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At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Perspectives on Regulatory Design and Development for Transport Security of Nuclear and other Radioactive Materials

The responsibility for nuclear security rests entirely with the State but the International Atomic Energy Agency (IAEA) recognizes that a cornerstone in the development of a nuclear security regime is the development of regulations that address security of nuclear material and other radioactive material in transport. While the IAEA provides recommendations and guidance for security of materials in transport through Nuclear Security Series (NSS) documents, the challenge of integrating those recommendations into a Member State’s national legal framework is significant. Specifically, the development of transport security regulations requires coordination between both legal and technical experts within the country’s competent authority and multiple governmental agencies, as well as input from transport operators and other stakeholders. In addition to these interactions, the regulations must reflect the legal traditions of the Member State to fit within its overall national legal framework. Recognizing the challenge, the IAEA has requested input from international legal and technical experts to develop a methodology for drafting regulations on transport security of nuclear material and other radioactive material. This paper discusses the progress made so far in developing the methodology and provides a perspective on the need for transport security regulations and the challenges associated with their development. Ultimately, the goal of this paper is to introduce the IAEA-proposed methodology and to facilitate the discussion of this methodology with Member States that need support in developing their transport security regulations.

Ladsous, David↗
DOE OSTI2020

Miniaturized pulsed discharge ionization detector, non-radioactive ionization sources, and methods thereof

The present application relates to pulsed discharge ionization detectors (PDIDs) and non-radioactive ionization sources, including miniaturized forms thereof. In some examples, the PDID includes annular electrodes, where each electrode is disposed between annular insulators. Also provided herein are methods of making and using such PDIDs, such as for detecting one or more volatile organic compounds, as well as non-radioactive ionization sources.

42 ENGINEERING↗
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 OSTI2023

Radioactive waste repository when contacted by water provides borates that absorb neutrons

A container holds radioactive material. A sub-criticality controller protects the radioactive material from reaching a criticality from contact with the water. The sub-criticality controller includes a metallic composition having at least one metal component and at least one borate component bonded to the at least one metal component. The metallic composition forms borates when the metallic composition contacts the water.

Xiong, Yongliang↗
DOE OSTI2015

Development and Application of collimated spectrometric systems for the characterization of radioactive contamination of decommissioned facilities - 15030

Different approaches use collimated spectrometric systems for the characterization of radioactive contamination of various objects: soil, constructions decommissioned facilities. From time of works on elimination of Chernobyl accident and other large scale contamination accidents due radioactive dawn-falls we start to use 'CORAD' method which gives values of surface contamination of soil (in Ci/m{sup 2}) and depth of soil contamination (in MFP or cm) using measurements with collimated spectrometric detector NaI(Tl). The realizations of method for measuring of surface contamination of walls, floors using spectrometric detectors of various types were carried out in last years. We present the main results of the work - the upgraded detector for measuring contamination in wells and concrete measurements in contaminated areas of engineering rooms. Realization of method using different types of spectrometric detectors scintillating NaI(Tl) and LaBr{sub 3}, semiconductor - HPGe and room temperature CdZnTe. Different types of detector are the most suitable for different conditions of measurements. The results of measurements of contaminated soil, sediment, concrete construction floor and walls in the rooms are presented. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

Determination of attenuation coefficients of heavyweight concretes containing colemanite by using {sup 133}Ba radioactive isotope source - Paper 140

As it is well known, colemanite is a boron ore. Boron is one of the most important underground richness of Turkey. Boron is employed as a constituent material of neutron shielding because of its high absorption of thermal neutrons. Barite ore includes BaSO{sub 4}, which is a well photon radiation absorbent. The literature about radiation attenuation in heavy concretes (including boron and barite ores) is fairly abundant. Abdo and Abdo et al. theoretically calculated and determined cross section by using XCOM computer program for photon and neutron radiations shielding. Akkurt et al. measured radiation transmission of heavy concretes including normal and barite aggregates for different γ-ray energies and calculated linear attenuation coefficients. Bashter et al. studied heavy concretes including hematite-serpentine, ilmenite-limonite as control absorber in nuclear reactors γ-rays and neutron particles shielding. Bashter and Bashter et al. studied heavy concretes including hematite-serpentine, ilmenite-limonite, basalt-magnetite, ilmenite, basalt, steel and magnetite for only photon radiation shielding and calculated linear and mass attenuation coefficients from 10 keV to 1 GeV. Dealmeid et al. studied radiation transmission and angular distribution for 25 MV X-rays by using LINAC. Demir and Keles measured radiation transmission of concrete including boron waste for 59.54 and 80.99 keV γ-rays. Kaplan reported about many heavy concretes studies in his book. Kitis et al. studied heavy concretes including barite aggregates for cyclotron shielding. Mollah et al. determined concretes including ilmenite and magnetite aggregates for neutron shielding by using {sup 252}Cf radioactive source and BF{sub 3} neutron detector. Yarar and Bayuelken studied concretes including colemanite aggregates for neutron shielding. In this paper, we have measured radiation transmission of thirteen different heavy concretes produced by using barite, colemanite and normal aggregates for 276, 303, 356 and 383 keV γ-ray energies of {sup 133}Ba radioactive isotope by using CdTe semiconductor detector and calculated linear and mass attenuation coefficients. Results were compared to X-ray mass attenuation coefficient values of NIST. The attenuation of γ-ray photons is governed primarily by the atomic number and density of the concrete as a shielding material. Photon attenuation depends on pair production, photoelectric effect and Compton scattering. In this study, linear attenuation coefficients (μ) decrease with the increasing the photon energy for these concretes. Heavy elements there is a fairly sharp minimum in the absorption cross section at about this energy region. It is indicated that attenuation coefficient is useful parameter for radiation shielding. As known, the linear attenuation coefficients of the concretes used in this study at these γ-ray energies are not available in the literature by using CdTe semiconductor detectors. The data shows the effects of including boron-rich colemanite or barium-rich barite in addition to normal aggregate in the concrete mixtures. As would be expected, the barite is desirable for photon attenuation, while the colemanite is desirable for neutron attenuation. The concretes including barite and colemanite would be preferred as materials for building construction of radiation oncology departments, radiology departments, nuclear reactors and radiation research laboratories against photon and neutron particle radiations. But other aggregate properties of colemanite such as abrasion resistance and freeze-thaw durability are also found to be important for concrete but they cannot be quantified at this study. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2023

Mobile Hot Cell Digital Twin: End-of-life Management of Disused High Activity Radioactive Sources

Sealed radioactive sources are used in various settings such as nuclear facilities, universities, hospitals, and industry. When these sources reach the end of their lifespan, they become waste and are recaptured and stored long-term. A digital twin can be used in the design of a solution for recapturing spent sources, offering benefits in deployment time, operator safety, and cost reduction. We propose a digital twin framework for managing the end-of-life process of disused high activity radioactive sources and have created a prototype implementation to demonstrate its feasibility.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

THE DESIGN OF RADIAL HONEYCOMB LATTICES FOR IMPACT ENERGY ABSORPTION IN RADIOACTIVE MATERIALS PACKAGES

In this research we present a variation on the corrugation technique of honeycomb lattices, for cylindrical honeycombs, making them much easier to design for impact energy absorption in radioactive materials packages. This variation, termed radial honeycomb lattices, eliminates the residual strain and saddle effect. The use of honeycomb lattices provides advantages over the typically used foams. While foams are effective at absorbing impact energy, they can burn, their material properties are difficult to control, they can degrade over time, and procurement of raw materials can be dependent on timing of manufacturer batch runs. While the weaknesses of foam are strengths for honeycomb lattices, lattices have a different set of problems. Typically, when cylindrical honeycomb lattices are manufactured, they are manufactured flat, wrapped around a mandrel of the desired radius, and then brazed. This approach introduces residual strains, resulting in the saddle effect, which limits both the radial thickness and cylinder length. The radial honeycomb lattice approach presented here makes the design of thicker and longer cylinder honeycombs possible. To address these issues, we propose a honeycomb lattice which changes in cross-section from the inner to the outer radius of the cylinder. This causes the lattice to automatically wrap into a cylinder as it exits the corrugating gears. The theoretically bounding case, of a square cross-section at the inner radius, transitioning through a hexagon, to a diamond cross-section at the outer radius, results in a maximum thickness of approximately 41% of the inner radius. Full mathematical derivations, implemented in computer code, allow for the design of an entire radial honeycomb lattice, including the corrugating gears. To accomplish this only the radial thickness, inner cross-section shape, cell size, and nominal gear radius need to be specified, making the design of these lattices very efficient. Radial honeycomb lattice prototypes have demonstrated that the honeycomb does indeed wrap into a cylinder as intended, without the saddle effect, and can therefore be used to create thick-walled cylinders of any length. These design improvements make cylindrical honeycomb lattices much more accessible as a design element for radioactive materials packages

Johnson, William R. [Savannah River National Labor↗
DOE OSTI2026

A Mobile Hot Cell for Conditioning Disused Sealed Radioactive Sources for Storage or Transportation

An innovative Mobile Hot Cell (MHC) has been developed for conditioning Disused Sealed Radioactive Sources (DSRS) category 1 and 2 for storage or transportation. The MHC is designed to provide both Radiological and Biological containment with a maximum capacity of 1000 Ci Co60 or 5000Ci Cs137 source and can be transported via standard cargo containers. This project has been supported through the National Nuclear Safety Administration (NNSA) Offsite Source Recovery Program (OSRP). The project is intended for the international community rather than domestic although domestic use is a possibility. Many countries have significant stockpiles of these devices that are often stored in less-than-optimal circumstances. This necessitates that these devices be addressed expeditiously, and the sources secured. The MHC utilizes robotics, automation, and other non-traditional methods for disassembling, characterizing, and packaging these sources that have reached end of life or are otherwise not needed. These innovative approaches are necessary to facilitate an expedited timeline to efficiently and safely secure these sources in a non-proliferation effort. Conditioning efforts include disassembling the device such as a teletherapy head used for cancer treatment, or blood/research irradiators such that the radioactive sources may be removed safely. The sources are then characterized. Leak checks are performed, dimensions are verified, and serial numbers are confirmed. Upon completion, the sources are typically placed into a Standard Forms Capsule which is seal welded closed. It is leak tested and placed into a Long-Term Storage Shield (LTSS) which can either be secured for storage directly or loaded into an appropriate cask for transportation. Further innovations include multiple deployment scenarios that include a full deployment of MHC components, deployment of the MHC automation internal components to an existing hot cell, deployment of minimally required MHC components and incorporation of sand for shielding, and integration of the MHC for Silo Storage, or Bore Hole Storage efforts. The MHC has evolved from a very specific use case to a “Swiss Army Knife” type of a tool in that it can be readily adapted to a large variety of situations. Innovative approaches such as the use of robotics, Computer Numeric Control (CNC) machining centers, automated welding equipment, HDMI Cameras, and LED lighting are some of the developed technologies incorporated into the MHC design. Shielding is accomplished with a steel walled Base Box which is surrounded by four nesting doll shield shells which when combined limits the external dose rate to 5mr/hr when a 1000 Ci Co60 source is exposed inside.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

The Performance of Underground Radioactive Waste Storage Tanks at the Savannah River Site: A 60-Year Historical Perspective

The Savannah River Site produced weapons-grade materials for nearly 35 years between 1953 and 1988. The legacy of this production is nearly 37 million gallons of radioactive waste. Since the 1950s, the liquid waste has been stored in large, underground carbon steel waste tanks. During the past 20 years, the site has begun to process the waste so that it may be stored in vitrified and grout forms, which are more suitable for long-term storage. Over the history of the site, some tanks have experienced leakage of the waste to the secondary containment. This article is a review of the instances of leakage and corrosion degradation that the tanks and associated equipment have experienced since the first tanks were built. Furthermore, the activities that the site has taken to mitigate the degradation and manage the service life of the tank for its anticipated lifetime are reviewed.

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 OSTI2022

Radioactive waste immobilization of Hanford sludge in magnesium potassium phosphate ceramic forms

Here, this paper evaluates immobilization of Hanford K-Basin tank sludge in magnesium potassium phosphate ceramic forms. The waste forms were produced using two simulated non-radioactive sludge streams, each with distinct characteristics and composition. Ceramicrete with wollastonite as filler was used as the matrix for this purpose. The resulting waste forms were tested for their mechanical properties, radiation stability, and leaching resistance. In another series of tests, the samples were vitrified in borosilicate glass and glass waste forms were produced. The Product Consistency Test, the American Nuclear Society's ANS 16.1 test, and the Toxicity Characteristic Leaching Procedure, which are used to develop waste acceptance criteria in the United States for permanent storage of treated waste, were used for evaluation of the leaching resistance of all waste forms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Public involvement on closure of Asse II radioactive waste repository in Germany

From 1967 to 1978, about 125,800 barrels of low- and intermediate level waste were disposed of - nominally for research purposes - in the former 'Asse' salt mine which had before been used for the production of potash for many years. Since 1988 an inflow of brine is being observed which will cause dangers of flooding and of a collapse due to salt weakening and dissolution if it should increase. Since several years the closure of the Asse repository is planned with the objective to prevent the flooding and collapse of the mine and the release of radioactive substances to the biosphere. The first concept that was presented by the former operator, however, seemed completely unacceptable to regional representatives from politics and NGOs. Their activities against these plans made the project a top issue on the political agenda from the federal to the local level. The paper traces the main reasons which lead to the severe safety problems in the past as well as relevant changes in the governance system today. A focus is put on the process for public involvement in which the Citizens' Advisory Group 'A2B' forms the core measure. Its structure and framework, experience and results, expectations from inside and outside perspectives are presented. Furthermore the question is tackled how far this process can serve as an example for a participatory approach in a siting process for a geological repository for high active waste which can be expected to be highly contested in the affected regions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Development of separation technique of sodium nitrate from low-level radioactive liquid waste using electrodialysis with selective ion-exchange membranes

An advanced method, in which electrodialysis separation of sodium nitrate and decomposition of nitrate ion are combined, has been developed to remove nitrate ion from low-level radioactive liquid wastes including nitrate salts of high concentration. An engineering scale apparatus with two electro-dialytic devices, in which the sodium and nitrate ions were separately removed by each device, was produced on the basis of the results of fundamental investigation previously reported, and the performance of the apparatus was tested. Both the ions were successfully removed at the same time, though these ions were separately transferred using two electro-dialytic devices. And also effect of several experimental parameters such as current and temperature on current efficiency of both the ions of each device was investigated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Geological criteria for site selection of an LILW radioactive waste repository in the Philippines

In the selection of sites for disposal facilities involving low- and intermediate-level radioactive waste (LILW), International Atomic Energy Agency (IAEA) recommendations require that 'the region in which the site is located shall be such that significant tectonic and surface processes are not expected to occur with an intensity that would compromise the required isolation capability of the repository'. Evaluating the appropriateness of a site therefore requires a deep understanding of the geological and tectonic setting of the area. The Philippines sits in a tectonically active region frequented by earthquakes and volcanic activity. Its highly variable morphology coupled with its location along the typhoon corridor in the west Pacific region subjects the country to surface processes often manifested in the form of landslides. The Philippine LILW near surface repository project site is located on the north eastern sector of the Island of Luzon in northern Philippines. This island is surrounded by active subduction trenches; to the east by the East Luzon Trough and to the west by the Manila Trench. The island is also traversed by several branches of the Philippine Fault System. The Philippine LILW repository project is located more than 100 km away from any of these major active fault systems. In the near field, the project site is located less than 10 km from a minor fault (Dummon River Fault) and more than 40 km away from a volcanic edifice (Mt. Caguas). This paper presents an analysis of the potential hazards that these active tectonic features may pose to the project site. The assessment of such geologic hazards is imperative in the characterization of the site and a crucial input in the design and safety assessment of the repository. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Characterization of radioactive wastes with respect to harmful materials

In addendum 4 to the license of the German KONRAD repository, which considers mainly radiological aspects, a water law permit was issued in order to prevent the pollution of the near-surface groundwater. The water law permit stipulates limitations for 10 radionuclides and 2 groups of radionuclides as well as mass limitations for 94 substances and materials relevant for water protection issues. Two collateral clauses, i.e. additional requirements imposed by the licensing authority, include demands on the monitoring, registering and balancing of non-radioactive harmful substances and materials /1/. In order to fulfill the requirements of the water law permit the German Federal Office for Radiation Protection (BfS) being the operator of the KONRAD repository has developed a concept, which ensures the compliance with all requirements of the water law permit and which provides standardized easy manageable guidance for the waste producers to describe their wastes. On 15 March 2011 the competent water authority, the 'Niedersaechsischer Landesbetrieb fuer Wasserwirtschaft, Kuesten- und Naturschutz' (NLWKN) issued the approval for this concept. Being the most essential part of this concept the procedural method and the developed description of nonradioactive waste package constituents by use of standardized lists of materials and containers is addressed and presented in this paper. The waste producer has to describe his waste package in a standardized way on the base of the lists of materials and containers. For each material in the list a comprehensive description is given comprising the composition, scope of application, quality control measures, thresholds and other data. Each entry in the list has to be approved by NLWKN. The scope of the lists is defined by the waste producers' needs. Using some particular materials as examples, the approval procedure for including materials in the list is described. The procedure of describing the material composition has to be considered in the KONRAD waste acceptance requirements. The respective part of these requirements will be introduced. In order to clarify the procedure of describing waste packages by use of the standardized lists of materials and containers some examples of typical waste package descriptions will be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Prioritization of remedial actions for radioactive waste burials in the Chernobyl exclusion zone

Large volumes of radioactive waste were urgently buried in the Chernobyl Exclusion Zone (ChEZ) in the aftermath of the 1986 accident. Twenty-six years later, decisions must be taken about whether and how to remediate these sites. An attempt to resolve two key issues is described here: 1. How to assess the hazards posed by these facilities, recognizing that, for the foreseeable future, the Chernobyl Exclusion Zone will remain under institutional control? and 2. What standards to apply in deciding the extent of remediation? This paper presents an examination of the issues and proposes a simple decision-making process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗