, a remotely operated welding system and construction of an outdoor engineered storage pad to store the 56 casks containing the relocated HLW canisters. Packaging and relocation of vitrified canisters is scheduled to begin in 2015. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Repair of Cracks and Spalls within the Concrete Façade Covering the Runit Island Waste Containment Structure (V.2.0)

The concrete repair shall be performed using a combination of pre-packaged cement-based mortar (Structural Concrete® V/O) supplied by Five Star Products, Inc., and CRACKBOND® products from Adhesive Technology Corp., Pompano Beach, Florida. Five Star Structural Concrete® V/O is a high strength, rapid setting, one component, permanent concrete repair material used primarily for vertical and overhead structural repairs. The material produces a repair which is dimensionally stable, develops an integral bond to existing concrete, and restores structural integrity within hours of placement (Five Star DESIGN-A-SPEC™ Guidelines). The CRACKBOND® family of products are two component, high strength structural epoxy formulations specifically designed to repair cracks in concrete.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2016

The New TN{sup R} MW Cask System Focused on Facilities End-of-Life Waste Management - 16627

When preparing for the decommissioning of a nuclear facility, during its 'end of life' management and while performing the actual dismantling operations, one has to consider a large diversity of nuclear waste in term of types, volumes and activities. Customers are frequently faced with the obligation to undertake multiple and costly waste management operations including handling, reconditioning or re-transferring from one package to another, for example when moving from on-site storage to transportation. To address this issue, a new - highly flexible - cask system named TN{sup R} MW is being developed. This cask has a total weight of 10 T and is compliant with the 2012 IAEA regulations. It is developed on a flexible concept basis, adaptable to the various nuclear needs, including: from IP2 to B(U) / B(U)F; onsite/ international transportation; long term interim storage. Licensing and manufacturing of number of items of this TN{sup R} MW family is underway. (authors)

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DOE OSTI2022

GDSA Framework Development and Process Model Integration FY2022

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 & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for SFWST disposal R&D is disposal system modeling (Sassani et al. 2021). The SFWST Geologic Disposal Safety Assessment (GDSA) work package is charged with developing a disposal system modeling and analysis capability for evaluating generic disposal system performance for nuclear waste in geologic media. This report describes fiscal year (FY) 2022 advances of the Geologic Disposal Safety Assessment (GDSA) performance assessment (PA) development groups of the SFWST Campaign. The common mission of these groups is to develop a geologic disposal system modeling capability for nuclear waste that can be used to assess probabilistically the performance of generic disposal options and generic sites. The modeling capability under development is called GDSA Framework (pa.sandia.gov). GDSA Framework is a coordinated set of codes and databases designed for probabilistically simulating the release and transport of disposed radionuclides from a repository to the biosphere for post-closure performance assessment. Primary components of GDSA Framework include PFLOTRAN to simulate the major features, events, and processes (FEPs) over time, Dakota to propagate uncertainty and analyze sensitivities, meshing codes to define the domain, and various other software for rendering properties, processing data, and visualizing results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

GDSA Framework Development and Process Model Integration FY2023

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DOE OSTI2016

Data Validation Package May 2016 Groundwater Sampling at the Bluewater, New Mexico, Disposal Site, September 2016

Groundwater samples were collected from monitoring wells at the Bluewater, New Mexico, Disposal Site to monitor groundwater contaminants as specified in the 1997 Long-Term Surveillance Plan for the DOE Bluewater (UMTRCA Title II) Disposal Site Near Grants, New Mexico (LTSP). Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). A duplicate sample was collected from location 16(SG).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Innovative Implementation of Consolidated Dry Fuel Storage at the INL - 19655

An innovative nuclear-fuel storage bucket design provides increased fuel storage capacity at the Irradiated Fuel Storage Facility (IFSF) on the Idaho National Laboratory (INL) Site. The IFSF is a dry storage facility that is at approximately 93% of its fuel storage capacity. This innovative fuel storage bucket provides for higher density packaging, thereby allowing for the receipt of more fuel at the IFSF and subsequent completion of the wet-to-dry fuel transfer scope at the INL's Idaho Nuclear Technology and Engineering Center (INTEC). The subject fuel originates from the Advanced Test Reactor (ATR) at the INL Site. All spent nuclear fuel (SNF) at INTEC is required by the Idaho Settlement Agreement (ISA) to be out of wet storage by December 31, 2023. The ISA is a tri-party agreement between the U.S. Department of Energy, the U.S. Navy, and the State of Idaho. Currently, the INTEC CPP-666 pools contain all of the INL's SNF remaining in wet storage, and INTEC's Spent Fuel Project is on schedule to meet the ISA deadline. Interestingly, the INL Site's environmental cleanup contract awarded to Fluor Idaho in 2016 also initially required the Spent Fuel Project to continue SNF receipts from the ATR into wet storage at the CPP-666 pools. The continued receipts of ATR SNF into the CPP-666 basin were able to be redirected to the IFSF because of the new storage package design, enabling the Spent Fuel Project to meet the ISA requirement to have all SNF out of wet storage by 2023. Complicating the need to redirect ATR receipts from wet storage is the lack of interim storage capacity at the ATR site. This lack of ATR storage capacity threatened to terminate reactor operations if the SNF could not be received at INTEC. The contractual requirement for continued ATR SNF receipts to prevent termination of reactor operations is resolved by the implementation of the innovative consolidated dry storage configuration in the IFSF. Consolidated fuel storage at the IFSF provides increased fuel storage capacity to facilitate meeting the ISA and the Fluor Idaho contract requirements. It provides for the completion of the ATR wet-to-dry fuel transfer scope and supports continued operation of the ATR by allowing the receipt of additional ATR SNF directly into IFSF dry storage. (authors)

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DOE OSTI2011

Qualitative and Quantitative Assessment of Nuclear Materials Contained in High-Activity Waste Arising from the Operations at the 'SHELTER' Facility

As a result of the nuclear accident at the Chernobyl NPP in 1986, the explosion dispeesed nuclear materials contained in the nuclear fuel of the reactor core over the destroyed facilities at Unit No. 4 and over the territory immediately adjacent to the destroyed unit. The debris was buried under the Cascade Wall. Nuclear materials at the SHELTER can be characterized as spent nuclear fuel, fresh fuel assemblies (including fuel assemblies with damaged geometry and integrity, and individual fuel elements), core fragments of the Chernobyl NPP Unit No. 4, finely-dispersed fuel (powder/dust), uranium and plutonium compounds in water solutions, and lava-like nuclear fuel-containing masses. The new safe confinement (NSC) is a facility designed to enclose the Chernobyl NPP Unit No. 4 destroyed by the accident. Construction of the NSC involves excavating operations, which are continuously monitored including for the level of radiation. The findings of such monitoring at the SHELTER site will allow us to characterize the recovered radioactive waste. When a process material categorized as high activity waste (HAW) is detected the following HLW management operations should be involved: HLW collection; HLW fragmentation (if appropriate); loading HAW into the primary package KT-0.2; loading the primary package filled with HAW into the transportation cask KTZV-0.2; and storing the cask in temporary storage facilities for high-level solid waste. The CDAS system is a system of 3He tubes for neutron coincidence counting, and is designed to measure the percentage ratio of specific nuclear materials in a 200-liter drum containing nuclear material intermixed with a matrix. The CDAS consists of panels with helium counter tubes and a polyethylene moderator. The panels are configured to allow one to position a waste-containing drum and a drum manipulator. The system operates on the ‘add a source’ basis using a small Cf-252 source to identify irregularities in the matrix during an assay. The platform with the source is placed under the measurement chamber. The platform with the source material is moved under the measurement chamber. The design allows one to move the platform with the source in and out, thus moving the drum. The CDAS system and radioactive waste containers have been built. For each drum filled with waste two individual measurements (passive/active) will be made. This paper briefly describes the work carried out to assess qualitatively and quantitatively the nuclear materials contained in high-level waste at the SHELTER facility. These efforts substantially increased nuclear safety and security at the facility.

Cherkas, Dmytro↗

DOE OSTI2025

Blending Compostable Plastics for Packaging Applications

Non-sustainable packaging materials (polymers) are a large portion of the increasing amount of plastic waste that become environmental pollutants. While many biodegradable or compostable polymers have been developed in recent years, most fail to compete with the non-sustainable polymers dominating the market, due to lack of certain desired properties, such as thermal (high melt temperature) and mechanical (high ductility) performance. For example, the popular compostable polymer polylactic acid (PLLA) has a high melt temperature but is very brittle. Another polymer, poly(d-valerolactone) (PVL) demonstrates great ductility but has a very low melt temperature. By blending PLLA and PVL, there is an opportunity to create a new biodegradable material with synergistic desired properties for packaging applications. In this study, we produce several (9) physical blends of PVL and PLLA with varying compositions of the two polymers while applying three different materials to make them more compatible (compatibilizers). We test the success of compatibilization by scanning electron microscopy (SEM), mechanic testing (strain at break) and differential scanning calorimetry (melting temperature). Overall, we highlight several promising materials with high compatibility and desired thermomechanical properties for sustainable packaging. Ideally, these materials could help mitigate future plastic pollution in Colorado and beyond.

blending↗

NASA NTRS1973

Feasibility of Space Disposal of Radioactive Nuclear Waste. 1: Executive Summary

This NASA study, performed at the request of the AEC, concludes that transporting radioactive waste (primarily long-lived isotopes) into space is feasible. Tentative solutions are presented for technical problems involving safe packaging. Launch systems (existing and planned), trajectories, potential hazards, and various destinations were evaluated. Solar system escape is possible and would have the advantage of ultimate removal of the radioactive waste from man's environment. Transportation costs would be low (comparable to less than a 5 percent increase in the cost of electricity) even though more than 100 space shuttle launches per year would be required by the year 2000.

Source record↗

DOE OSTI2012

Unconventional Staging Package Selection Leads to Cost Savings

In late 2010, U.S. Department of Energy (DOE) Deputy Secretary of Energy, Daniel Poneman, directed that an analysis be conducted on the U-233 steel-clad, Zero Power Reactor (ZPR) fuel plates that were stored at Oak Ridge National Laboratory (ORNL), focusing on cost savings and any potential DOE programmatic needs for the special nuclear material (SNM). The NA-162 Nuclear Criticality Safety Program requested retention of these fuel plates for use in experiments at the Nevada National Security Site (NNSS). A Secretarial Initiative challenged ORNL to make the first shipment to the NNSS by the end of the 2011 calendar year, and this effort became known as the U-233 Project Accelerated Shipping Campaign. To meet the Secretarial Initiative, National Security Technologies, LLC (NSTec), the NNSS Management and Operations contractor, was asked to facilitate the receipt and staging of the U-233 fuel plates in the Device Assembly Facility (DAF). Because there were insufficient staging containers available for the fuel plates, NSTec conducted an analysis of alternatives. The project required a staging method that would reduce the staging footprint while addressing nuclear criticality safety and radiation exposure concerns. To accommodate an intermediate staging method of approximately five years, the NSTec project team determined that a unique and unconventional staging package, the AT-400R, was available to meet the project requirements. By using the AT-400R containers, NSTec was able to realize a cost savings of approximately $10K per container, a total cost savings of nearly $450K.

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DOE OSTI2017

Novel Gamma Attenuation Methods Utilized for Waste Characterization of UO{sub 2}F{sub 2}-Bearing Materials at the Paducah GDP - 17327

The Paducah Gaseous Diffusion Plant (PGDP) is currently under deactivation with demolition in the decades to follow. UO{sub 2}F{sub 2}-bearing waste is generated at PGDP by UF{sub 6} hydrolysing to materials such as pipes, valves, PPE, etc. when UF{sub 6} is exposed to humid air. In addition, Al{sub 2}O{sub 3} and NaF trap materials contain significant amounts of Uranium. The UO{sub 2}F{sub 2} deposits within the generated waste can be of sub-gram levels to several hundreds of localized grams and are much denser than the surrounding interfering matrix materials and may be of very complex geometric and chemical forms such as with the Al{sub 2}O{sub 3} and NaF trap materials. The main objective is to characterize the fissile (U-235) content of the UO{sub 2}F{sub 2}-bearing waste so as to meet the Data Quality Objectives (DQO) for proper disposal. If a waste container cannot meet the DQO then it must be subjected to costly remediation, repackaging and re-characterization processes. We have developed, at PGDP, several novel model-based methods and techniques that vastly improve the standard gamma attenuation correction methods and do not require Working Reference Materials (WRM) for calibration. These new methods involve rotating and non-rotating 18.9 liter (5 U.S. gallon) waste containers with and without a transmission modality. A new model-based variant of differential attenuation has been developed that provides a high-quality self-absorption correction that does not require knowledge of the Uranium enrichment or isotopic ratios. For more challenging waste containers, a method using multi-view (non-rotating) 18.9 liter waste containers has been considered using a recently developed Advanced ISOCS{sup C} Uncertainty Estimator (AIUE) package. The new methods suffice to significantly improve accuracy and precision and to lower the Total Measurement Uncertainty (TMU) leading to satisfactory DQO's that were not previously achievable using the standard gamma attenuation correction methods. In this paper we present real 18.9 liter waste container and Working Reference Material (WRM) data subjected to the new methods and compared to standard methods depicting the significant improvement in the DQO that is realized. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Cell G 7606A Mockup Measurement Analysis Methods and Results

The production of research isotopes at the Radiochemical Engineering Development Center (REDC) invariably leads to the accumulation of radioactive waste. Storage of this waste onsite at ORNL serves as an interim solution prior to shipment to long term storage facilities, however onsite capacity is limited. Proper characterization of the activity of waste products is essential for determining the appropriate waste stream and, ultimately, mitigating the cost of disposal. This is typically done via gamma spectrometry and use of the In-Situ Object Counting System (ISOCS), a software package from Mirion that serves as an accepted community standard. However, variability in the contents, density, and activity distribution in waste containers can introduce large errors in ISOCS quantification. A measurement campaign using a mockup of the proposed setup in Cell G seeks to quantify the magnitude and source of these systematic errors. Analysis of this data will guide the creation of ISOCS geometry templates and measurement methods designed to minimize overall uncertainty in reported activities.

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DOE OSTI2017

Preliminary Concept of Operations for the Spent Fuel Management System--WM2017

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the U.S. Department of Energy s Office of Nuclear Energy is tasked with identifying, planning, and conducting activities to lay the groundwork for developing interim storage and transportation capabilities in support of an integrated waste management system. The system will provide interim storage for commercial spent nuclear fuel (SNF) from reactor sites and deliver it to a repository. The system will also include multiple subsystems, potentially including; one or more interim storage facilities (ISF); one or more repositories; facilities to package and/or repackage SNF; and transportation systems. The project team is analyzing options for an integrated waste management system. To support analysis, the project team has developed a Concept of Operations document that describes both the potential integrated system and inter-dependencies between system components. The goal of this work is to aid systems analysts in the development of consistent models across the project, which involves multiple investigators. The Concept of Operations document will be updated periodically as new developments emerge. At a high level, SNF is expected to travel from reactors to a repository. SNF is first unloaded from reactors and placed in spent fuel pools for wet storage at utility sites. After the SNF has cooled enough to satisfy loading limits, it is placed in a container at reactor sites for storage and/or transportation. After transportation requirements are met, the SNF is transported to an ISF to store the SNF until a repository is developed or directly to a repository if available. While the high level operation of the system is straightforward, analysts must evaluate numerous alternative options. Alternative options include the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if any), repackaging technology, the types of containers used, repository design, component sizing, and timing of events. These alternative options arise due to technological, economic, or policy considerations. As new developments regularly emerge, the operational concepts will be periodically updated. This paper gives an overview of the different potential alternatives identified in the Concept of Operations document at a conceptual level.

Cumberland, Riley M↗

DOE OSTI2024

Preliminary Analysis of Advanced Reactor Spent Nuclear Fuel Storage, Transportation, and Disposal

Due to increased interest in advanced reactor deployment and their associated potential new fuel cycles, the U.S. Department of Energy (DOE) Spent Fuel and Waste Science and Technology (SFWST) program has begun to evaluate the possible implications of long term management and final disposition of the spent nuclear fuel (SNF) generated. Safely managing and dispositioning this SNF, along with any other associated radioactive waste, is the primary focus of this initial preliminary assessment. This paper summarizes efforts to evaluate the characteristics and packaging options for three types of advanced reactor SNF forms: (1) tristructural isotropic (TRISO), (2) metallic, and (3) irradiated fuel salt presented in the report titled “Storage, Transportation, and Disposal of Advanced Reactor Spent Nuclear Fuel and High-Level Waste”. TRISO and metallic SNF and their associated waste streams were emphasized because of the near-term anticipated demonstrations of X-energy’s Xe-100 and TerraPower and GE Hitachi’s Natrium advanced reactors. Preliminary information on spent fuel salts discharged from molten-salt reactors (MSRs) was also examined to provide a baseline for future efforts. All calculations and assumptions were based on publicly available information. This paper identifies several different reactors that produce either TRISO or metallic SNF as well as a few of the reactor and fuel characteristics used for the assessments. Based on these characteristics, calculations were performed to determine the applicability of packaging SNF into existing canister designs. The 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, and as-loaded radiation, thermal, and criticality analyses to identify constraints on storage, transportation, and disposal. Gaps in publicly available data pertaining to reactor operation and/or fuel composition provide increased uncertainty in some evaluations. Additionally, uncertainty in packaging and SNF management operations provide additional uncertainty. However, preliminary conclusions can still be assessed through this work and are presented in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Effects of Aging and Different Mechanical Recycling Processes on the Structure and Properties of Poly(lactic acid)-clay Nanocomposites

The growing use of poly(lactic acid) (PLA) and PLA-based nanocomposites in packaging has raised the interest of studying the mechanical recycling of the wastes and the properties of the recycled materials. The main objective of this work was to study the effect of two different mechanical recycling processes on the structure and properties of a PLA-montmorillonite nanocomposite. The two recycling processes included accelerated thermal and photochemical aging steps to simulate the degradation experienced by post-consumer plastics during their service life. One of them also included a demanding washing process prior to the reprocessing. A decrease in the molecular weight of PLA was observed in the recycled materials, especially in those subjected to the washing step, which explained the small decrease in microhardness and the increased water uptake at long immersion times. Water absorption at short immersion times was similar in virgin and recycled materials and was accurately described using a Fickian model. The recycled materials showed increased thermal, optical and gas barrier properties due to the improved clay dispersion that was observed by XRD and TEM analysis. The results suggest that recycled PLA-clay nanocomposites can be used in demanding applications.

77 NANOSCIENCE AND NANOTECHNOLOGY↗

DOE OSTI2017

Design update of the ITER upper wide angle viewing system

One of the diagnostic systems being provided by the US is the Upper Wide Angle Viewing System (UWAVS) which provides real-time, simultaneous visible and infrared (IR) images of the ITER divertor regions via optical systems located in five upper ports. UWAVS is designed in three main sections: in-vessel, interspace and port cell assemblies. Each assembly utilizes multiple steering and relay mirrors to direct the in-vessel light out of the tokamak to the port cell camera sensors. The primary design challenge of the in-vessel assembly is maximizing performance of the overall system while surviving the severe electromagnetic and nuclear ITER environment. A first mirror material study was conducted and determined that single crystal molybdenum was the best choice for the first two mirrors in the optical train. A fail open, bellows actuated shutter with cross pivot flexure design was determined to be the most reliable mechanism to protect the first mirror. A geometrically representative glow discharge mirror cleaning system is being created and will be tested to maximize cleaning effectiveness while minimizing optical degradation of the molybdenum mirrors adjacent to the plasma. The shutter and first mirror assemblies were packaged and designed for replacement via remote handling methods to minimize radiation waste and cost associated with eventual component replacement. The preliminary optical and structural design provides a robust and reliable system while maximizing the field of view. The R&D efforts, the technical challenges and issues, and the design and analysis results are introduced.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗