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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Nuclear Security Interface Considerations for Regaining Control of Discovered Nuclear and Radioactive Materials

Recognizing that regaining control of orphan radioactive and nuclear material is challenging, the international community has developed requirements and guidance to support countries in addressing the problem. However, limited guidance is available on the practical nuclear security related issues that arise when nuclear or other radioactive material out of regulatory control (MORC) is encountered and efforts are undertaken to regain control. For example, the IAEA Specific Safety Guide No. SSG-19, “National Strategy for Regaining Control over Orphan Sources and Improving Control over Vulnerable Sources”, provides recommendations on a methodology for establishing a national strategy for regaining control of orphan sources. However, except for Pu239 in radioactive sources, nuclear material is outside the scope of this safety guide, and practical safety and security measures below the level of national strategy are not discussed adequately.This paper discusses the interfaces between the stakeholder groups and the responsibilities of each when encountering common scenarios that deal with MORC transitioning to regulatory control. The stakeholders include but are not limited to individuals who encounter MORC, and providers of formal and informal transport systems, and traditional and ad hoc storage solutions. This paper explores practical legal, financial, and institutional issues that hinder implementation of required safety and security practices. This paper also offers strategies that can be implemented on the individual, organizational, national, and international level to regain control of MORC.

Shannon, Michael↗
DOE OSTI2013

RECERTIFICATION OF THE MODEL 9977 RADIOACTIVE MATERIAL PACKAGING

The Model 9977 Packaging was initially issued a Certificate of Compliance (CoC) by the Department of Energy’s Office of Environmental Management (DOE-EM) for the transportation of radioactive material (RAM) in the Fall of 2007. This first CoC was for a single radioactive material and two packing configurations. In the five years since that time, seven Addendums have been written to the Safety Analysis Report for Packaging (SARP) and five Letter Amendments have been written that have authorized either new RAM contents or packing configurations, or both. This paper will discuss the process of updating the 9977 SARP to include all the contents and configurations, including the addition of a new content, and its submittal for recertification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2016

Radioactive Material Inventory Optimization Using the Plutonium-239 Equivalent Curie Methodology

ABSTRACT The purpose of this paper is to provide the technical basis for application of the Plutonium 239 equivalent curies (PEC) concept used to optimize and limit the material-at-risk (MAR) in the Irradiated Materials Characterization Laboratory (IMCL) located at the Materials and Fuels Complex (MFC). The PEC concept offers a method to normalize all radionuclides to a common radiotoxic hazard index that can be summed to a single value. This paper describes the technical approach to applying the PEC concept, documents the weighting factors and specific activities to be utilized in the calculation, and provides an example of the application of the concept to a radionuclide distribution that is representative of materials encountered in the IMCL. The IMCL is operated to provide a dynamic, re-configurable capability to analyze and characterize irradiated or non-irradiated radioactive material samples. The operations unique to the IMCL include sample storage, preparation, and analysis using experiment-specific instrumentation such as, Electron probe micro-analyzer (EPMA), Focused ion beam (FIB), Transmission electron microscope (TEM), and Scanning electron microscope (SEM). The methodology used to characterize the radioactive material inventory enables the IMCL to operate in such a way that complements their capability and program goals without compromising the safety of the worker, co-located worker, or the general public. The PEC concept offers a method in which all radionuclide inventory is normalized to a common radiotoxic hazard index that can be summed to a single value. The PEC concept is useful in the comparison of various material types for hazard categorization and for identifying worst-case and bounding inventory.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Capability Needs for Irradiated and Radioactive Materials Research (Ad hoc Committee Summary Report)

It is essential for the U.S. nuclear energy community to have access to world-leading equipment suitable for conducting research on both nuclear fuels and structural materials, including neutron irradiated–and therefore activated–materials. To address future research infrastructure requirements to support the DOE-NE mission, the Office of Reactor Fleet and Advanced Reactor Deployment established an ad hoc committee to gather information on potential capability gaps for radioactive materials and radiation effects research. This document summarizes the discussions of the committee addressing the high-level challenges in irradiated and radioactive materials research and the capabilities needed to address these challenges. After considering the various needs, the committee agreed on the four top-level targets and priority capability gaps.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Peer Review Comments on the Authorized Limits Technical Document (OREM-19-2573) for Disposal of Waste Containing Residual Radioactive Material at the U.S. Department of Energy Industrial Landfill IV, Y-12 National Security Complex, Oak Ridge, Tennessee

Argonne National Laboratory is assigned responsibility by DOE/EM-4 to conduct a peer review of the proposed Authorized Limits (ALs) for disposal of waste containing residual radioactive material at the U.S. Department of Energy (DOE) Industrial Landfill IV, Y-12 National Security Complex, Oak Ridge, Tennessee. The ALs document package (Document Number OREM-19-2573, dated May 2020) was sent to Argonne by DOE on June 1, 2020 (Oak Ridge 2020a). Argonne started the review immediately and held several BlueJeans meetings with Oak Ridge and DOE to discuss review comments. The first BlueJeans meeting was June 17, 2020. Oak Ridge presented a briefing of the AL package, followed by discussion of the assumptions, methodology, and parameters used in the analysis. A follow-up BlueJeans meeting was held on June 24, 2020, at which Argonne presented both general and specific review comments in PowerPoint slides. This set of slides - Review Comments on the “Authorized Limits for Disposal of Waste Containing Residual Radioactive Material at DOE Industrial Landfill IV, Y-12 National Security Complex, Oak Ridge, Tennessee” - was then emailed to Oak Ridge and DOE after the BlueJeans meeting on June 24, 2020. A PDF version of this slide set is included as Appendix A.

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

Sandia-developed Transport of Radioactive Material (STORM)

SAND2022-12604 O Sandia-developed Transport of Radioactive Material (STORM) reads in the release record output file and transports the material through the atmosphere to determine the potential health effects and land contamination. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.

Clayton, Daniel↗
DOE OSTI2018

Immobilization of Radioactive Materials Using Porous Iron Composite Media - 18582

High levels of nitrate (NO{sub 3}) derived from agriculture and naturally occurring radioactive materials (NORM), primarily uranium (U) and natural decay chain products that have been mobilized by the influx of oxygenated, high-alkalinity water, are a threat to shallow groundwater resources in California. Porous Iron Composite (PIC) materials were evaluated as a water treatment option in a series of batch experiments. While the use of Zero Valent Iron (ZVI) to address groundwater contaminated with NO{sub 3} or U has been demonstrated, the ability of PICs to chemically reduce and immobilize U and other associated contaminants in oxygenated alkaline groundwater where soluble uranyl-carbonate species dominate and high NO{sub 3} levels may compete for reductive capacity remain a concern. Additionally, ZVI materials have been used to immobilize other redox sensitive radionuclides (e.g., technetium-99 ({sup 99}Tc)) with limited success. Batch tests were conducted using an artificial groundwater surrogate (i.e., CAGW for California Groundwater) containing ∼ 100 mg L{sup -1} NO{sub 3}{sup -}, 200 mg L{sup -1} alkalinity, 100 μg L{sup -1} U, 100 μg L{sup -1} rhenium (Re) as a non-radioactive surrogate for the nuclear fission product technetium-99 ({sup 99}Tc), 100 μg L{sup -1} stable strontium (Sr) as a surrogate for radium (Ra) and the fission product strontium-90 ({sup 90}Sr), and 50 μg L{sup -1} arsenic (As), a common groundwater contaminant. Five gram samples of the PIC material were reacted with 200 mL of CAGW under two different atmospheric conditions to evaluate the impact of O{sub 2}. Batch tests were also conducted in the presence and absence of another oxidized species, NO{sub 3}. At specific time intervals, a small aliquot was collected from each batch treatment for chemical analysis (i.e., U, Re, As, Fe, NO{sub 3}, etc.) and the pH and ORP of the remaining suspension was measured. The PIC was effective at fully reducing NO{sub 3}{sup -} levels over the extended reaction time, even though the reactor pH quickly increased above 8.0 during equilibration. Soluble levels of U and Re decreased in a similar fashion to NO{sub 3}{sup -} in the presence and absence of O{sub 2}. Soluble As levels decreased more rapidly than other contaminants, but remained at ∼ 1 μg L{sup -1}. Very little nitrite (NO{sub 2}{sup -}) was detected in the batch tests, with NH{sub 4}{sup +} levels consistent with the amount of NO{sub 3}{sup -} reduced. Similar rates of immobilization for U, Re, and As were observed in the presence and absence of NO{sub 3}{sup -}. (authors)

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

Eddy covariance towers as sentinels of abnormal radioactive material releases

Ensuring accurate detection and attribution of abnormal releases of radioactive material is critical for protecting human health and safety. Most commonly, such detection is accomplished via active monitoring approaches involving the collection of physical samples. Further, this is labor intensive and limits the temporal and spatial resolution of any detected events to a relatively coarse level. As an alternative first step towards passive monitoring, we developed an approach using eddy flux tower data records to identify signals from a known abnormal release and quantify the extent to which that signal also occurs at other times in the data record. Through two case studies, one of which targeted the Fukushima nuclear disaster and the other targeting an abnormal release event at a radioisotope production facility in Fleurus, Belgium, we tested our approach and identified several potential heretofore unidentified abnormal events that were consistent with atmospheric circulation patterns and/or wind direction from known release sites. Because our approach is relatively simple and is resistant to systematic errors in the observational record, it has broad applicability beyond specific constituents and ecosystem types to identify a wide variety of limited-duration anomalies in flux tower data to ensure human health and industrial safety.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Distribution of Radioactive Materials in the Absheron Peninsula, Azerbaijan - 13567

The Absheron Peninsula forms the extreme Eastern part of Azerbaijan and juts into the Caspian Sea. The region has a long history of oil and gas exploration, transport, and processing and includes a number of abandoned chemical plants that were used in the separation of iodine from formation waters. As a result of lax environmental standards during the Soviet era, the industrial activity has led to serious contamination from oils residues, heavy metals and naturally occurring radioactive materials (NORM). Radiometric surveys performed over a wide range of the Absheron Peninsula showed generally low NORM concentrations. However, radiation levels two to three orders of magnitude above background levels were detected at two abandoned iodine separation plants near the capital city, Baku. These elevated radiation levels are mainly due to Ra-226 and U-238 with lower contributions from Ra-228 and U-235. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Remote detection of radioactive material using a short-pulse CO 2 laser

Detection of radioactive material at distances greater than the radiated particle range is an important goal with applications in areas such as national defense and disaster response. Here, we demonstrate avalanche-breakdown-based remote detection of a 3.6 mCi α-particle source at a standoff distance of 10 m, using 70 ps, long-wave infrared (λ = 9.2 µm) CO 2 laser pulses. This is ∼10 times longer than our previous results using a mid-IR laser. The primary detection method is direct backscatter from microplasmas generated in the laser focal volume. The backscatter signal is amplified as it propagates back through the CO 2 laser chain, enhancing sensitivity by >100 times. Here we also characterize breakdown plasmas with fluorescence imaging, and present a simple model to estimate backscattered signals as a function of the seed density profile in the laser focal volume. All of this is achieved with a relatively long-drive laser focal geometry (f/200) that is readily scalable to >100 m.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Monte Carlo characterization of the cosmic ray muon flux in shallow subsurface geological repositories intended for disposal of radioactive materials

Recent challenges in monitoring subsurface geological repositories intended for disposal of radioactive materials such as spent nuclear fuel call for new, innovative concepts that are facility independent, cost-effective, passive, and reliable. Once nuclear material is in place at these facilities, reverifying the inventory may no longer be feasible if continuity of knowledge is lost or unavailable to the inspectors. Using cosmic ray muons may present several potential advantages over conventional photon/neutron signatures, and their use in safeguards applications have only received attention in the past decade. However, there have been limited efforts to explore the integration of cosmic ray muons into repository safeguards and study potential gains, risks, and costs. This paper presents a Monte Carlo-based methodology to characterize the cosmic ray muon flux, including muon angular and energy differential distributions at depths representative of subsurface geological repositories. Since there have been limited measurements at these sites and a measurement made in one site is not always transferable to another site, the objective is to develop an efficient simulation method and useful parametrizations to provide a convenient tool for enabling muon simulations at any geological repository site. It is expected these results will provide a better understanding of how muons can be integrated into an existing geological repository safeguards framework.

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

An experimental program to characterize confined explosions of hydrogen/oxygen mixtures in the context of radioactive material transport between CEA nuclear facilities - 15125

CEA, the French Alternative Energies and Atomic Energy Commission, carries out a variety of nuclear research programs in many research reactors and nuclear facilities, decommissioning and dismantling operations that require multiple transports of different type of radioactive material such as radioactive waste or experimental fuel rods using a wide range of transportation packages. Packages are designed, constructed and operated to meet safety and regulatory requirements and must be approved by the French Safety Authority. Application for package approval (e.g. type B package) requires a complete and thorough safety demonstration in which flammable gas generation like radiolysis has become a major issue in the last few years. To face these challenges, the CEA unit in charge of transport package construction and transport operations (CEA/STMR, Cadarache Center, France), has developed a many-sided approach based on one hand on a better knowledge of both hydrogen gas generation and oxygen consumption in waste materials, and on the other hand, on design of transport packages or waste containers capable of withstanding dynamic pressure loads due to an inner hydrogen/air mixture explosion. This paper describes the experimental program that CEA/STMR has decided to implement in order to better characterize confined space explosions of hydrogen/oxygen mixtures. It presents how pressure profiles were experimentally measured in various pressure and temperature conditions which are representative of normal and accident transport conditions. It describes how different experimental set-ups were designed and improved progressively so as to obtain reliable and reproducible detonation conditions, investigate the influence of water vapor and characterize pressure profiles (peak pressure, residual pressure, peak duration). This experimental data will be used to qualify computational models and tools that are used to calculate the structural response of waste containers or fuel canisters, and finally demonstrate the safety of transport packages following an inner hydrogen explosion. (authors)

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

Constraint-Based Routing Models for the Transport of Radioactive Materials

The Department of Energy (DOE) has a historic programmatic interest in the safe and secure routing, tracking, and transportation risk analysis of radiological materials in the United States. In order to address these program goals, DOE has funded the development of several tools and related systems designed to provide insight to planners and other professionals handling radioactive materials shipments. These systems include the WebTRAGIS (Transportation Routing Analysis Geographic Information System) platform. WebTRAGIS is a browser-based routing application developed at Oak Ridge National Laboratory (ORNL) focused primarily on the safe transport of spent nuclear fuel from US nuclear reactors via railway, highway, or waterway. It is also used for the transport planning of low-level radiological waste to depositories such as the Waste Isolation Pilot Plant (WIPP) facility. One particular feature of WebTRAGIS is its coupling with high-resolution population data from ORNL s LandScan project. This allows users to obtain highly accurate population count and density information for use in route planning and risk analysis. To perform the routing and risk analysis WebTRAGIS incorporates a basic routing model methodology, with the additional application of various constraints designed to mimic US Department of Transportation (DOT), DOE, and Nuclear Regulatory Commission (NRC) regulations. Aside from the routing models available in WebTRAGIS, the system relies on detailed or specialized modal networks for the route solutions. These include a highly detailed network model of the US railroad system, the inland and coastal waterways, and a specialized highway network that focuses on the US interstate system and the designated hazardous materials and Highway Route Controlled Quantity (HRCQ) -designated roadways. The route constraints in WebTRAGIS rely upon a series of attributes assigned to the various components of the different modal networks. Routes are determined via a constrained shortest-path Dijkstra algorithm that has an assigned impedance factor. The route constraints modify the various impedance weights to bias or prefer particular network characteristics as desired by the user. Both the basic route model and the constrained impedance function calculations are determined by a series of network characteristics and shipment types. The study examines solutions under various constraints modeled by WebTRAGIS including possible routes from select shut-down reactor sites in the US to specific locations in the US. For purposes of illustration, the designated destinations are Oak Ridge National Laboratory in Tennessee and the Savannah River Site in South Carolina. To the degree that routes express sameness or variety under constraints serves to illustrate either a) the determinism of particular transport modes by either configuration or regulatory compliance, and/or b) the variety of constrained routes that are regulation compliant but may not be operationally feasible.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Comparison of German and American fabrication standards for transportation and storage containments of high level radioactive material type B packages - 15055

As the competent authority for the evaluation of German radioactive material (RAM) transport and storage packages, BAM (the German Federal Institute for Materials Research and Testing) has evaluated various international cask manufacturers regarding compliance with the German Type B(U) licensing requirements. The US NRC has recommended paragraphs in the ASME Boiler and Pressure Vessel (B and PV) Code, Section III, Division 3, as the measure to quantify the safety requirements of the US regulations regarding fabrication of RAM transportation and storage containments. The objective of this study is to compare the fabrication criteria of the German Dangerous Goods Procedural Rules (BAM-GGR) and the German Institute for Standardization (DIN) with the US quantitative measures for fabrication of RAM package containment vessels. This comparison will cover major similarities and differences between the requirements of the German standards and the ASME Code requirements for Type B package containment fabrication QC (quality control). While the material most often used for German RAM containment applications is ductile cast iron (DCI), in the US forged austenitic stainless steel is typically used. This study focuses on fabrication QC measures required to ensure reproducible and acceptable properties for DCI containments. (authors)

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

A Pill to Treat People Exposed to Radioactive Materials

Berkeley Lab's Rebecca Abergel discusses "A pill to treat people exposed to radioactive materials" in this Oct. 28, 2013 talk, which is part of a Science at the Theater event entitled Eight Big Ideas. Go here to watch the entire event with all 8 speakers:

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

A Methodology for Determining the Concentration of Naturally Occurring Radioactive Materials in an Urban Environment

The detection of radioactive sources in an urban setting is greatly complicated by natural background radiation, which emanates from various materials including roadways, sidewalks, soil, and building exteriors. The method presented and demonstrated here represents an effort to characterize the concentration of naturally occurring radioactive material (NORM) in these types of materials. The location surveyed in this work was the Fort Indiantown Gap Combined Arms Collective Training Facility in Lebanon County, Pennsylvania. Over 70 measurements with a high-purity germanium detector were performed to ascertain the NORM concentrations present in the soil, asphalt, gravel, concrete, and walls found throughout the site. Monte Carlo radiation transport calculations were used to obtain detector responses for these various geometries and materials to convert these measurements into NORM concentration estimates. Lastly, synthetic spectra were simulated using the predicted source terms and compared to actual measurements, showing acceptable agreement.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗