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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.

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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Removal of Metal Oxides from Reactor Metals/Alloys by Laser Ablation

Storage of low and intermediate level radioactive waste in a deep geological repository (L and ILW DGR): - Large volumes of metallic waste = increased production of hydrogen gas. - Large fraction of the metallic wastes are out-of-core reactor components. - Carbon steel feeder pipes. - Radioactive contamination is concentrated within the surface oxides. - Removal of surface contaminants would lead to volume reduction of metallic waste that is stored in L and ILW DGR. - Current decontamination techniques generate large volumes of secondary waste. Objectives: Oxide removal from carbon steel using a ytterbium fiber laser (1064 nm) with a particular focus on identifying the key laser parameters that affect the removal efficiency. Advantages of laser ablation: Generate low volumes of waste, Flexibility of deployment, Remote operation. Grow an oxide film on carbon steel with the incorporation of cobalt. Optimize laser cleaning technology for the removal of surface oxides. Incorporate an oxide collection system for the ablated particles. Conclusions: Objective 1 - Simulated inactive oxides on carbon steel, using Co{sup 2+} as a surrogate for Co radionuclide. Objective 2 - Optimized laser parameters (200 MW/cm{sup 2}, 3 scans, 15 μm overlap); Complete removal of cobalt from surface and bulk metal with optimized cleaning settings. Objective 3 - Glass fiber media H14 HEPA filters had highest collection yield; High oxide collection yields (∼100%). Future Work: Further testing of the liquid trap will verify the efficacy of the HEPA filter and determine the amount of ablated oxides that pass through. Testing of the laser ablation system needs to be done on other reactor materials along with the incorporation of different transition metal cations. Improvements to the oxide collection system are required to improve collection yields.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Overview of Salt Repository Research and Development for Spent Nuclear Fuel and High-level Nuclear Waste in the United States - 20307

This paper summarizes the current US Department of Energy Office of Nuclear Energy's (DOE-NE) work towards developing and executing a research and development program that addresses both scientific and technical issues related to long-term disposal of spent nuclear fuel (SNF) and high-level waste (HLW) in a hypothetical bedded salt based geological repository. A primary goal of the program is to create a generic Geologic Disposal Safety Assessment (GDSA) Framework that can be used to help guide decisions on siting a possible future bedded salt repository. The generic GDSA work includes analysis of the impacts of heat generation caused by decay of short-lived radionuclides. We report progress in four primary areas. First, we discuss the development and recent modifications of a research and development road-map. Second, we briefly describe an experimental approach to better understand thermal processes in salt. Third, we highlight collaborations with the international research community that leverage salt-based repository science around the world. Finally, we discuss how our findings are being used to aid in the development of a generic safety assessment for a bedded salt repository containing SNF and HLW. (authors)

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DOE OSTI2018

Generic HLW Repository In Bedded Salt In Germany: Emplacement Technologies and Repository Designs - 18288

In Germany, a new site selection act for a repository for HLW and SNF entered into force in July 2013 and was amended in spring 2017 [1]. It stipulates that the safety of repository systems for heat-generating radioactive waste and spent nuclear fuel in different host rock formations be compared. This requires that generic repository concepts and adequate safety- and safety demonstration concepts for all potential host rock formations in Germany (salt, clay and crystalline rock) exist. For several decades, development of disposal concepts and safety analyses for a repository in a salt dome was prioritized while repository concepts for clay and crystalline rock have been taken into account for two decades only. So far, bedded salt formations have not been the subject of comparative investigations. Thus, in summer 2015, the Federal Ministry for Economic Affairs and Energy (BMWi) launched the R and D project KOSINA to develop repository designs and carry out safety analyses for different emplacement alternatives in a generic bedded salt formation and a generic salt pillow. KOSINA is an acronym for 'Concept development for a generic repository for heat-generating waste in bedded salt formations in Germany as well as development and testing of a safety and demonstration concept'. Together with its partners BGR, GRS, and IfG, DBE Technology GmbH has successfully carried out an almost three years lasting scientific programme, which will be concluded in April 2018. The core areas of the programme are the development of generic geologic models, including derivation of model parameters, the development of a safety and safety demonstration concept, the development of technical repository designs for four different emplacement alternatives, the analysis of geomechanical integrity, the evaluation of operational safety as well as the analysis of radiological consequences. An interim report, which compiles the basic data of the waste inventory, the repository design requirements, the data basis for geologic model developments, and the outline of a safety and safety demonstration concept, was already published in February 2016. The main achievements of the R and D project KOSINA are the development of four different repository designs, two per each generic geologic model (flat-bedded salt, salt pillow) based on thermomechanical calculations. This includes the implementation of a safety concept that provides containment of the radioactive materials with regard to the technical designs of waste packages, transport and emplacement technologies, and backfilling and sealing concepts. Furthermore, the integrity of the geologic barriers could be demonstrated for all four repository designs, and the long-term predictions showed no radiological releases during the demonstration period of one million years. (authors)

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DOE OSTI2019

Management and Disposal of U.S. Department of Energy Spent Nuclear Fuel - 19133

Congress created the U.S. Nuclear Waste Technical Review Board (NWTRB) in the 1987 Nuclear Waste Policy Amendments Act (Public Law 100-203) to evaluate the technical and scientific validity of activities undertaken by the Secretary of Energy to implement the Nuclear Waste Policy Act. In accordance with this mandate, the NWTRB reviewed the U.S. Department of Energy's (DOE) efforts to manage the inventory of spent nuclear fuel (SNF) that is under its control (DOE SNF) at several facilities around the country: the Hanford Site in Washington, the Idaho National Laboratory (INL) in Idaho, the Savannah River Site (SRS) in South Carolina, and the Fort St. Vrain independent spent fuel storage installation (ISFSI) in Colorado. While disposal of SNF and high-level radioactive waste (HLW) in a deep geologic repository remains the ultimate objective of the DOE nuclear waste management program, there is significant uncertainty about when such a repository will be constructed in the United States. Until disposal occurs, it is essential to manage DOE SNF in a way that will facilitate its eventual disposal, and it is also important to improve understanding of processes related to packaging and storage of the SNF that could affect its future transportation and disposal. The NWTRB's review, which was a multi-year effort, was documented in a report to Congress and the Secretary of Energy and is summarized here. The NWTRB reviewed the quantities and characteristics of DOE SNF at each of the four sites. It also examined DOE packaging and storage activities and plans at each of the sites. The NWTRB highlighted three main issue areas in its review. First, the review identified issues related to managing the aging of DOE SNF and the facilities used to store the DOE SNF. Second, the review addressed issues related to packaging of stored non-naval DOE SNF into a standard canister. Third, the NWTRB assessed issues that could affect disposal of DOE SNF if DOE continues to evaluate a range of repository settings. Based on the information and findings developed as part of its review, the NWTRB made six recommendations to Congress and the Secretary of Energy related to aging management, packaging, and disposal of DOE SNF. DOE has begun to address some of the NWTRB recommendations. (author)

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

The Challenges of Repository Operation in an Existing Mine - 19339

A generally accepted principle related to long term management of radioactive waste/ spent fuel is the isolation in a geological repository. In Germany, the disposal operation of low-level waste (LLW) and intermediate-level waste (ILW) started in the Asse II mine in 1967 and in the former German Democratic Republic in the Morsleben repository in 1971. After the German reunification, the responsibility for the existing sites changed several times until the Bundesgesellschaft fuer Endlagerung (BGE) became new license holder for both sites in 2017. The BGE is responsible to carry out the operative tasks of searching for a high-level waste (HLW) repository, the planning and construction of the Konrad repository for the emplacement of low- and intermediate-level waste (LILW), the operation of the Asse II mine including the retrieval of the emplaced waste, and the preparation for the closure of the Morsleben repository. Nevertheless, the challenges of repository operation in an existing mine have not changed in all these years, as the regulatory framework is mainly written for Nuclear Power Plants (NPPs) and is difficult to adapt for a repository mine. In particular, conditions in a mine are substantially different from 'clean' above-ground nuclear installations. Dust due to mining activities, relatively high temperatures, and a corrosive atmosphere in a salt mine might be present. Therefore, the usability of the repository for several decades during the different operational phases, i.e. planning, construction, emplacement, and closure, can only be ensured by continuous maintenance. Furthermore, the atomic law is applicable for above-ground facilities and the mining law for mines of every description. The difficulty of geological repositories is to ensure that both laws are going to be applied above and below ground with no regulatory framework tailored especially for such sites. A special case is the Asse II mine where the mining law was applied from the very beginning, before the atomic law needed to be established later on and now requires the retrieval of the radioactive waste before decommissioning. In this presentation, similarities and differences in the legal framework of the repositories in the responsibility of BGE and some operational experiences throughout the years are described. This applies to the radiation protection regime, the excavation ratio, the host rock stability and the backfilling concepts of the mining areas. Further challenges arise from the demonstration of long term safety according to the state-of-the-art of science and technology for either existing repositories/ mines or those under construction or planned. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Historic and modern nuclear graphite impurities: Pathways to improved waste strategies

Graphite has been used in large volumes as a structural material and neutron moderator since the earliest days of nuclear fission. However, no international consensus exists on the disposal of irradiated graphite, leaving much of the historic radioactive graphite inventory in interim vault or silo storage. With several new graphite-moderated reactors planned or under construction, the issue of graphite waste management is becoming increasingly urgent. This paper reviews and quantifies impurities in both historic and modern nuclear graphite, with emphasis on nitrogen—responsible for much of the 14 C inventory—and chlorine, which plays a critical role in repository performance and design. Modern graphites, benefitting from stringent quality-control measures developed for non-nuclear industries, meet or exceed the ASTM Ultra-High Purity nuclear standards, even without halide purification. Both chlorine and nitrogen concentrations have declined over time. For chlorine, identified as a key impurity influencing U.S. waste repository design, we propose a target of 0.1 appm in as-fabricated billets as a reasonable benchmark. Nitrogen sources are traced throughout the graphite production process, with surface and bulk concentrations characterized for all materials studied. Modern graphites commonly exhibit nitrogen levels below 5 appm, with values approaching 1 appm achievable. Using such reduced-nitrogen grades is critical to keeping graphite-induced radioactivity below the greater-than-Class-C waste threshold, thereby avoiding disposal cost penalties of nearly an order of magnitude.

Chlorine↗
DOE OSTI2019

Deep Isolation - Innovation for the Storage and Disposal of High-level Waste and Spent Nuclear Fuel - 19367

Deep Isolation, Inc. is a private, Berkeley based company that has brought innovation to nuclear waste disposal. Deep Isolation has developed a disposal technique that adapts mature drilling technology and customizes it for the purpose of providing a safe and secure solution. Though the stalled nuclear waste program in the US has created unwanted burdens and costs, it has also provided object lessons which Deep Isolation has learned from and designed a novel approach. The solution proposed by Deep Isolation combines the best of the geosciences and social sciences to present the US and other countries with an optimized and tailored approach to disposing of a variety of nuclear waste. This system can complement existing potential storage and disposal sites, or in the case of countries with smaller amounts of waste--the national solution. Deep Isolation innovates where others have stalled and offers a means of deep geological disposal that costs less, takes less time to implement, and cuts down and in some cases eliminates the need for transportation of the waste. The technical solution optimized by Deep Isolation uses established directional drilling technology from the oil and gas industry to drill a vertical drill hole thousands of meters deep and then transition to a horizontal drill hole that is thousands of meters in length. The target media is in or below shale formations, which are present at various depths throughout much of the United States and throughout the world. The availability of suitable geology allows for permanent disposal at multiple locations - including existing nuclear sites, without a need to transport the waste to a central repository. This innovative solution not only offers options for commercial used nuclear fuel and defense spent nuclear fuel and high-level waste, but also other nuclear wastes such as those considered by DoE's past deep bore-hole program, i.e. cesium and strontium capsules. Greater-Than-Class-C (GTCC) waste is also an opportunity for Deep Isolation. DoE's preferred alternative for GTCC includes 'land disposal at generic commercial facilities.' Deep Isolation could provide safe, secure, and permanent disposal at a commercial facility. Deep Isolation complements the technical innovation with best practices from the social sciences. In acknowledgement that the past challenges to implement nuclear waste disposal facilities have been primarily due to socio-political failings, we have designed a means to deeply engage with public and stakeholders. Our approach is committed to partnering with communities, local jurisdictions, states, and Tribes to determine a path forward that benefits all involved, that is fully transparent, collaborative and accountable. (authors)

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

Requalification and Declaration of High-alpha Cemented TRU-waste from the Late 70's and Early 80's - 20499

KTE and Siemens launched a project to qualify a series of 391 waste drums and 115 concrete containers produced from 1979 to 1984. The waste is characterized by a high content of long-lived alpha-nuclides (TRU-waste - Pu, U, Am in the range of 1 E+11-1 E+12 Bq/ drum. This high radionuclide inventory requires a special packaging according to the waste acceptance criteria (WAC) of the repository Konrad [1] despite the fact, that the waste was treated manually. The raw waste originally was delivered in nitric acid solution and was solidified using ordinary Portland cement. The existing documentation comprises analyses, transfer documentation for the nuclear fuel nuclides and waste treatment documentation yielding information, which waste was treated together in a batch. Additionally all drums were subjected to non-destructive gamma-spectrometry assay. A small number of 7 drums were opened and sampled to support the existing declaration. For the entire series the waste products were attributed to defined batches using the accountancy data supported by the analyses and the gamma-spectrometry. Each batch was produced from a homogeneous stock of waste solution and can therefore be declared and documented together. The declaration concept is based on the gamma spectrometry results performed on the entire drums, which were corrected and extended for the uncertainty of the analyses. The declaration concept is verified by the regulator and its independent expert. The state of the project is in the documentation of the solidification campaigns followed by the planning of the container configuration. (authors)

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DOE OSTI2020

Geology and Design of Major Spent Fuel Repositories

This report summarizes the geology and design concepts of four major spent fuel repositories from around the world. The crystalline rock with bentonite backfill is currently implemented in Finland with another repository being constructed in Sweden. The salt formation design focuses on the Waste Isolation Pilot Plant (WIPP) repository in the United States. The volcanic tuff open air design is based on the Yucca Mountain location. The volcanic tuff open air design at Yucca Mountain is only a concept as this repository has not been approved for operations. The last spent fuel repository design, clay/shale formation with backfill, focuses on the Cigeo Project in France, which is also pending.

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

Management of Documents and Information on Radioactive Waste - 19455

The assurance of the safety of a repository for low- and intermediate-level radioactive waste, during the 300 years of institutional control after closure, depends on assuming that the responsible institution will be capable to control the documents with information regarding the facility, the disposed of waste and the established measures for counteracting events that could breach the isolation of the radioactive material. Management of documents or information does not represent a major challenge in itself in view of consolidated practices and international standards, such as ISO 15489. However, considering the time scale involved in the disposal of the radioactive wastes, efforts should be made to ensure the preservation and proper access to documents and information. In this matter, the technological aspect of document production can be a complicating factor. In fact, it is common to be unable to access electronic documents produced 20 or 30 years ago, due either to problems related to the media where the information is stored, or to a lack of compatible systems for reading the files. This is just one example of complication when thinking about production, organization, representation, preservation and retrieval of recorded information that is to last for centuries. In this way, this work seeks to present the situation of the management of documents and information on radioactive waste in the scope of the Institute of Energy and Nuclear Research (IPEN) and with national regulation in Brazil. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

RANGERS: Methodology for Design and Performance Assessment of Engineered Barrier Systems in a Salt Repository for HLW/SNF: Synthesis Report

Salt formations have long been recognized as a highly favorable host rock for the final disposal of high-level radioactive waste (HLW) in deep geological repositories. Their unique properties, including exceptional impermeability, self-healing capabilities, and thermal conductivity, make them a reliable natural barrier for the deep disposal of radioactive waste. This report focuses on the development and application of a methodology for assessing the integrity and per formance of the Engineered Barrier System (EBS) within salt-based repositories, a critical component of the multi-barrier system ensuring safe radioactive waste disposal.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Fact Sheet - Long-length Scintillating Fibers for Radiation Detection

As the world looks to underground geological repositories for storing nuclear waste, technologies to safeguard the material are needed. An example of a project that developed a new technology is Tripwire. Tripwire proposes a multi-sensor system approach for geological spent nuclear fuel repositories that relies of radiation, vibration, and electromagnetic detection. Tripwire was developed by the Applied Radiation Measurements and Systems (ARMS) group at INL and sponsored by the National Nuclear Security Administration (NNSA). The ARMS group performs research and development, testing and evaluation, operational support, and training focused on applied ionizing radiation detection and measurement. Application areas include nuclear engineering for advanced reactors and fuel cycle operations, nuclear nonproliferation, nuclear counterproliferation, nuclear forensics, and arms control and disarmament.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2017

Public Preferences Related to Radioactive Waste Management in the United States: Methodology and Response Reference Report for the 2016 Energy and Environment Survey.

This report presents the questions and responses to a nationwide survey taken June 2016 to track preferences of US residents concerning the environment, energy, and radioactive waste management. A focus of the 2016 survey is public perceptions on different options for managing spent nuclear fuel, including on-site storage, interim storage, deep boreholes, general purpose geologic repositories, and geologic repositories for only defense-related waste. Highlights of the survey results include the following: (1) public attention to the 2011 accident and subsequent cleanup at the Fukushima nuclear facility continues to influence the perceived balance of risk and benefit for nuclear energy; (2) the incident at the Waste Isolation Pilot Plant in 2014 could influence future public support for nuclear waste management; (3) public knowledge about US nuclear waste management policies has remined higher than seen prior to the Fukushima nuclear accident and submittal of the Yucca Mountain application; (6) support for a mined disposal facility is higher than for deep borehole disposal, building one more interim storage facilities, or continued on-site storage of spent nuclear fuel; (7) support for a repository that comingles commercial and defense related waste is higher than for a repository for only defense related waste; (8) the public’s level of trust accorded to the National Academies, university scientists, and local emergency responders is the highest and the level trust accorded to advocacy organizations, public utilities, and local/national press is the lowest; and (9) the public is willing to serve on citizens panels but, in general, will only modestly engage in issues related to radioactive waste management.

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

Hydro-mechanical behavior of heated bentonite buffer for geologic disposal of high-level radioactive waste: A bench-scale X-ray computed tomography investigation

The behavior of heated bentonite buffer is critical for the security and long-term performance of a geological repository for high-level radioactive waste (HLW). While laboratory column experiments have been conducted to investigate compacted bentonite and coupled THMC (thermal-hydro-mechanical and chemical) processes for a moderate temperature range of up to 100 °C, data for a higher temperature range are limited. Understanding bentonite behavior and coupled THMC processes under higher temperatures (e.g., up to 200 °C) could allow for a more economic repository design and would expand the data and knowledge base for more reliable modeling. In this study, a bench-scale experiment was conducted in a compacted bentonite column experiencing both heating up to 200 °C in the center and hydration from a sand-clay boundary surrounding the column. During the experiment run for 1.5 years, frequent X-ray computed tomography (CT) scanning of bentonite provided insights into the spatiotemporal evolution of (1) hydration/dehydration, (2) clay swelling/shrinkage, (3) displacement, and (4) mineral precipitation. After the experiment, a comprehensive post-dismantling characterization of bentonite samples was conducted. Results showed that the bentonite hydration was axi-symmetrical despite the initial heterogeneity due to packing, confirming the ability of bentonite to seal fast flow/transport paths. Further, compared to a non-heated control experiment, the heated column showed greater CT density variations along the radial distance, indicating that homogenization of bentonite might be more difficult if a temperature gradient is maintained in the repository. Precipitation of an anhydrite layer occurred in the inner hot zone, pointing to potential concerns about salt precipitation causing canister corrosion. Ultimately, the experiments provided a high-resolution window into the strongly dynamic and coupled behavior of bentonite exposed to heating, hydration and swelling, which will be valuable for improving modeling of coupled processes, especially for the early state of a HLW repository.

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

Developing Materials for Extreme Environments and Their Use in Nuclear Waste Canisters and Reactors - 20559

The paper focuses on the development of materials which can be utilized in extreme industrial environments. The paper discusses the advancement of materials and their application in components which are utilized in extreme industrial environments as such that of the nuclear field and especially in the nuclear waste field. Nuclear power plants have become an indispensable part of the electricity generation, medical, food processing and many other industries. Though burning small amount of fuel as compared to conventional power plants, they also create waste which is not easy to handle as other types of wastes. The nuclear waste sitting at the bottom of spent fuel tanks around the globe has raised public concerns on how it will be dealt with. Europe has started to give an answer by working on geological repositories. Though till yet, waste has been cooled in spent fuel tanks and after processing, is precisely placed in nuclear waste canisters and casks and transported to sites across the country for further cooling and storage. Though in all this process there had been a great deal of advancement on the type of casks and canisters used and the processes for transporting nuclear material and waste. Often reports about radioactive material leakage, moisture content inside barrels etc. has surfaced. Be it official or unofficial report, there seems a technical failure on how the nuclear waste has been stored and placed at facilities around the globe. The other aspect of this material development is to be beneficial for nuclear reactors. As the technology gets advanced, nuclear reactors change their titles from Gen II to Gen III and so on. The new fleet of upcoming reactors will be exceptionally safer, reliable and more efficient. But these benefits will also bring drawbacks with them. As these reactors will operate at high temperature (above 800 degree Celsius) they also ask for advancement in materials used in the reactor core, fuel used, reactor control etc. The paper tries to answer such concerning questions. The work shows the study done on different material that can be better candidates for extreme environments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Modeling nuclear waste disposal in crystalline rocks at the Forsmark and Olkiluoto repository sites – Evaluation of potential thermal–mechanical damage to repository excavations

We conduct coupled thermo-hydro-mechanical modeling of a KBS-3V repository design in crystalline rocks, using data and conditions from the Forsmark in Olkiluoto repository sites in Sweden and Finland. The study focuses on repository performance related to the impact of thermal and hydraulic evolution on the potential for thermal–mechanical damage to underground repository excavations. For the designs and conditions considered at the Forsmark and Olkiluoto repository sites, the simulations show a peak temperature well under the adopted performance target of a 100°C maximum temperature, whereas there is still a high potential for thermal–mechanical damage to the KBS-3V waste deposition holes. The thermal–mechanical damage is much more likely if rock permeability is so low that it delays saturation and swelling of bentonite-clay-based backfill beyond the time for the thermal–mechanical peak, which occurs 50 to 100 years after nuclear waste deposition. We also found that sidewalls of the KBS-3V emplacement tunnels are vulnerable to tensile fracturing due to the combined effect of thermal stressing and backfill swelling. The study highlights a strong interaction between bentonite-based backfill and host rock through capillary suction along with induced rock desaturation. A careful design and selection of the bentonite-clay-based backfill materials for KBS-3V tunnels and deposition holes can facilitate a timely saturation and backfill swelling that in turn can minimize thermal–mechanical damage.

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

The Off-Site Source Recovery Program Capabilities in Recovering Transuranic Sealed Sources - 17599

Over the past 16 years, the mission of the Off-Site Source Recovery Program (OSRP) has been recovering disused radioactive sealed sources in the interest of national security, public health and safety. This program, which is sponsored by the National Nuclear Security Administration's Office of Global Material Security's Office of Radiological Security, is managed at Los Alamos National Laboratory and Idaho National Laboratory and has recovered sources worldwide. To date, OSRP has recovered more than 38,000 sources from over 1,300 locations worldwide. Since its inception, OSRP has grown from recovering only sealed sources containing one transuranic (TRU) isotope, Pu-239, in the early days of the program, to recovering sealed sources containing several other TRU and non-TRU isotopes. Primarily, OSRP's mission includes Pu-238, Pu-239, Am-241, Cf-252, Cm-244, Sr-90, Cs-137, Co-60, Ir-192 and Ra-226 sources that are US-origin, defense related and not commercially disposable. Over the lifetime of this program the types of recoveries performed have expanded from a single site to multi-site round-up recoveries. An additional approach, referred to as self-ship recoveries, is also used in cases where only a few sources may be present at a licensee's site. In these situations, the licensee ships their source(s) to an OSRP staging facility for ultimate consolidation and disposal. OSRP recovers sources from many types of licensees, with the majority of sources being removed from industry, universities and hospitals. These include sealed sources used in industrial radiological devices like moisture-density gauges and thickness gauges, medical devices such as blood irradiators and radiotherapy devices and sealed sources from nuclear power plants, among others. The recovery and management of sealed sources is a complex process covering multidisciplinary areas, including source construction, characteristics of the radioactive material, source packaging and transportation per DOT regulations, safe and secure storage, and final disposition per waste acceptance criteria of repository sites. (authors)

07 ISOTOPE AND RADIATION SOURCES↗